ML16036A381

2015-11-Proposed Written Exam

ML16036A381
Person / Time
Site:	Wolf Creek
Issue date:	11/13/2015
From:	Vincent Gaddy Operations Branch IV
To:	Wolf Creek
References
Download: ML16036A381 (208)

Text

1 ID: 98316 Points: 1.00 Step 5 in EMG E-0, REACTOR TRIP OR SAFETY INJECTION, states 'Check if SI is required'.

What is the purpose of this step?

A. Ensure BIT valves are closed so flow can be directed to the normal flow path if indications warrant.

B. To determine which signal caused the SI for future use in the procedure.

C. To ensure BOTH trains of SI have actuated and if NOT to actuate BOTH trains.

D. Maximize the time available to prevent a possible PZR over pressurization caused by going water solid.

Answer: D Answer Explanation:

Correct - per BD E-0 step 5 checks for if this SI is inadvertent and if so it stops all but one charging pump then closes the ECCS flow path valves to the RCS from the CCPs. This action maximizes the time the crew will have to restore normal letdown to prevent the PZR from going water solid and lifting a PORV.

Incorrect - ensure bit valves are closed. The RNO for this step will close the BIT valves IF the SI is inadvertent. Plausible if the student remembers that this step RNO will close the valves but doesn't understand why.

Incorrect - determine which signal caused the SI. The step asks for each signal that could have caused the SI. Plausible if the student uses the check step as a need to know what caused the SI as to if the SI is needed.

Incorrect - ensure both trains of SI have actuated. Step 4 of the procedure checks for this. Plausible as this step deals with the SI but only to determine if its actuated not if its required.

Meets the K/A because it asks for the student to have knowledge of the reasons for steps in EMG E-0 This question is RO level because it asks for bases or knowledge of EMG procedure steps Rev 0

Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98316 User-Defined ID: 98316

Reference:

BD EMG E-0 Topic: 1 RO Reason for steps in E-0 RO Importance Rating: 4.0 SRO Importance Rating: 4.6 K/A Number: 007 EK 3.01 Comments: NEW Lesson Plan Objective: LO1732313 R4, EXPLAIN the bases and any knowledge requirements for selected procedure steps.

Tier #1 Group #1 Last Used - N/A Memory 55.41 part 10 KA - Reactor trip - Knowledge of the reasons for the following as they apply to a reactor trip - Actions contained in EOP for reactor trip - Safety function 1 Modification History:

0 - Revised based on Scott's comments from 4/25/15 1 - revised based on Robs comments from 8/6/15 Rev 0

2 ID: 98317 Points: 1.00 Given the following plant conditions:

• Plant is operating at 100%

• The crew is responding to a reactor trip and SI actuation

• EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, is in progress at step 2

• PZR is 79% and up slow

• RCS pressure is 1780 psig and down slow Based on these conditions, the failure is a leak through the...

A. RCS hot leg.

B. PZR safety valve(s).

C. PZR liquid space sample valve.

D. charging header connection to the RCS loop.

Answer: B Answer Explanation:

Correct - A steam space leak will lower pressure in the RCS but PZR level will raise.

Incorrect - PZR liquid space. This leak location will cause RCS pressure and PZR level to lower. Plausible if the student doesn't understand the leak location.

Incorrect - charging header connection. This leak location will cause RCS pressure and PZR level to lower. Plausible if the student doesn't understand the leak location.

Incorrect - RCS hot leg. This leak location will cause RCS pressure and PZR level to lower. Plausible if the student doesn't understand the leak location.

Meets the K/A because it asks the student to evaluate the indications given and determine the leak is a vapor space leak via the safety valve which would be an interrelation between the vapor space leak phenomena and the valve.

RO level because it asks only for what could cause the indications given not what or where to go.

Rev 0

Question 2 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98317 User-Defined ID: 98317

Reference:

USAR 15.6.1.1 Topic: 2 RO PZR steam space leak indications RO Importance Rating: 2.7 SRO Importance Rating: 2.7 K/A Number: 008 AK 2.01 Comments: BANK - Braidwood Lesson Plan Objective: LO1610722, R8, DISCUSS the effects of the inadvertent opening of a Pressurizer safety valve transient.

Tier #1 Group #1 Last Used - 2013 Braidwood # 40 Comprehension 55.41 part 14 KA - PZR vapor space accident - Knowledge of the interrelations between the PZR vapor space accident and the following - Valves - Safety function 3 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

3 ID: 98318 Points: 1.00 Given the following:

• BB PS-455F, PZR Master Pressure Controller, is selected to P457/P456

• Then BB PI-457 suddenly fails to 2500 psig.

Which, if any, PORV Block valve(s) MUST to be closed to stop the pressure from lowering?

A. BB HV-8000A ONLY B. BB HV-8000B ONLY C. Neither PORV Block valve D. Both PORV Block valves Answer: A Answer Explanation:

Validated on desktop simulator BB HIS-455A PORV is controlled by the output of PZR master pressure controller BB PK-455. Since the selected channel failed hi the associated PORV will open but the other PORV will not. BB HV-456A is controlled by 2/4 pressure channels above setpoint only.

Correct - selected controlling pressure channel fails hi 455A will open and 456A will not. If the block valve for the open PORV were closed then RCS pressure will stabilize out until the problems can be corrected.

Incorrect - Both PORV block valve. Plausible if the student thinks this failure will affect both PORVs in the same way that the controller will operate both valves.

Incorrect - BB HV-8000B. Plausible if the student confuses which PORV the controlling pressure channel controls. If they think of this in reverse then this would be the correct response.

Incorrect - Neither PORV block valve. Plausible because if the actual pressure lowered to less than 2185 psig then the associated PORV block valve will close and since the PORV is a pilot operated valve it will lose its opening force and reclose. The non-affected PORV would be closed before the block valves could close.

Meets the K/A because the opening of the PZR PORV is a SBLOCA and it asks the student to understand how the PORVs function.

RO knowledge due to asking how the controller works and responds to inputs.

Rev 0

Question 3 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98318 User-Defined ID: 98318

Reference:

BD EMG E-1 3 RO PORV valve positions during a pressure transient Topic:

caused by a pressure channel failure RO Importance Rating: 3.9 SRO Importance Rating: 4.1 K/A Number: 009 EA 1.15 Comments: New Lesson Plan Objective: SY1301000 R3, Realize how a Pressurizer Pressure Control Channel regulates the operation of the master pressure controller, heaters, spray valves, PORV and PORV block valves.

Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 7 KA - Small break LOCA - Ability to operate and monitor the following as they apply to a small break LOCA - PORV and PORV block valve - Safety function 3 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

4 ID: 98319 Points: 1.00 Using the attached NPIS indication, determine what indication AND procedure action that is required by the crew.

A. 'A' RCP stator winding temperature is too high, trip the reactor, trip 'A' RCP B. 'B' RCP motor bearing temperature is too high, trip the reactor, trip 'B' RCP C. 'C' RCP #1 seal and bearing water temperature is too high, perform a controlled shutdown to remove 'C' RCP from service D. 'D' RCP thrust bearing temperature is too high, perform a controlled shutdown to remove 'D' RCP from service Answer: B Answer Explanation:

Correct - The value for an immediate RCP shutdown has been exceeded (bearing temp of 195F) per foldout page of OFN BB-005, RCP MALFUNCTIONS, the RCP must be shutdown immediately.

Incorrect - A RCP and trip. The value given is high but for this to cause an immediate shutdown this would have to exceed 299F. Plausible if the student confuses which values go with which component (common misconception).

Incorrect - C RCP and controlled shutdown. The value give for seal and bearing temperature is high but it would have to exceed 230F to be and immediate shutdown and this also only calls for a controlled shutdown. Plausible if the student confuses the values for immediate shutdown and controlled shutdown.

Incorrect - D RCP and controlled shutdown. The thrust bearing temperature is high but would need to be over 195F for this to cause any action for the crew. Plausible if the student confuses the values for immediate and controlled shutdown (common misconception)

Meets the K/A because asks student to interpret plant computer screen to evaluate system status and make a decision as to actions taken based on that evaluation.

RO knowledge because foldout page items for off normal Rev 0

Question 4 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98319 User-Defined ID: 98319

Reference:

OFN BB-005 4 RO RCP malfunctions that require action per OFN BB-Topic:

005 RO Importance Rating: 3.9 SRO Importance Rating: 3.8 K/A Number: 015 2.1.19 Comments: New Lesson Plan Objective: LO1732415 R2, RECOGNIZE the available situations which are addressed by procedure OFN BB-005.

Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 10 KA - Reactor coolant pump malfunctions - Conduct of ops -

Ability to use plant computers to evaluate system or component status - Safety function 4 Modification History:

0 - Revised based on Scott's comments from 4/25/15 1 - revised based on Robs comments 8/6/15 2 - replaced based on incorrect KA selection Rev 0

Rev 0 5 ID: 98320 Points: 1.00 Given the following:

• The plant is being cooled down for a refueling outage

• RCS pressure is 345 psig stable

• RCS temperature is 333°F stable

• PZR level 25% stable

• 'A' RHR pump in cooldown mode of operation

• 'B' RHR pump in STBY ECCS lineup

• Then: 'A' RHR pump TRIPS With NO operator action, RCS pressure/temperature will _______________, RHR hot leg suction isolation valve will ____________, and the RHR suction relief will ____________.

A. lower, close, remain closed B. lower, remain open, remain closed C. rise, close, lift if pressure rises to 475 psig D. rise, remain open, lift if pressure rises to 475 psig Answer: D Answer Explanation:

Wolf Creek has no auto isolation of RHR piping if pressure rises.

Correct - If RHR is lost during a plant cooldown then RCS pressure and temperature will rise. The RHR suction valves have an interlock to not open unless pressure is lower than 360 psig. If pressure rises after the valve is open there is no effect due to no interlock to close. The suction relief valve lifts at 450 psig so the valve should be open at the given pressure.

Incorrect - Lower, open, and relief closed. The suction valve will be open. The suction relief being closed is plausible if the student confuses the lift setpoint with the RHR discharge relief.

Incorrect - rise, closed and relief open. Plausible if the student recalls the interlock for opening the RHR suction valve and misunderstands that it will not reclose if pressure rises. The relief will be open at this pressure.

Incorrect - lower, closed and relief closed. Plausible if the student recalls the interlock for opening the RHR suction valve and misunderstands that it will not reclose if pressure rises. The suction relief being closed is plausible if the student confuses the lift setpoint with the RHR discharge relief.

Meets the K/A because the question poses a loss of RHR and asks if the system will isolate due to pressure rise. Since there is no auto response for the RHR isolation the only response is that of RCS pressure and its effect on the reliefs.

Rev 0

RO knowledge system interlock and design understanding Question 5 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98320 User-Defined ID: 98320

Reference:

SY1300500 5 RO RHR hot leg suction valve and suction relief valve Topic:

operations RO Importance Rating: 3.3 SRO Importance Rating: 3.7 K/A Number: 025 AK 3.02 Comments: NEW Lesson Plan Objective: SY1300500 R6, Explain the operation of the system during normal, off normal, and emergency operating modes.

Tier #1 Group #1 Last Used - N/A Fundamental 55.41 part 7 KA - Loss of RHR - Knowledge of the reasons for the following responses as they apply to the loss of RHR -

isolation of RHR low pressure piping prior to pressure increase above specified level.

Modification History:

Rev 0

6 ID: 98321 Points: 1.00 Given the following:

• Reactor power is 100%

• RO notes 053D, CCW SRG TK B LEV HILO, is lit

• BOP notes 'B' CCW surge tank level is 18% and down slow

• 'B' CCW loop is supplying all cooling Per procedure the operators will...

A. place the 'B' train CCW pumps in P-T-L, trip the reactor, trip RCP's.

B. place the 'B' train CCW pumps in P-T-L, place ALL safety related pumps in P-T-L.

C. manually align BL water to the 'B' CCW surge tank prior to level lowering to less than 15% while continuing to locate the leak.

D. manually align ESW to the 'B' CCW surge tank prior to level lowering to less than 15% while continuing to locate the leak.

Answer: D Answer Explanation:

OFN EG-004, ALR 53D 053D comes in at CCW surge tank level of 19%

The OFN will have the operators check if auto normal makeup is aligned and then if level is still lowering to manually lineup ESW and continue to look for the leak. The 15% is based on foldout page that states if surge tank level drops to less than 15% then to trip the reactor and continue with the procedure.

Correct - manually align ESW before 15% and continue to look for leak Incorrect - place B train CCW pumps in PTL and trip reactor, plausible since foldout page will direct this AFTER surge tank level lowers to less than 15%.

Plausible if the student doesn't remember when the foldout page directs this action.

Incorrect - place B train CCW pumps PTL, place all safety related pumps in P-T-L plausible since foldout page will direct this if safety loop is aligned to train being affected if level lowers to less than 15%. Plausible if the student doesn't remember when the foldout page directs this action.

Incorrect - manually align BL water plausible if student confuses normal auto makeup and ESW manual makeup.

Rev 0

Meets the K/A because it asks the operator to operate the CCW and related system for a loss of CCW which will happen if surge tank level continues to lower. It contains surge tank level, level control. At Wolf Creek the low level alarm for the surge tank doesn't come in until 19% so auto makeup should have refilled the surge tank without the alarm from coming in. The radiation alarm will not come in on a leak out of the system only a radioactive leak into the system.

For this question the loss of surge tank level is the cause of the loss of CCW since a radiation alarm would indicate a leak into the system and not be a loss of CCW only that a leak is into the system.

RO knowledge since it asks for the operator to understand the purpose and overall mitigative strategy of the procedure.

Question 6 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98321 User-Defined ID: 98321

Reference:

OFN EG-004 6 RO manual ESW makeup for CCW surge tank on Topic:

lowering level RO Importance Rating: 3.1 SRO Importance Rating: 3.1 K/A Number: 026 AA 1.05 Comments: NEW Lesson Plan Objective: LO1732414 R3, Given a procedure flow path, EXAMINE the available options for procedure actions.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Loss of CCW - Ability to operate and or monitor the following as they apply to the loss of CCW - The CCWS surge tank, including level control and level alarms and radiation alarm - Safety function 8 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

7 ID: 98322 Points: 1.00 Given the following:

• Reactor power is 100%

• PZR Master Pressure Controller output drifts low (indicator needle down scale)

Five minutes after the failure RCS subcooling will be ______ AND PZR subcooling will be A. higher, higher B. same, higher C. same, same D. higher, same Answer: D Answer Explanation:

Correct - output lowering will cause heaters to turn on to raise pressure and sprays would not respond to control pressure since the whole system is contorted by the output of the master pressure controller. A PORV will open after pressure gets to setpoint to stop the pressure rise. Since pressure is now higher and temperature is the same RCS subcooling is higher and the PZR pressure and temperature is higher so subcooling would be same.

Incorrect - higher, higher. Plausible if the student confuses pressure and temperature in the PZR since that system will stay at saturated.

Incorrect - same, same. Plausible if the student confuses the fact that the PZR will not change subcooling and thinks the RCS will follow what the PZR will do.

Incorrect - same, higher. Plausible if the student reverses the change in pressure and temperature between the RCS and PZR.

Meets the K/A because ask for understanding of subcooling with respect to the PZR.

RO because it asks about how the system works.

Rev 0

Question 7 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98322 User-Defined ID: 98322

Reference:

SY1301000 Topic: 7 RO Failure of a PZR pressure controller RO Importance Rating: 3.1 SRO Importance Rating: 3.4 K/A Number: 027 AK 1.01 Comments: NEW Lesson Plan Objective: SY1301000 R3, Realize how a Pressurizer Pressure Control Channel regulates the operation of the master pressure controller, heaters, spray valves, PORV and PORV block valves.

Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 14 KA - PZR pressure control system malfunction - Knowledge of the operational implication of the following concepts as they apply to PZR pressure control malfunctions - Definition of saturation temperature - Safety function 3 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

8 ID: 98323 Points: 1.00 Given the following:

• 087F, TURB TRIP & P9 RX TRIP, first out is lit

• BOP reports all main stop valves have closed

• RO reports that NO rod bottom lights are lit

• RO reports RX power is 55% down slow

• RO operates SB HS-1, REACTOR TRIP MAN ACTUATION, with NO change in indications

• BOP operates SB HS-42, REACTOR MAN TRIP, with NO change in indications For this condition which of the following are the breaker indications?

A.

B.

C.

D.

Answer: D Answer Explanation:

Correct - For an ATWS the normal plant lineup would be the trip breakers closed and the bypass breakers racked out.

Rev 0

Incorrect - all red lights lit. Plausible if the student thinks that the normal lineup would be all trip and bypass breakers closed for the ATWS.

Incorrect - all green lights lit. Plausible if the student thinks that after the trip switches have been actuated the breakers should be open even for an ATWS.

Incorrect - trip breakers green. Plausible if the student thinks that after the trip switches have been actuated the breakers should be open even for an ATWS.

Meets the K/A because it has the student identify the breaker positions from what is expected for an ATWS condition.

RO knowledge because it is system knowledge to determine breaker position.

Question 8 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98323 User-Defined ID: 98323

Reference:

EMG E-0 Topic: 8 RO ATWS and reactor trip breaker indicating lights RO Importance Rating: 4.2 SRO Importance Rating: 4.3 K/A Number: 029 EA 2.07 Comments: NEW Lesson Plan Objective: SY1300100 R5, EXPLAIN operation of the Rod Control System motor generators and reactor trip breakers.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - ATWS - Ability to determine or interpret the following as they apply to a ATWS - Reactor trip breaker indicating lights - Safety function 1 Modification History:

0 - changed based on Scotts feedback 4/25/15 Rev 0

9 ID: 98324 Points: 1.00 During the performance of EMG C-21, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, the following conditions exist:

• RCS cooldown rate is determined to be 125°F/hr

• All S/G NR levels are off-scale low

• RCS hot leg temperatures are lowering

• Containment pressure is normal

1. Which of the following describes how the crew is directed to control AFW flow?

2. What is the basis for this action?

A. 1. Flow is reduced and maintained at 30,000 lbm/hr to any S/G with less than 6% NR level

2. Help stabilize hot leg temperatures to establish conditions for SI termination B. 1. Flow is reduced and maintained at 30,000 lbm/hr to any S/G with less than 6% NR level

2. Maintain RCS cooldown rate C. 1. Flow is maintained greater than 270,000 lbm/hr until at least ONE S/G NR level is above 6%

2. Help stabilize hot leg temperatures to establish conditions for SI termination D. 1. Flow is maintained greater than 270,000 lbm/hr until at least ONE S/G NR level is above 6%

2. Maintain RCS cooldown rate Answer: A Answer Explanation:

Correct - Flow is reduced per step 5 but the caution above the step discusses minimum flow to any SG with NR level less than 6%. The basis discusses that after feed flow is lowered hot leg temperature will rise and step 5 is a continuous action step so flow and steam dumps will control hot leg temperature to establish conditions for SI termination.

Incorrect - flow reduced to 30,000 and maintain RCS cooldown rate. First part is correct. If the student knows cooldown needs to be slowed down then lower AFW flow would help that. Plausible if the student doesn't understand that stabilizing RCS temperature and pressure will allow for SI termination.

Incorrect - flow maintained at 270,000 and maintain RCS cooldown rate. Flow at 270,000 is the number used for all accidents until at least one SG has 6% NR level. This is not made for this case since the feed flow is helping to cause the cooldown. If the student believes this cooldown rate is appropriate. Plausible if the student doesn't understand that stabilizing RCS temperature and pressure will allow for SI termination.

Rev 0

Incorrect - flow maintained at 270,000 and establish conditions for SI termination.

Flow at 270,000 is the number used for all accidents until at least one SG has 6% NR level. This is not made for this case since the feed flow is helping to cause the cooldown. SI termination part is correct. Plausible if the student forgets that for this specific accident AFW flow is not desirable.

Meets the K/A because it asks for interrelations between the uncontrolled depressurization of all SGs and AFW flow (safety system instrumentation). This interrelation is what to do with AFW flow if this happened.

RO knowledge because it asks for mitigative strategy of the procedure.

Question 9 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98324 User-Defined ID: 98324

Reference:

BD EMG C-21 Topic: 9 RO What to throttle AFW flow to in C-21 and why RO Importance Rating: 3.4 SRO Importance Rating: 3.7 K/A Number: E12 EK 2.1 Comments: BANK - Vogtle Lesson Plan Objective: LO1732334 R3, Discuss the bases for procedural actions in EMG C-21.

Tier # 1 Group # 1 Last Used - 2011 Vogtle #65 Comprehension 55.41 part 7 KA - Uncontrolled depressurization of all steam generators

- Knowledge of interrelations between the uncontrolled depressurization of all steam generators and the following -

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features - Safety function 4 Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

10 ID: 98325 Points: 1.00 Given the following with the unit at 30% power:

• 'B' Main Feed Reg Valve fails open

• 'B' S/G level rises to the P-14 setpoint

• All S/G NR levels have remained 25% or higher

• NO operator actions What affect does this failure have on the AFW system?

A. All AFW pumps start All SMART valves throttle appropriately B. NO AFW pumps start NO SMART valves throttle C. ONLY the TDAFW pump starts NO SMART valves throttle D. ONLY 'A' and 'B' MDAFW pumps start All SMART valves throttle appropriately Answer: D Answer Explanation:

LER 15003 Wolf Creek manual reactor trip due to high steam generator level at low power Correct - The P-14 causes a reactor trip and a trip of both MFW pumps which is a start signal for the MDAFW pumps only since SG level never lowered to the low setpoint for the TD pump. The SMART valves will throttle to limit total flow to each SG.

Incorrect - all AFW pumps start. If all the AF pumps start all the SMART valves will be closed due to the TD pump running. Plausible as on a normal trip SG levels shrink lower than the low SG level AFW actuation setpoint. Since the stem states all levels are higher than that and the TDAFW pump didn't get a start signal it would not be running Incorrect - no AFW pumps start. The SMART valves would not move if no pumps start. Plausible since the stem states that no SG level is lower than the actuation setpoint. If the student doesn't understand that the P-14 causes a trip of both MFW pumps which is a start for the MDAFW pumps.

Incorrect - only the TD pump starts. The D SMART valve will throttle but the others will stay open. Plausible if the student confuses which signal starts which pump Meets the K/A because the ability to determine the state of the AFW system on a loss of MFW is asked.

RO knowledge because it asks for system understanding.

Rev 0

Question 10 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98325 User-Defined ID: 98325

Reference:

SY1505900 Topic: 10 RO trip of both MFW pumps start of MD pumps RO Importance Rating: 4.2 SRO Importance Rating: 4.3 K/A Number: 054 AA 2.04 Comments: NEW Lesson Plan Objective: SY1505900 R11, Discuss the instrumentation and controls of the Feedwater System, including trips and automatic actions of a Feedwater Isolation Signal (FWIS).

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - Loss of main feedwater - Ability to determine and interpret the following as they apply to the loss of main feedwater - Proper operation of AFW pumps and regulating valves - Safety function 4 Modification History:

0 - changed based on Scotts comments 4/25/15 Rev 0

11 ID: 98326 Points: 1.00 A station black out occurred at 0800. The crew entered EMG C-0, LOSS OF ALL AC POWER. At 0845 the crew had shed non-essential AC and DC loads per procedure from all NK batteries.

Which of the following is correct with regards to NK battery life?

A. NK11 and NK14 are expected to be totally discharged by 1200 AND NK12 and NK13 are expected to be totally discharged by 1400 B. NK11 and NK14 are expected to be totally discharged by 1400 AND NK12 and NK13 are expected to be totally discharged by 1200 C. NK11, NK12, NK13, and NK14 are expected to last until 1600 D. NK11, NK12, NK13, and NK14 are expected to be totally discharged by 1200 Answer: C Answer Explanation:

NK11 and NK14 are 1600 amp/hr batteries and NK12 and NK13 are 864 amp/hr batteries. Per BD C-0 all batteries are designed to last 240 minutes (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

Per AI 21-016, OPERATOR TIME CRITICAL ACTIONS VALIDATION and BD C-0 step 27, as long as the non-essential loads are shed off the batteries by the 60 mark all batteries will last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Correct - Having loads shed it is expected to last a full 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per AI-21-016 Incorrect - be totally discharged by 1200. This is the design capacity of the batteries but since a load shed was completed within the 60 minute window then all batteries are expected to last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Plausible if the student fails to recall why the load shed is performed.

Incorrect - NK11 and NK14 totally discharged by 1400 and NK12 and NK13 totally discharged by 1200. NK11 and NK14 are larger batteries and if the student confuses this with the load shed and time then it would be plausible.

Incorrect - NK11 and NK14 totally discharged by 1200 and NK12 and NK13 totally discharged by 1400. NK11 and NK14 are larger batteries and if the student confuses this with the load shed and time then it would be plausible.

Meets the K/A because it asks for battery discharge rate over time (capacity).

RO knowledge because it asks about system response to transient or how the system works.

Rev 0

Question 11 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98326 User-Defined ID: 98326

Reference:

AI 21-016 11 RO Loss of all AC battery discharge rate vs time with Topic:

load shed RO Importance Rating: 3.3 SRO Importance Rating: 3.7 K/A Number: 055 EK 1.01 Comments: NEW Lesson Plan Objective: LO1120201 R3, Solve for various parameters in simple AC and DC circuits using the following: Power Equations Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 8 KA - Loss of offsite and onsite power - Knowledge of the operational implications of the following concepts as they apply to the station blackout - Effect of battery discharge rates on capacity - Safety function 6 Modification History:

0-Rev 0

12 ID: 98327 Points: 1.00 A loss of offsite power has occurred.

• The Control room staff responded using the appropriate procedure.

• Four minutes into the event the RO re-scans the alarm panels and notes the following alarms locked in.

Which of the following alarms would be the highest priority for the operators to address AND why?

A. 014D, S/U XFMR TROUBLE, restoring power B. 039A, LTDN HX TEMP HI DIVERT, RCS chemistry concerns C. 055B, ESW PMP B PRESS LO, cooling flow to the EDG D. 077E, SR HI VOLT FAIL, energizing SR detectors Answer: C Answer Explanation:

ALR 055C, OFN EF-033 Correct - with the loss of offsite power the only power left is safety related EDGs which are cooled by ESW. The ALR for 055C entry condition state that the pump has tripped and locked out so with no cooling water the EDG will trip on high temperature Incorrect - 014D comes in on any loss of normal power to it so with a loss of offsite power it will be in until power is restored. Not a priority over the EDG.

Plausible if the student believes that offsite power is more important than safety related power.

Incorrect - 039A comes in on the loss of offsite power due to the loss of CCW for a time and the letdown outlet temperature rising. Plausible if the student misunderstands this is a normal alarm for the loss of offsite power and thinks since there is no cleanup of the RCS water chemistry would be negatively affected.

Incorrect - 077E should be in until the SR detectors energize which should take place about 9 minutes after the reactor trip. Plausible if the student doesn't realize that at four minutes this should still be in.

Meets the K/A because this asks for which alarm is more important than other alarms.

RO knowledge because this is an entry condition into a procedure ALR.

Rev 0

Question 12 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98327 User-Defined ID: 98327

Reference:

ALR 055C 12 RO Priority of alarms on loss of offsite power with ESW Topic:

issues RO Importance Rating: 4.1 SRO Importance Rating: 4.3 K/A Number: 056 2.4.45 Comments: NEW Lesson Plan Objective: LO1732443 R2, RECOGNIZE the available situations which are addressed by OFN EF-033.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Loss of offsite power - Emergency procedures - Ability to prioritize and interpret the significance of each annunciator or alarm - Safety function 6 Modification History:

Rev 0

13 ID: 98328 Points: 1.00 The plant is operating at 75% power when a loss of NN02 occurs.

If NN02 can NOT be restored which of the following sets of instruments will still be available to monitor actual plant parameters?

A. AE LI-553, 'C' S/G level BB LI-461, PZR level B. AB PI-525, 'B' S/G Pressure BG LI-185, VCT level C. BB PI-455, PZR pressure BB PI-405, RCS WR pressure D. AB FI-522, 'B' S/G steam flow SE NI-42, Power Range Answer: C Answer Explanation:

Correct - All instruments given are powered from other NN buses Incorrect - AE LI-553, this is NN02 powered. Plausible if the student confuses instrument power supplies.

Incorrect - SE NI-42, this is NN02 powered. Plausible if the student confuses instrument power supplies.

Incorrect - AB PI-525, this is NN02 powered. Plausible if the student confuses instrument power supplies.

Meets KA asks if student knows what the backup indications are for a loss of NN bus RO knowledge power supplies Rev 0

Question 13 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98328 User-Defined ID: 98328

Reference:

SY1506300 Topic: 13 RO NN02 powered instruments RO Importance Rating: 3.2 SRO Importance Rating: 3.4 K/A Number: 057 AA 1.05 Comments: NEW Lesson Plan Objective: SY1506300 R7, Integrate system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 1 Group # 1 Last Used - N/A Fundamental 55.41 part 7 KA - Loss of vital AC electrical instrument bus - Ability to operate and or monitor the following as they apply to the loss of vital AC instrument bus - backup instrument indications Modification History:

Replaced Rev 0

14 ID: 98329 Points: 1.00 STS KJ-005A, MANUAL/AUTO START, SYNC & LOADING OF EDG NE01, is in progress. NE01 has been started from the Control Room and is running with all operating parameters indicating normal. The crew is about to sync NE01 with NB01 (offsite power) when NE01 output voltage goes to 0.

1. Which of the following will cause the above indications?

2. How must NE01 be shutdown given this failure?

A. 1. Loss of NK04 DC bus

2. By manually closing the fuel racks due to a loss of DC control power.

B. 1. Loss of NK01 DC bus

2. By manually closing the fuel racks due to a loss of DC control power.

C. 1. Loss of NK04 DC bus

2. From NE-107 using the STOP pushbutton because the control room switch has lost power.

D. 1. Loss of NK01 DC bus

2. From NE-107 using the STOP pushbutton because the control room switch has lost power.

Answer: B Answer Explanation:

NK01 supplies DC control power to NE01 for field flashing and control from the control room. If this DC is lost when the diesel is running the output voltage will go to 0 and the engine must be shutdown locally since no power to the control room switch no exists.

Correct - per the OFN notes (and the lesson plan) if NK01 is lost with NE01 operating the output voltage will drop to zero and the diesel cannot be synced with the bus. The diesel cannot be shutdown from the control room either it must me locally stopped using the manual lever on the fuel racks.

Incorrect - loss of NK04 and local fuel rack shutdown. NK04 is the control power to NE02. The second part is correct with this loss the fuel racks are the only way to stop the engine. Plausible if the student forgets which DC power supplies which EDG.

Incorrect - NK01 and shutdown from NE-107. The DC power is correct. The local panel and the control room use the DC power to control the EDG so with this loss the local pushbuttons will not work either. Plausible if the student forgets that the same DC power controls the switches in the control room and the local panel.

Incorrect - NK04 and shutdown from NE-107. NK04 is the control power for NE02. The local panel and the control room use the DC power to control the EDG so with this loss the local pushbuttons will not work either. Plausible if the student forgets that the same DC power controls the switches in the control room and the local panel.

Rev 0

Meets the K/A because it asks for knowledge of the loss of DC control power to the diesels has on the diesels. Student must recognize the reason for closing the fuel racks vice just pushing the stop pushbutton which has lost power.

RO knowledge because it asks for system knowledge of the emergency diesel.

Question 14 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98329 User-Defined ID: 98329

Reference:

OFN NK-020 Topic: 14 RO Loss of NK01 with NE01 running affects RO Importance Rating: 3.4 SRO Importance Rating: 3.7 K/A Number: 058 AK 3.01 Comments: NEW Lesson Plan Objective: SY1406401 R4, Assess the functional interrelationship with the DC Distribution System, including electrical power supplies.

Tier # 1 Group # 1 Last used - N/A Comprehension 55.41 part 5 KA - Loss of DC power - Knowledge of the reasons for the following responses as they apply to the loss of DC power -

Use of dc control power by EDGs - Safety function 6 Modification History:

0 - changed based on Scotts feedback 4/25/15 Rev 0

15 ID: 98330 Points: 1.00 Given the following:

• The unit is operating at 100%

• A failure in the normal feeder breaker for NB01 causes the breaker to open

• As the crew is working through the appropriate procedure the RO notes that the ESW system parameters are NOT as expected Which ONE of the following valves and positions would cause abnormal ESW operating parameters?

A. EF HIS-60, ESW TRN B FROM CCW HEAT EXCHANGER, is CLOSED and needs to be OPEN B. EF HIS-32, ESW TRN B TO CTMT AIR COOLERS, is OPEN and needs to be CLOSED C. EF HIS-37, ESW TRN A DISCHARGE TO UHS, is OPEN and needs to be CLOSED D. EF HIS-41, ESW TRN A TO SERVICE WTR SYS, is OPEN and needs to be CLOSED Answer: D Answer Explanation:

OFN EF-033 attachment B valve positions for actuation After an ESW system actuation signal valves reposition automatically Correct - EF-41 should be closed on an actuation of the ESW system since flow is directed back to the UHS and not to service water Incorrect - EF-32 should be open for an actuation to allow full flow through the containment coolers. Plausible if the student confuses which valves in the ESW system reposition and to what position.

Incorrect - EF-37 should be open to allow for the ESW system to discharge back to the UHS. Plausible if the student mistakes the return back to service water and UHS.

Incorrect - EF-60 should be closed since the bypass valves are throttled for proper flow through these heat exchangers. Plausible if the student confuses the bypass and the normal outlet valve positions.

Meets the K/A because it asks for ability to interpret indications and understand how actions taken can correct these indications.

RO knowledge because it asks for ESW system operations understanding.

Rev 0

Question 15 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98330 User-Defined ID: 98330

Reference:

OFN EF-033 Topic: 15 RO ESW valve mispositioning effect on system RO Importance Rating: 4.2 SRO Importance Rating: 4.4 K/A Number: 062 2.2.44 Comments: MODIFIED - 91855 Lesson Plan Objective: SY1408900 R7, EXPLAIN the instrumentation and controls of the ESW System, including symptoms/failure modes.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 5 KA - Loss of nuclear service water - Equipment control -

Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions - Safety function 4 Modification History:

0 - modified based on feedback from Scott 4/25/15 Rev 0

16 ID: 98331 Points: 1.00 Given the following:

• A LOCA outside containment has occurred

• SI has actuated

• EMG C-12, LOCA OUTSIDE CONTAINMENT, has been entered Which ONE of the following actions will be attempted to isolate the break AND which indication is used to determine if the leak has been isolated?

A. Close RCP seal water return isolation valves (BB HIS-8141A, B, C, D) and monitor PZR level.

B. Close RCP seal water return isolation valves (BB HIS-8141A, B, C, D) and monitor RCS pressure.

C. Isolate RHR to Accumulator Injection Loop (EJ HIS 8809A) and monitor RCS pressure.

D. Isolate RHR to Accumulator Injection Loop (EJ HIS 8809A) and monitor PZR level.

Answer: C Answer Explanation:

EMG C-12 Correct - RHR to accumulator injection loop and RCS pressure. This is the only indication that is looked for in C-12. The RHR cold leg injection valve is specifically called out.

Incorrect - RCP seal return isolation valves and PZR level. C-12 doesn't look at PZR level as an indication that the leak is isolated. RCP common seal return valve is closed not each individual valve. Also the individual valves are inside containment not outside. Plausible if the student doesn't understand what indications to look for in C-12 for leak isolation and if RCP seal return valves are mistaken for being outside of containment.

Incorrect - RCP seal return isolation valves and RCS pressure. RCP wrong for reasons above. RCS pressure is correct. Plausible if the student thinks the RCP seal return valves are outside of containment.

Incorrect - RHR to accumulator injection loop and PZR level. RHR valves are correct but PZR level is not for reasons given above. Plausible if the student doesn't understand what indications to look for in C-12 for leak isolation.

Meets the K/A by having the crew isolate items and then check indications so knowledge of operational implications and indicating signals dealing with LOCA outside containment.

RO knowledge because it asks for system knowledge of the RHR system and an understanding of the isolation indication Rev 0

Question 16 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98331 User-Defined ID: 98331

Reference:

EMG C-12 16 RO What to isolate and what to look for in LOCA outside Topic:

containment C-12 RO Importance Rating: 3.5 SRO Importance Rating: 3.9 K/A Number: E04 EK 1.3 Comments: BANK - Indian Point Lesson Plan Objective: LO1732333 R3, EXPLAIN major actions accomplished by procedure EMG C-12.

Tier # 1 Group # 1 Last Used - 2010 Indian Point NRC exam Comprehension 55.41 part 7, 10 KA - LOCA outside containment - Knowledge of the operational implications of the following concepts as they apply to the LOCA outside containment - Annunciators and conditions indicating signals and remedial actions associated with the LOCA - Safety function 3 Modification History:

Added justification items 4/25/15 Rev 0

17 ID: 98332 Points: 1.00 The plant was tripped from full power due to a LOCA outside of containment. The crew has transitioned to EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION.

Which ONE of the following signals will the crew reset while in this procedure AND why?

A. P-4 (reactor trip) to allow reset / re-arming of SI.

B. Phase A isolation to restore control of valves to operators.

C. ONLY the SI signal to prevent ECCS pumps from restarting when they are placed in Normal-After-Stop.

D. ONLY the SI (RWST) signal to prevent possible loss of RWST inventory to the containment sump and damage to the RHR pumps.

Answer: D Answer Explanation:

System knowledge of an SI signal and the RWST water level coming down makes this question not a procedure knowledge question but understanding how the system works. The SI reset which was performed in the previous procedures that sent you to C-11, is only part of the swapover for the RWST reset logic.

There is a separate switch for the SI (RWST) swapover that now needs to be reset or the swapover will still occur. The original SI reset just allowed ECCS pumps and valves to be operated as normal without the signal changing any operations of the equipment back to its safeguards lineup.

Correct - Since the SI signal has been reset this will allow the reset of the SI (RWST) signal now to stop the auto swapover to the CTMT sump (that is empty) which would allow RWST water to be lost from injection and damage to the RHR pumps due to no suction source.

Incorrect - only the SI signal. This has been reset in other procedures prior to this point but if this were reset (which is needed to reset the SI (RWST) signal then this would only be part of the signals that need reset in this procedure.

Plausible if the student doesn't understand how the SI and the swapover for the RWST signals work together since the LOCA is outside of CTMT there is no water in the sump for the RHR pump to use if an auto swap were to occur.

Incorrect - P-4. This is an active signal since the plant was tripped. Plausible if the student thinks that P-4 needs to be reset to allow for a second SI for this accident.

Incorrect - Phase A. This is an active signal since the plant was tripped from full power. Plausible if the student thinks that the phase A signal is preventing the proper positioning of valves.

Meets the K/A because it asks for the knowledge of interrelations between the loss of recirc and a component (RHR). Since this procedure assumes there is no water in the CTMT sump protecting the RHR pumps is an interrelation for the loss of recirc.

Rev 0

RO knowledge because it does not ask for the procedure step it only puts the crew in the procedure but asks what signals should be reset to protect what equipment. This is system knowledge of the SI signal and the RWST swapover signal.

Question 17 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98332 User-Defined ID: 98332

Reference:

EMG C-11 Topic: 17 RO which signals get reset in C-11 and why RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: E11 EK 2.1 Comments: NEW Lesson Plan Objective: LO1732332 R3, SUMMARIZE the major action categories and the bases for the steps that accomplish each category.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Loss of emergency coolant recirc - Knowledge of the interrelations between the loss of emergency coolant recirc and the following - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features - Safety function 4 Modification History:

0 - revised based on feedback from Scott 4/25/15 Rev 0

18 ID: 98333 Points: 1.00 Given the following plant conditions:

• The plant is at 100% power

• DG 'A' has been paralleled with bus NB01 and is carrying 5.8 MWe of load in accordance with STS KJ-015A, MANUAL/AUTO FAST START SYNCHRONIZATION AND LOADING OF EDG NE01

• ANN 019E, NB01 BUS DGRD VOLT, alarms

• 55 seconds later, BKR 13-48, #7 XFMR OUTPUT BKR, trips open With NO Operator actions, what is the current status of the 'A' Train Safeguards Power system?

A. NB01 Normal Feeder Breaker is CLOSED, NE01 will stop.

B. NB01 Emergency Supply Breaker is OPEN, NE01 will stop.

C. NB01 Normal Feeder Breaker is CLOSED, NE01 will remain running.

D. NB01 Emergency Supply Breaker is OPEN, NE01 will remain running.

Answer: C Answer Explanation:

OFN NB-042, SY1406401 R7 Correct - NB01 Normal Feeder Breaker will remain CLOSED, NE01 will remain running. Once the 13-48 breaker detaches the NB bus from the switchyard, the NB voltage returns to normal values (or a little higher), and the Normal supply breaker remains shut. The EDG remains running despite the load reject.

Incorrect - NB01 Normal Feeder Breaker will remain CLOSED, NE01 will stop.

Right breaker position, D/G will not stop. The loss of load that the EDG will experience when the BKR 13-48 Normal supply breaker opens will not overspeed the EDG.

Incorrect - NB01 Emergency Supply Breaker will OPEN, NE01 will stop. Breaker will remain closed, D/G will not stop. -- There is nothing in the current condition to cause the EDG output breaker to open, and the EDG will not overspeed or stop itself.

Incorrect - NB01 Emergency Supply Breaker will OPEN, NE01 will remain running. Wrong breaker position, right NE01 status. There is nothing in the current condition to cause the EDG output breaker to open.

Comments: ALR 00-019E states that if the EDG is in parallel then to stop parallel operation. OFN NB-042, LOSS OF OFFSITE POWER TO NB01 (NB02)

WITH EDG PARALLELED requires removing power from the switchyard first and then opening the EDG breaker to cause the UV on the bus for emergency operation.

Rev 0

Additional insights: This question is essentially the basis for why OFN NB-042 was created and why we keep a dedicated operator during diesel runs with the engine paralleled. The issue is that an upstream transient may be isolated prior to the bus feeder breaker. If the bus feeder breaker remains closed, the diesel will not shift to emergency mode. Without an SI present, 55 seconds is not long enough for the degraded voltage condition to open the bus feeder breaker. The transformer feeder breaker opens, but there is no transformer lockout. The diesel is tested by surveillance to withstand up to full load reject. The EDG will restore voltage once divorced from the degraded offsite source. The voltage will likely go high, but not excessively due to the power factor limitations in the STS procedure. The operators will perform the actions of OFN NB-042 to place the EDG in emergency mode.

Meets the K/A because it asks for ability to interpret grid disturbances and status of the EDG RO knowledge since this is system understanding of the output breaker of the EDG and NB01 normal power supply breaker.

Rev 0

Question 18 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98333 User-Defined ID: 98333

Reference:

OFN NB-042 18 RO EDG status after loss of offsite power to NB bus with Topic:

EDG on bus RO Importance Rating: 3.9 SRO Importance Rating: 4.3 K/A Number: 077 AA 2.09 Comments: BANK - 45856 Lesson Plan Objective: SY1406401 R7, INTEGRATE system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 1 Group # 1 Last Used - 2011 Wolf Creek NRC Comprehension 55.41 part 7 KA - Generator voltage and electric grid disturbances -

Ability to determine and interpret the following as they apply to generator voltage and electric grid disturbances -

Operational status of emergency diesel generators - Safety function 6 Modification History:

added to pedigree Rev 0

19 ID: 98334 Points: 1.00 Given the following:

• Unit is operating at 100% power

• Main Turbine is operating in Open Loop mode

• Rod Control selector switch is in Auto

• Control Bank 'D' rods are parked at 231 steps

• Control Rod B10 drops fully into the core

• NO operator actions have been taken Which of the following statements is correct with regard to this event AND why?

A. Tave will return to program due to rods stepping OUT.

B. RCS pressure will remain lower than program due to the lower Tave.

C. Main Turbine electrical load lowered due to lower S/G pressures.

D. Tref returns to previous value due to control valves not changing position.

Answer: C Answer Explanation:

SY1511701 4.1.1 Open Loop/Valve Management In OPEN LOOP, the main turbine control system will maintain a steady control valve position without any feedback mechanism. Ovation will position the turbine control valves as needed to establish its interpretation of megawatts based on valve position curves developed by the vendor. Valve curves have been updated and more will be more finely tuned by DCP 14452 (Rev 1) software upgrade.

With a dropped rod at 100% power Tave will lower causing auto rod motion OUT.

Since D bank rods are already at 231 they will only move out to 232 and stop (C-11). With this small rod step and the fact that rods don't have much 'bite' at this height Tave will not return to target and will stay lower than program. This lower Tave will cause the SG pressures to lower thereby lowering overall steam flow to the main turbine. Since the turbine is in open loop mode the control valves will receive no feedback to change valve position so the overall affect is lower electrical output with no operator action.

Correct - main turbine load will lower due to lower SG pressures due to lower Tave Incorrect - Tave will return to program. Tave will remain lower since control rods can only step 4 steps and don't affect temperature much at this height. Tave will stay lower even with the rods stepping out. Plausible if the student thinks that rods will return temperature back to program.

Rev 0

Incorrect - RCS pressure will remain lower than program. Pressure is dictated by PZR program from the master pressure controller which is not affected by the dropped rod or any other things that happen due to this. Plausible if the student confuses the pressure controller with the level which is controlled from Tref and Tave.

Incorrect - Tref returns to previous value. Plausible if the student confuses what controls Tref. Tref is controlled by first stage impulse pressure which will remain lower due to lower SG pressure and valves not moving in this mode of operation Meets K/A because it asks for knowledge of rods and turbine load changes (reason)

RO knowledge because this is fundamental understanding of the reactor and the secondary plant Question 19 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98334 User-Defined ID: 98334

Reference:

LO1130641 Topic: 19 RO main turbine impacts from a dropped rod RO Importance Rating: 3.2 SRO Importance Rating: 3.7 K/A Number: 003 AK 1.01 Comments: NEW Lesson Plan Objective: LO1130641 R22, Explain reactor response to a control rod insertion.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 5 KA - Dropped control rod - Knowledge of the operational implications of the following concepts as they apply to dropped control rod - Reason for turbine following reactor on dropped rod event - Safety function 1 Modification History:

Rev 0

20 ID: 98335 Points: 1.00 Given the following:

• Plant is in MODE 5.

• Chemistry reports that RCS boron concentration is now 150 ppm lower than the previous sample.

• The crew recognizes that Shutdown Margin is NOT being met, and initiates an emergency boration using 'A' BA Transfer pump.

• The RO reports BG FI-183A, EMERG BORATE FLOW, indicates 25 gpm.

Based on this information, the crew will A. align the RWST to the CCP suction.

B. check when emergency boration can be stopped.

C. align a Safety Injection Pump for Emergency Boration.

D. check if normal letdown can be established with current plant conditions.

Answer: A Answer Explanation:

Correct - if flow is less than 30 gpm then the procedure directs using the RWST as a suction source Incorrect - check when emergency boration can be stopped. Plausible if the student thinks this is enough flow from the emergency boration flow path.

Incorrect - Use an SI pump. This alignment is used when Charging System is NOT aligned as operable flowpath (step 1) and since the stem indicates the crew has passed that step because they have indicated flow. Plausible if the student confuses the low flow with the need to use a different pump.

Incorrect - check if normal letdown can be established. Plausible if the student thinks the emergency boration flow rate is good and wants to continue to put the system back in a normal alignment.

Meets K/A because after the BA pump is started the ability to monitor its normal operation is needed to understand that 25 gpm is not enough flow from it so it is broke and the procedure has the crew use a different pump for this.

RO knowledge to understand that anything less than 30 gpm for emergency boration is to low and the next source needed.

Rev 0

Question 20 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98335 User-Defined ID: 98335

Reference:

OFN BG-009 20 RO emergency boration due to loss of SDM while in Topic:

mode 5 RO Importance Rating: 3.7 SRO Importance Rating: 3.5 K/A Number: 024 AA 1.02 Comments: BANK - 98095 Lesson Plan Objective: LO1732419 R3, Given a procedural flow path, EXAMINE the available options for procedure actions.

Tier # 1 Group # 2 Last Used - N/A Memory 55.41 part 10 KA - Emergency boration - Ability to operate and or monitor the following as they apply to emergency boration - Boric acid pump - Safety function 1 Modification History:

added to pedigree Rev 0

21 ID: 98336 Points: 1.00 A reactor startup is in progress. The RO has just completed pulling Control Rods and is checking nuclear instrumentation. The following indications are observed:

Based on the observed indications above which detector is malfunctioning AND can the startup continue?

A. SR NI N-31, NO B. SR NI N-32, YES C. SR NI N-31, YES D. SR NI N-32, NO Rev 0

Answer: A Answer Explanation:

Proper overlap is SR at ~ 2X104 and IR ~ 1X10-10.

OFN SB-008, attachment P step p12 if you can block the channel and power is above P-6 then continue the startup and fix it later.

Correct - N-31 is reading too low compared to the others.

Incorrect - N-32 and NO. It is showing proper overlap with both the IR detectors for this power. Plausible if the student doesn't know where the overlap between the SR and IR is. Second part is correct.

Incorrect - N-32 and YES. This overlap is correct based on both of the other IR detectors agreeing with it. Per the OFN the start up must be stopped. Plausible if the student misunderstands overlap and TS for the SR detectors.

Incorrect - N31 and YES. Correct detector but per the OFN the startup must be stopped. Plausible if the student misunderstands that the detector is still needed at this power level per TS.

Meets the K/A because it asks for ability to determine SR and IR overlap with the malfunction of a SR detector RO knowledge since this is system knowledge of the excore NIs Rev 0

Rev 0 Excore NIS Ranges "ABSOLUTE" POWER INTERMEDIATE POWER POST-ACCIDENT RANGE RANGE POWER RANGE 4

10

-3 2 10 10  % 100 100

-4 10 P 50 10

-5 W 0 10 10 SOURCE A R 0 1 RANGE -6 M 10 -1 P POST-ACCIDENT 10

-7  %

% -2 E 10 SOURCE RANGE -2 P 10 10 R -8 P O 6 10 -3 10 E O 10 W S -9

-4 5 10 5 W -4 E 10 10 10 10

-10 E R 4 -5 4 10 R 10 10 10 C -11

-6 3 10 3 -6 10 P 10 10 10 S C 2 2 -7 10 P 10 10

-8 1 S 1 -8 10 10 10 10 0

10 1

-10 -1 10 10 Westinghouse Gamma Metrics For Training Use Only K:\TRNG_CommonDrawings\SE\SE01.vsd

Question 21 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98336 User-Defined ID: 98336

Reference:

SY1301501 Topic: 21 RO SR overlap with IR RO Importance Rating: 3.1 SRO Importance Rating: 3.5 K/A Number: 032 AA 2.04 Comments: NEW Lesson Plan Objective: SY1301501 R3, Explain the operation of the Excore Nuclear Instrumentation System Source Range channels.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 7 KA - Loss of SR NI - Ability to determine and interpret the following as they apply to the loss of SR NI - Satisfactory source range intermediate range overlap - Safety function 7 Modification History:

0 - modified the stem and distractors based on Scotts feedback 4/25/15 Rev 0

22 ID: 98337 Points: 1.00 Given the following:

All Shutdown rods have been withdrawn An Intermediate Range NI detector fails high The Level Trip Bypass Switch is placed in Bypass After the switch is placed in bypass, annunciator 82D, IR HI FLUX ROD STOP, light will be ________ AND the intermediate range 'instrument power on' light will be ________.

A. ON, OFF B. OFF, OFF C. ON, ON D. OFF, ON Answer: D Answer Explanation:

Ran on desktop simulator to determine which alarms were received and not received.

Correct - After the bypass switch is placed in bypass the IR rod stop alarm will clear and the instrument will still have power so the power on light would still be lit.

Incorrect - OFF, OFF. Plausible if the student doesn't understand the function of the bypass switch which is to bypass the trips and rod stops from the IR detector Incorrect - ON, OFF. Plausible if the student reverses the logic for the use of this switch.

Incorrect - ON, ON. Plausible if the student doesn't understand the function of the bypass switch and thinks that the switch only is used to test the detector.

Meets K/A because asks ability to use the bypass switch RO knowledge because system knowledge Rev 0

Question 22 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98337 User-Defined ID: 98337

Reference:

SY1301501 Topic: 22 RO IR trip bypass switch understanding RO Importance Rating: 3.0 SRO Importance Rating: 3.1 K/A Number: 033 AA 1.02 Comments: NEW Lesson Plan Objective: SY1301501 R4, Explain the operation of the intermediate range nuclear instrumentation detector.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 2 KA - Loss of IR NI - Ability to operate and or monitor the following as they apply to the loss of IR NI - Level trip bypass - Safety function 7 Modification History:

0 - changed based on Scotts feedback 4/25/15 Rev 0

23 ID: 98338 Points: 1.00 Which of the following ALARMING radiation monitors is listed as an entry condition for OFN BB-006, HIGH REACTOR COOLANT ACTIVITY?

A. SJ RE-01, CVCS Letdown Monitor B. SD RE-39, Reactor Seal Table Area Monitor C. GT RE-59, CTMT High Range Area Monitor D. GT RE-31, Containment Atmosphere Monitor Answer: A Answer Explanation:

Correct - this monitor is listed as the only rad monitor that is a direct entry into the off normal procedure. It monitors the letdown line for activity in the RCS water which would be an indication of high RCS activity.

Incorrect - GT RE-59. This monitor is the containment high range rad monitor whose function is to give rad levels in the event of a LOCA inside containment.

This monitor would show higher than normal radiation levels but since it only shows a high range the letdown monitor would alarm well prior to this one showing any change since the bottom range is 1 R. Plausible if the student confuses monitor functions.

Incorrect - SD RE-39. This monitor checks for high radiation at the reactor seal table which would show higher than normal rad levels but is not intended to be an RCS activity monitor. Plausible if the student confuses monitor functions.

Incorrect - GT RE-31. This monitor checks for high containment airborne activity.

This is an atmosphere monitor so this would look for a leak of RCS water into containment. Plausible if the student confuses monitor functions.

Meets the K/A because it asks for the monitor that the procedure will look for to enter.

RO knowledge system understanding and procedure entry conditions.

Rev 0

Question 23 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98338 User-Defined ID: 98338

Reference:

OFN BB-006 23 RO which rad monitor confirms high reactor coolant Topic:

activity RO Importance Rating: 2.6 SRO Importance Rating: 3.0 K/A Number: 076 AK 2.01 Comments: BANK - 59116 Lesson Plan Objective: LO1732416 R1, IDENTIFY the procedure entry conditions.

Tier # 1 Group # 2 Last Used - 2007 Callaway Fundamental 55.41 part 11 KA - High reactor coolant activity - Knowledge of the interrelations between the high reactor coolant activity and the following - Process radiation monitors. Safety function 9

Modification History:

0 - feedback from Scott 4/25/15 Rev 0

24 ID: 98339 Points: 1.00 OFN RP-017, CONTROL ROOM EVACUATION, has been entered by the CRS due to a fire in the Control Room.

Which of the following is vital equipment that the RO will operate for this condition AND why?

A. Borate the RCS to ensure proper SDM exists.

B. Locally line up CCW to containment to ensure thermal barriers have cooling flow.

C. Ensure NK4421, BB PCV-456A PORV, breaker is CLOSED to allow for auto cycling to control RCS pressure.

D. OPEN NK4401, BUS NB02 BRKR CONTROL POWER, to prevent 'B' train equipment from loading on the bus due to possible hot shorts.

Answer: D Answer Explanation:

OFN RP-017 has the crew shutdown the plant and maintain it in a hot standby condition until subsequent actions can be taken to place the plant in the desired long term condition. OFN RP-017A is used to cooldown the plant if desired.

Correct - step C2 (immediate action step for the RO) has them open breakers to prevent control systems and buses from operating erratically due to hot shorts caused by the fire.

Incorrect - Borate the RCS to ensure proper SDM. This step will be performed in OFN RP-017A if the plant is to be cooled down and placed in cold shutdown.

Plausible if the student confuses which steps are performed in which procedure.

Incorrect - locally line up CCW to containment. This step will be performed in OFN RP-017A if the plant is to be cooled down and placed in cold shutdown.

Plausible if the student confuses which steps are performed in which procedure.

Incorrect - Close PORV breaker. This breaker is OPENED by the RO in the immediate actions steps to prevent them from cycling due to hot shorts.

Plausible if the student doesn't understand the intent of this procedure.

Meets KA asks what vital equipment will be operated during a plant fire.

RO knowledge these steps are in the immediate action section of the procedure.

Rev 0

Question 24 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98339 User-Defined ID: 98339

Reference:

OFN RP-017 REV 46 Topic: 24 RO vital equipment operated in case of fire RO Importance Rating: 3.3 SRO Importance Rating: 4.0 K/A Number: 067 AA 2.16 Comments: NEW Lesson Plan Objective: LO1732427, R2, RECOGNIZE the available situations which are addressed by procedure OFN RP-017.

Tier # 1 Group # 2 Last Used - N/A Memory 55.41 part KA - Plant fire on site - Ability to determine and interpret the following as they apply to the plant fire on site - Vital equipment and control systems to be maintained and operated during a fire - Safety function 8 Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

25 ID: 98340 Points: 1.00 During the performance of EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE, which of the following valves will be checked closed AND why?

A. KC HV-253, Fire Protection System Header Outside CTMT Isolation To ensure all non-essential containment penetrations are closed to prevent the release of radioactive materials.

B. EG HV-058, CCW To RCS Outer CTMT Isolation Valve To ensure all the containment penetrations are isolated since this could be the source of the high containment pressure.

C. EG HV-058, CCW To RCS Outer CTMT Isolation Valve To ensure all non-essential containment penetrations are closed to prevent the release of radioactive materials.

D. KC HV-253, Fire Protection System Header Outside CTMT Isolation To ensure all the containment penetrations are isolated since this could be the source of the high containment pressure.

Answer: A Answer Explanation:

Correct - per FR-Z1 basis for step 1 and 2 the reason for ensuring the valves are isolated is to prevent a radioactive release from a non-essential containment penetration. CISA closes the fire header valve (this valve is manually closed during normal ops but still receives a signal to close)

Incorrect - EG HV-058 and prevent the release of rad materials. Phase B will isolate the CCW system from containment not phase A. Plausible if the student confuses which signal closes which valves. The reason is correct.

Incorrect - EG HV-058 and pens closed for high pressure. Phase B will isolate the CCW system from containment not phase A. Plausible if the student confuses which signal closes which valves. The reason is plausible since water injected from any system could raise CTMT pressure.

Incorrect - KC HV-253 and pens closed for high pressure. Correct valve wrong reason. Plausible as any injection from any water source could cause CTMT pressure to rise.

Meets K/A asks for a loss of CTMT integrity what is checked for rad control, ability to control the release by ensuring valves are in correct position for event.

RO knowledge due to step basis of EMGs Rev 0

Question 25 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98340 User-Defined ID: 98340

Reference:

BD EMG FR-Z1 25 RO Basis for closing phase A valves on high Topic:

containment pressure RO Importance Rating: 3.8 SRO Importance Rating: 4.3 K/A Number: E14 2.3.11 Comments: NEW Lesson Plan Objective: LO1732350 R3, DISCUSS the major action steps for procedure EMG FR-Z1.

Tier # 1 Group # 2 Last Used - N/A Fundamental 55.41 part 9 KA - Loss of containment integrity (high pressure) - Rad control - Ability to control radiation release - Safety function 5

Modification History:

Rev 0

26 ID: 98341 Points: 1.00 The crew is performing EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION, with the following conditions:

RCS pressure is 585 psig down slow RCS temperature is 404°F down slow PZR level is 50% stable RCP A and D are running RCP B and C are stopped Both CCPs are running A SI pump is running B SI pump is tagged out for maintenance Both RHR pumps are stopped Containment pressure is 6.5 psig stable The crew is determining if CCPs and SI pumps should be stopped Using the attached reference, determine the CCP and SI pump configuration after the procedure steps are completed based on the given indications?

A. The B CCP will be running with the A SI pump running.

B. The A CCP will be running with NO SI pumps running.

C. The A CCP will be running with the A SI pump running.

D. The B CCP will be running with NO SI pumps running.

Answer: A Answer Explanation:

To get to this point in this procedure you will have had to go from E-0 to E-1 to ES-11. All prior procedures were checked to determine the appropriate indications given for this setup.

RCS subcooling - with pressure at 585 psig saturation temperature is 486°F so 486°F - 404°F = 82°F of subcooling In ES-11 steps 21 and 22 Correct - 'B' CCP and A SI pump. There are two CCPs running so the step will be performed. There are two RCP's running and one SI pump running and with adverse containment 70°F of subcooling is needed to stop the A CCP. There is a note stating that pumps should be stopped in alternate trains so you would not stop the B CCP with the B SI pump tagged out. The next step asks if any SI pumps are running which is yes, so with one CCP running and two RCP's running and one SI pump running 150°F of subcooling is needed. Subcooling is not met so step 22.c RNO directs the crew if RCS temperature is greater than 375°F then go to step 40 and not to stop the running SI pump.

The procedure step states pumps should be stopped in opposite trains. Without a reason for not stopping pumps in this manner they will be.

Rev 0

Incorrect - B CCP with NO SI pumps. There is a note that states pumps should be stopped in alternate trains so the B CCP will be the pump that is stopped. SI pump will not be stopped because subcooling is not met. Plausible if the student fails to use the adverse containment subcooling value and chooses to stop the pump.

Incorrect - A CCP with NO SI pumps. A CCP is correct. SI pump will not be stopped because subcooling is not met. Plausible if the student fails to use the adverse containment subcooling value and chooses to stop the pump.

Incorrect - B CCP with A SI pump running. There is a note that states pumps should be stopped in alternate trains so the B CCP will be the pump that is stopped. The SI pump is correct Meets K/A uses procedure steps to determine if operator can perform steps of LOCA cooldown RO knowledge procedure is given and steps are performed Rev 0

Question 26 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98341 User-Defined ID: 98341

Reference:

EMG ES-11 Topic: 26 RO Which ECCS pumps to stop in ES-11 RO Importance Rating: 3.9 SRO Importance Rating: 4.2 K/A Number: E03 2.1.25 Comments: Handout provided NEW Lesson Plan Objective: LO1732321 R3, DISCUSS the major action steps of procedure EMG ES-11.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 10 KA - LOCA cooldown and depressurization - Conduct of ops - Ability to interpret reference materials such as graphs curves and tables etc - Safety function 4

• Reference provided EMG ES-11 steps 21 and 22 Modification History:

Rev 0

27 ID: 98342 Points: 1.00 The unit has tripped due to a LOCA and ESF equipment has failed to start.

EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING, has been entered.

A depressurization of the Steam Generators (SGs) is being performed in accordance with EMG FR-C2, when the STA reports that there is a Red Path on the Integrity CSF Status Tree.

Which ONE (1) of the following describes the actions that will be taken?

A. Complete the S/G depressurization and then transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS, if the red path still exists B. Immediately transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS C. Complete FR-C2 and then transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS, if the red path still exists D. Stop the S/G depressurization and, if the red path does NOT clear, transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS Answer: C Answer Explanation:

Correct - Because of the SG depressurization, a Red condition on Integrity is expected. If this is not performed and a transition is made then the overall affect would end up being entry into C-1 inadequate core cooling and that is worse than staying in C-2. There is a NOTE in C2 telling the crew NOT to go to P1 if a red condition exists.

Incorrect - immediately transition to P-1. Plausible if the student misunderstands the direction in the note with respect to the direction in the procedure users guide about higher tear procedures.

Incorrect - stop the SG depressurization. Plausible since the action the operator did caused the red path so if they stop it and the red path clears then they might want to stay. Procedure use and adherence understanding.

Incorrect - complete the SG depressurization. Plausible since the FR procedure directed the action after the action then the student may think a transition now needs to be completed.

Meets KA asks about PTS with regards to cooldown and depressurization using SG at max rate RO knowledge EMG procedure note basis Rev 0

Question 27 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98342 User-Defined ID: 98342

Reference:

BD EMG FR-C2 27 RO FR-C2 note to not go to P-1 on red path until after Topic:

C-2 is complete RO Importance Rating: 3.6 SRO Importance Rating: 4.0 K/A Number: E08 EK 2.2 Comments: BANK - 59338 Lesson Plan Objective: LO1732341 R9, EXPLAIN the bases and knowledge requirements for selected procedure steps of EMG FR-C2.

Tier # 1 Group # 2 Last Used - 2007 Callaway exam Comprehension 55.41 part 10 KA - PTS - Knowledge of the interrelations between the PTS and the following - Facility heat removal systems including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility -

Safety function 4 Modification History:

Rev 0

28 ID: 98343 Points: 1.00 The plant is operating at 45% power, steady state when 'A' RCP experiences an Under-Frequency trip due to an electrical fault. The PA buses continue to operate as normal.

Which of the following describes the 'A' loop RCS temperatures and 'A' S/G response from time 0 to 1 minute after the RCP trip?

(assume NO operator actions)

A. Initially to several seconds later there is little change in Thot and Tcold, after that out to 1 minute the difference is lowering rapidly.

S/G pressure is higher than pre-trip and steam flow is higher.

B. Initially to several seconds later the difference between Thot and Tcold is changing rapidly, after that out to 1 minute the change is smaller.

S/G pressure is higher than pre-trip and steam flow is higher.

C. Initially to several seconds later there is little change in Thot and Tcold, after that out to 1 minute the difference is lowering rapidly.

S/G pressure is the same as pre-trip and steam flow is close to 0.

D. Initially to several seconds later the difference between Thot and Tcold is changing rapidly, after that out to 1 minute the change is smaller.

S/G pressure is the same as pre-trip and steam flow is close to 0.

Answer: C Answer Explanation:

Ran this on the desktop simulator to gather information from. From the time the RCP trip was inserted the RCS loop temperatures slowly converged toward each other. After the flywheel lost forward inertia the temperature difference rapidly converged and then Thot ended up lower than Tcold because of reverse flow in the loop and the SG still removing some heat (not much). The SG pressures were equal to the other SGs since they are all still connected together but the steam flow from the affected loop went down to near 0. Flow from the other pumps goes up due to the loss of the flow from the affected loop.

Correct - The response of Th and Tc is due to the design coast down of the RCPs which lasts approximately 1-1.5 minutes (30 seconds per TS). Although both temperatures will be lower, Th will lower faster than Tc due to the sudden, significant reduction in heat generated by the reactor going to that SG. SG pressure will be relatively stable Incorrect - time 0 little change then rapid change and SG pressure higher. The first part is correct. The affected SG is now not producing any power for the turbine so the other SG must make up the loss of steam flow from this loop so overall pressure is lower not higher. Plausible if the student thinks that the loss of steam flow is because the pressure is higher.

Incorrect - time 0 rapid change then Th and Tc equal and SG pressure same.

The SG pressure is correct. Plausible if the student forgets about the flywheel effect on the RCS indications.

Rev 0

Incorrect - time 0 rapid change then Th and Tc equal and SG pressure higher.

Plausible if the student forgets about the flywheels effect on the RCS indications.

Also the affected SG is now not producing any power for the turbine so the other SG must make up the loss of steam flow from this loop so overall pressure is lower not higher. Plausible if the student thinks that the loss of steam flow is because the pressure is higher.

Meets the KA because it asks for RCS parameters after a loss of an RCP (coastdown with the flywheel)

RO knowledge since this is system understanding of how the flywheel on the RCP will affect the other RCS indications after the loss of the pump Question 28 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98343 User-Defined ID: 98343

Reference:

SY1300300 28 RO RCS temperature response to loss of RCP's during Topic:

the first minute RO Importance Rating: 2.8 SRO Importance Rating: 3.2 K/A Number: 003 K 5.02 Comments: Modified - Millstone Lesson Plan Objective: SY1300300 R9, DETERMINE the operational implications that a loss of flow will have on the core operating parameters.

Tier # 2 Group # 1 Last Used - Millstone 2011 Comprehension 55.41 part 2, 3 KA - RCP - Knowledge of the operational implications of the following concepts as they apply to the RCPs - Effects of RCP coastdown on RCS parameters Modification History:

0 - changed based on feedback from Scott 4/25/15 Rev 0

29 ID: 98344 Points: 1.00 Given the following:

The plant is recovering from a forced outage Core is at middle of life The crew is performing a reactor startup per GEN 00-003 Reactor power is 1 X 10-8 amps and stable Auto makeup to the VCT occurs During the auto makeup, BG FK-110, BA FLOW CTRL, was inadvertently set at 0.0 turns With NO operator action which of the following describes the reactor power response to this event?

A. Reactor power will continue to rise until trip on IR high flux.

B. Reactor power will continue to rise too slightly above the POAH.

C. Reactor power will continue to lower until subcriticality is reached.

D. Reactor power will remain constant because charging is flowing to RCP seals ONLY at this point.

Answer: B Answer Explanation:

Correct - At this power level there is no feedback from temperature of the RCS.

As the dilution continues power will rise until the POAH is reached at then power will level off and stay there.

Incorrect - no plant response since charging to seals only. This is a possible lineup for the RCP seals and charging. Plausible if the student mistakes the current normal plant lineup.

Incorrect - lower power until subcritical. This would be true if the setting on the pot was higher than the current RCS boron concentration. Plausible if the student forgets which way the controller works.

Incorrect - trip on IR high flux. If power were to rise above the POAH and did not turn this would happen next. Plausible if the student mistakes the dilution and no operator action with the low power trips and forgets the POAH effects of temperature.

Meets the KA by asking the operational implications of a dilution from the CVCS system. This has a part about understanding that the pot has an effect on this (dilution) but if the stem states a dilution is in progress then the question will be a LOD1 RO knowledge based on system response for a dilution.

Rev 0

Question 29 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98344 User-Defined ID: 98344

Reference:

LO1130641 Topic: 29 RO dilution effect on the reactor without operator action RO Importance Rating: 3.6 SRO Importance Rating: 3.7 K/A Number: 004 K 5.20 Comments: MODIFIED - PI Lesson Plan Objective: LO1130641 R12, Discuss the concept of the point of adding heat (POAH) and its relationship to reactor power.

SY1300400 R17 Tier # 2 Group # 1 Last Used - 2010 Prairie Island Comprehension 55.41 part 1 KA - CVCS - Knowledge of the operational implications of the following concepts as they apply to the CVCS -

Reactivity effects of xenon, boration, and dilution Modification History:

0 - feedback from Scott 4/25/15 Rev 0

30 ID: 98345 Points: 1.00 The plant is operating at 50% power and stable. An electrical failure in controller BG PK-131, LETDOWN HEAT EXCHANGER OUTLET PRESSURE CONTROL, has caused the controller to fail to 0% output.

Which of the following lists the changes in the indications for this failure?

LTDN HX OUTLET LTDN HX OUTLET FLOW PRESS A. DOWN UP B. UP UP C. UP DOWN D. DOWN DOWN Answer: A Answer Explanation:

Correct- With a failure of the BG PK-131 controller to 0% demand this closes the valve fully. This will cause the letdown relief valve inside containment to lift, causing all letdown flow to be directed to the PRT. This will cause pressure to go up and flow back to the VCT (outlet flow) to go down.

Incorrect - Down, Down. Plausible if the student doesn't understand how the system works under failure modes.

Incorrect - Up, Up. Plausible if the student doesn't understand how the system works under failure modes.

Incorrect - Up, Down. This is exactly backwards if the student thinks that 0% is with the valve open. Plausible if the student confuses what 0% means for the valve position.

Meets the KA asks for ability to monitor auto ops of CVCS during letdown RO knowledge since this asks for system understanding Rev 0

Question 30 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98345 User-Defined ID: 98345

Reference:

SY1300400 Topic: 30 RO CVCS indications with BG PF-131 failure closed RO Importance Rating: 3.6 SRO Importance Rating: 3.4 K/A Number: 004 A 3.11 Comments: NEW Lesson Plan Objective: SY1300400 R24, Integrate system and plant response to transient and equipment failures, including interactions with related systems Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - CVCS - Ability to monitor auto operation of the CVCS including - Charging and letdown Modification History:

Rev 0

31 ID: 98346 Points: 1.00 Which of the following busses directly supply power to RHR Pumps 'A' and 'B'?

A RHR Pump B RHR Pump A. NN03 NN04 B. NG01 NG02 C. NB01 NB02 D. PB03 PB04 Answer: C Answer Explanation:

Obj 2 SY1300500 Correct - NB01 and NB02 is CORRECT. The electrical power supplies for the RHR pumps are as follows:

NB01 RHR pump 'A', (Breaker NB0101).

NB02 RHR pump 'B', (Breaker NB0204).

Incorrect - PB03 and PB04. Plausible because PB buses power the condensate, heater drain, and NCP pumps Incorrect - NG03 and NG04. Plausible because RHR valves are powered from NG01 and NG02 (e.g. EJ HV-8809A from NG01B and EJ HV-8809B from NG02B)

Incorrect - NN03 and NN04. Plausible because RHR instrumentation is powered from NN.

Meets the KA by asking the power supply to the RHR pump RO knowledge based on power supply understanding Rev 0

Question 31 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98346 User-Defined ID: 98346

Reference:

KD-7496 ONE LINE POWER DISTRIB Topic: 31 RO RHR power supplies RO Importance Rating: 3.0 SRO Importance Rating: 3.2 K/A Number: 005 K 2.01 Comments: BANK - 59457 Lesson Plan Objective: SY1300500 R2, Explain the function, location, operation, and electrical interlocks of the major components.

Tier # 2 Group # 1 Last Used - Wolf Creek 2006 Fundamental 55.41 part 8 KA - RHR - Knowledge of bus power supplies to the following - RHR pump Modification History:

Rev 0

32 ID: 98347 Points: 1.00 The unit is at 100% power with the following conditions given for the 'B' Accumulator:

Boron concentration sample 2305 ppm Pressure 485 psig Borated Water Volume 6294 gal Outlet isolation valve OPEN breaker OPEN If a LBLOCA were to now occur on the 'A' loop hot leg what would be the effect (if any) on the RCS?

A. None, the accumulator would inject as design.

B. The low boron concentration could allow for a reactor re-start.

C. N2 injection could impede natural circulation flow for the 'B' loop.

D. Fuel clad could overheat due to NOT enough water injection from the accumulator.

Answer: D Answer Explanation:

Correct - With the low N2 pressure of the accumulator the full volume of water would not inject and the purpose of the accumulator is to flood the core on a LBLOCA until the ECCS pumps have started and begin to inject water to continue to remove heat.

Incorrect - None. Plausible if the student confuses the TS requirements for the accumulator and believes all spec are made which would indicate that it would perform its design function and fully inject Incorrect - Low boron. Plausible if the student confuses the TS requirements for the accumulator and believes the boron is low and the reason for the boron is to prevent a restart.

Incorrect - N2 injection. Plausible if the student confuses the TS requirements for the accumulator and believes this will cause N2 to inject which would impede NC.

Meets the KA asks for effect on the ECCS injection (blowdown phase) of an accumulator that is not within TS (malfunctioned)

RO knowledge system understanding Rev 0

Question 32 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98347 User-Defined ID: 98347

Reference:

TS 3.5.1 Topic: 32 RO accumulator malfunction due to N2 out of TS low RO Importance Rating: 3.4 SRO Importance Rating: 3.9 K/A Number: 006 K 6.02 Comments: NEW Lesson Plan Objective:

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 8 KA - ECCS - Knowledge of the effect of a loss or malfunction on the following will have on the ECCS - Core flood tanks accumulators Modification History:

Replaced 8/25/15 based on Scotts feedback of missing the KA Rev 0

33 ID: 98348 Points: 1.00 The Pressurizer Relief Tank (PRT) can be drained, using no pumps, to the:

A. Waste Holdup Tank B. Recycle Hold Up Tank (RHUT)

C. Instrument Tunnel Sump D. Containment Normal Sump Answer: D Answer Explanation:

M-12BB02 SYS BB-202 Correct - The PRT has a drain line that will drain the tank directly to the containment normal sump Incorrect - RHUT. This is a normal place to put contaminated drains. Plausible if the student forgets the pumps are required to drain the PRT to this location.

Incorrect - Waste Holdup Tank. The path takes water from the PRT through the RCDT pumps then to the waste tank. Plausible if the student forgets the pumps are required to drain the PRT.

Incorrect - Instrument Tunnel sump. Since this is inside containment this is a possible solution but no drains off of the PRT will take water to this location.

Plausible if the student links the PRT location and the reactor cavity sump location together inside containment.

Meets KA by knowledge of physical connection between PRT and containment RO knowledge bases on understanding the system interconnections Rev 0

Question 33 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98348 User-Defined ID: 98348

Reference:

M-12HB01 Topic: 33 RO Where can the PRT be drained to RO Importance Rating: 2.9 SRO Importance Rating: 3.1 K/A Number: 007 K 1.01 Comments: BANK - Indian Point Lesson Plan Objective: SY1300200 R2, EXPLAIN the function and operation of the major RCS components.

Tier # 2 Group # 1 Last Used - Indian Point 2010 Fundamental 55.41 part 3 KA - PRT - Knowledge of the physical connections and or cause effect relationships between the PRT and the following systems - Containment system Modification History:

Rev 0

34 ID: 98349 Points: 1.00 The unit is operating at 100% power when the following takes place:

061B, PROCESS RAD HI, comes in alarm 061A, PROCESS RAD HIHI, comes in alarm shortly after 061B

'B' CCW surge tank vent valve indicates closed

'B' CCW surge tank level is up slow Based on the given indications:

1) Where is the leak coming from?

2) What actions will the crew perform per procedure?

A. 1. Letdown Heat Exchanger

2. Bypass the heat exchanger B. 1. Seal Water Heat Exchanger

2. Bypass the heat exchanger C. 1. Seal Water Heat Exchanger

2. Isolate seal water D. 1. Letdown Heat Exchanger

2. Isolate letdown Answer: D Answer Explanation:

The given conditions are indicative of a leak into the CCW system and since the vent valve has closed the water leaking in is radioactive. The surge tank level rising with no makeup also gives indication that water is leaking into the CCW system. Other indications do change but are not required to diagnose this leak location. Of the two locations given in the distractors for the leak only one is higher pressure than CCW. The seal water heat exchanger is lower pressure so the leak would be lowering surge tank level. Since the water in the seal water heat exchanger is radioactive and the procedure does have the crew bypass this component if the leak is suspected there the distractors are very plausible if the student confuses which pressure is higher.

Correct - Per procedure if the leak is not in the seal water heat exchanger the crew is directed to isolate the leak and return to procedure in effect.

Incorrect - Letdown, bypass. The location is correct but there is no procedure direction to bypass this heat exchanger only to isolate it. Plausible if the student confuses the letdown and seal water heat exchanger procedure directions.

Incorrect - Seal water, isolate. This is the wrong location since this heat exchanger is lower pressure than CCW. The procedure has this heat exchanger bypass and not isolated. Plausible if the student confuses which is at higher pressure and which gets bypassed vs isolated.

Rev 0

Incorrect - Seal water, bypass. This is the wrong location but the correct action for a leak in this heat exchanger. Plausible if the student confuses which heat exchanger is higher pressure since the procedure will direct bypassing the seal water one.

Meets the KA ability to predict how a rad alarm on the CCW system will affect operations and use procedures to address RO knowledge based on system understanding.

Question 34 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98349 User-Defined ID: 98349

Reference:

OFN EG-004 Topic: 34 RO Leak in letdown heat exchanger procedure actions RO Importance Rating: 3.3 SRO Importance Rating: 3.5 K/A Number: 008 A 2.04 Comments: NEW Lesson Plan Objective: SY1400800 R9, DETERMINE the protection afforded by the design of the CCW System for different portions of the system.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 5, 7 KA - Component Cooling Water - Ability to predict the impacts of the following malfunctions or operations on the CCWS and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - PRMS alarm Modification History:

Rev 0

35 ID: 98350 Points: 1.00 Reactor power is at 60% and stable awaiting repairs to the 'B' MFP. The RO notices the following:

• BB ZL-455B, PZR SPRAY LOOP 1 CTRL VLV, both green and red lights lit with controller demand output at 35%

• BB ZL-455C, PZR SPRAY LOOP 2 CTRL VLV, green light lit red light out with controller demand output at 0%

• BB PK-455A, PZR PRESS MASTER CTRL, demand indicates 40% and down slow

• ALR 033C, PZR PRESS LO HTRS ON, is lit Which of the following procedures AND actions will mitigate this event?

A. Enter ALR 033C, PZR PRESS LO HTRS ON, and energize PZR backup heaters.

B. Enter ALR 033C, PZR PRESS LO HTRS ON, trip the reactor and stop 'A' and

'D' RCP's.

C. Enter OFN SB-008, INSTRUMENT MALFUNCTIONS, take manual control of BB PK-455B and CLOSE the valve.

D. Enter OFN SB-008, INSTRUMENT MALFUNCTIONS, and select an alternate PZR pressure channel for control.

Answer: C Answer Explanation:

Correct - with 35% demand and both green and red lights lit on one spray valve RCS pressure will be lowering. OFN SB-008 Att V memory action steps have the RO take manual control of either the spray valve or the PZR pressure master controller and control spray flow, in this case close the spray valve is the only choice since the master controller is at 40% and down slow and the other spray valve is closed. The master pressure controller will not correct this failure.

Incorrect - ALR 033C and trip reactor and stop A and D RCP's. This ALR would be correct for this issue but not to trip the reactor and stop RCP's. Plausible if the student feels no actions taken will stop the pressure from lowering to less than the reactor trip setpoint and stopping RCP's will stop the pressure from dropping any lower.

Incorrect - ALR 033C and energize PZR heaters. This ALR would be correct for this issue but energizing heaters will only at most delay a reactor trip on low pressure if more action is not taken. Plausible if the student feels that the heaters alone will mitigate the pressure lowering.

Rev 0

Incorrect - OFN SB-008 and select alternate pressure channel. Correct procedure but with the indications given a pressure channel has not failed.

Plausible if the student feels that selecting out a failed pressure channel will stop the pressure from lowering any more.

Meets KA by asking how to respond to a PZR malfunction using the procedure RO knowledge since this is a memory action step of the procedure but high cog to interpret what the lights mean Question 35 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98350 User-Defined ID: 98350

Reference:

OFN SB-008 35 RO PZR spray open which procedure to enter and Topic:

action to take RO Importance Rating: 4.3 SRO Importance Rating: 4.4 K/A Number: 010 2.1.23 Comments: MODIFIED - 58685 Lesson Plan Objective: LO1732418 R4, EXPLAIN the plant response for each instrument failure identified in procedure OFN SB-008.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 5, 7, 10 KA - PZR pressure control - Conduct of ops - Ability to perform specific system and integrated plant procedures during all modes of plant operation Modification History:

Rev 0

Q36, Rev 1 ID: 98351 Points: 1.00 The unit has experienced a complete loss of bus NN02. 076A - SSPS B GENERAL WARNING alarm is received.

Which of the following describes why this alarm is received?

A. Due to a loss of a DC power supply on one train of RPS.

B. Due to a loss of an output relay cabinet in RPS.

C. Due to a loss of input relays on one train of RPS.

D. Due to the loss of the ability to trip the 'B' reactor trip breaker.

Answer: A Answer Explanation:

Correct - NN02 supplies power to RPS in three places. Two are the input relays for the white train of logic. The other one is 15/48 VDC logic cabinet power supply to B train only.

Incorrect - output relay cabinet. The output relay cabinets are powered from NN01 and NN04.

Incorrect - input relay on one train. NN02 powers up both trains white channel inputs not just one.

Incorrect - loss of ability to trip the B trip breaker. General warning alarm is associated with the rod control system and trip breakers are the source of power for the rod control system. Plausible if the student puts this together wrong and believes this will prevent the trip breaker opening from the control room.

Meets KA by asking what RPS power supply is lost RO knowledge of power supplies to RPS. High cog because of integrating the NN02 loss and the SSPS B general warning alarm.

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q36, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98351 User-Defined ID: 98351

Reference:

SY1301200 Topic: 36 RO RPS power supply SSPS GW from loss of NN02 RO Importance Rating: 3.3 SRO Importance Rating: 3.7 K/A Number: 012 K 2.01 Comments: NEW Lesson Plan Objective: SY1301200 R5, Explain the RPS operation.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - Reactor protection - knowledge of bus power supplies to the following - RPS channels components and interconnections Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

37 ID: 98352 Points: 1.00 The crew is responding to a reactor trip when it becomes necessary to manually actuate SI. 50 seconds after the crew actuates SI the CRS has the RO reset the SI signal.

Will the SI signal reset AND what effect (if any) will this action have on related plant equipment?

A. YES, H2 fans will need to be manually started.

B. NO, H2 fans will need to be manually started.

C. YES, there will be NO effect on plant equipment.

D. NO, there will be NO effect on plant equipment.

Answer: D Answer Explanation:

M-744-00025 Correct - SI signals once started require a 60 second timer to complete before any other actions with the related equipment can be completed. This also locks out any reset of the SI prior to this timer completing.

Incorrect - No, H2 fan need manual start. The SI signal part is correct you cannot reset it until the timer is complete. H2 fans will start at 60 seconds since there is no input from the SI signal except to start an independent timer that will restart these fans after 60 seconds regardless of any other action taken by the crew.

Incorrect - Yes, H2 fan need manual start. The SI signal will block all reset attempts until its 60 second timer completes. H2 fans will start at 60 seconds since there is no input from the SI signal except to start an independent timer that will restart these fans after 60 seconds regardless of any other action taken by the crew.

Incorrect - Yes, on effect to plant equipment. The SI signal will block all reset attempts until its 60 second timer is complete. Second part is correct plant equipment will continue to operate as design.

Meets the KA by understanding the physical connection of ESFAS and reset RO knowledge system understanding Rev 0

Question 37 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98352 User-Defined ID: 98352

Reference:

M-744-00025 Topic: 37 RO Reset SI before 60 seconds RO Importance Rating: 3.7 SRO Importance Rating: 4.1 K/A Number: 013 K 1.18 Comments: NEW Lesson Plan Objective: SY1301301 R4, Describe Operation Of The Engineered Safety Features Actuation System; Including Automatic Actuation, And Bypass And Reset Operation.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - ESFAS - Knowledge of the physical connections and or cause effect relationship between the ESFAS and the following systems - Premature reset of ESF actuation Modification History:

Rev 0

38 ID: 98353 Points: 1.00 The plant has experienced a LOCA. SI, CISB, and CSAS have all auto actuated as required.

Which ONE of the following Annunciators would direct the Control Room staff to transfer the Containment Spray Pump Suctions to the Recirc Sump?

A. ALR 047B, RWST EMPTY B. ALR 047C, RWST LEV LO-LO 2 C. ALR 047E, RWST LEV HI/LO D. ALR 047D, RWST LEV LO-LO 1 AUTO XFR Answer: B Answer Explanation:

ALR 00-047C, EMG ES-12 Correct - per ES-12 and the ALR 047C if the RWST level is 12% or lower then spray pumps are placed on recirc.

Incorrect - RWST EMPTY. This alarm has the crew stop all pumps taking a suction from the RWST but the tank is not empty it still has 6% in it of usable volume per the ALR. Plausible if the student forgets which alarm provides for which action.

Incorrect - RWST LEV HI/LO. This alarm is if the RWST is lower than the TS value to alert the crew to refill it and also if it is to hi overflowing it will come in.

Plausible if the student forgets which alarm provides for which action.

Incorrect - RWST LO-LO 1 AUTO XFR. This alarm has the crew perform ES-12 to transfer all the ECCS equipment over to the recirc sump. It doesn't have the crew swap spray at this time. Plausible if the student forgets which alarm provides for which action.

Meets the KA asks for ability to monitor changes in RWST level for ESFAS changes RO knowledge this is a setpoint of an alarm Rev 0

Question 38 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98353 User-Defined ID: 98353

Reference:

EMG ES-12 Topic: 38 RO RWST LO LO 2 realignment of spray pumps RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: 013 A 1.06 Comments: BANK - 46455 Lesson Plan Objective: LO1732322 R3, DISCUSS the major action steps of procedure EMG ES-12.

Tier # 2 Group # 1 Last Used - 2009 Callaway Fundamental 55.41 part 7, 10 KA - ESFAS - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the ESFAS controls including - RWST level Modification History:

Rev 0

39 ID: 98354 Points: 1.00 The unit is operating at 100%. The 'B' Cavity Cooling fan was running but its output breaker trips and will NOT reset.

Which of the following is the correct response to this event AND why?

The 'A' Cavity Cooling fan...

A. auto starts to provide cooling to the incore instrumentation vessel connections.

B. must be manually started to provide cooling to the incore instrumentation vessel connections.

C. must be manually started to provide cooling to the excore NIs to prevent possible damage due to high temperatures.

D. auto starts to provide cooling to the excore NIs to prevent possible damage due to high temperatures.

Answer: D Answer Explanation:

SY1302600 Correct - a loss of power to the running fan will auto start the other fan. One of the functions of these fans is to cool the excore NI detectors to a max of 135F.

This allows them to operate properly and give reliable indications. As stated in the NI LP if temperature gets too high for the detectors they will not provide accurate indication.

Incorrect - Auto start and cool the instrument vessel connections. The first part is correct it will auto start. The instrument vessel connections are on top of the head and not cooled by the cavity cooling fans. Plausible if the student confuses where the fans discharge to.

Incorrect - manual start and cools excore NIs. The fans will auto start on a loss of power, that is all that will auto start them. A failure of the running fan will not start its counterpart. The second part is correct. Plausible if the student remembers that the fans don't start on a malfunction of the fan but forgets that loss of power is not a malfunction of the fan.

Incorrect - manual start and cools instrument vessel connections. The fans will auto start on a loss of power, that is all that will auto start them. A failure of the running fan will not start its counterpart. The instrument vessel connections are on top of the head and not cooled by the cavity cooling fans. Plausible if the student forgets what starts the fans and where they discharge to.

Meets the KA by asking what the RO monitors from the control room, cooling fan, and what is the expected response to the plant, the other should auto start.

RO knowledge based on system understanding Rev 0

Question 39 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98354 User-Defined ID: 98354

Reference:

SY 1302600 Topic: 39 RO cavity cooling fan loss of power and function RO Importance Rating: 3.2 SRO Importance Rating: 3.1 K/A Number: 022 A 4.02 Comments: NEW Lesson Plan Objective: SY1302600 R11, DESCRIBE the Containment Cooling System design feature(s) and system interlocks.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 7 KA - Containment cooling system - Ability to manually operate and or monitor in the control room: - CCS pumps Modification History:

Rev 0

40 ID: 98355 Points: 1.00 Review the attached test data sheet from STS EN-100A, CONTAINMENT SPRAY PUMP A INSERVICE PUMP TEST.

Which of the following statements is correct with regard to the information contained on the test data sheet?

The RO should notify the SM/CRS that the 'A' Containment Spray Pump...

A. has failed the surveillance due to pump Dp being too low.

B. static suction pressure is UNSAT but NO actions are required at this time.

C. valve ENV0099 leak rate is UNSAT and a CR should be initiated ONLY.

D. has failed the surveillance due to dynamic suction pressure being too high.

Answer: C Answer Explanation:

AP 29B-003, SURVEILLANCE TESTING, is the admin procedure that governs performing surveillances.

Per the surveillance test data sheet attached there are only one item that are not within the normal range of acceptance. ENV0099 leak rate being too high. Per required action 2 this is only an UNSAT item and the pump is considered OPERABLE and a CR should be initiated ONLY.

Correct - ENV0099 leak rate to high and the note states that only a CR needs to be initiated.

Incorrect - pre suction pressure unsat. Plausible if the student mis reads or gets confused on the table.

Incorrect - failed due to pump Dp. Plausible if the student mis reads or gets confused on the table.

Incorrect - failed due to post suction pressure to high. Plausible if the student mis reads or gets confused on the table.

Meets KA because the surveillance is over the 026 Containment Spray system and is determining that the surveillance for this equipment is or is not being met (knowledge of surveillance procedures).

RO knowledge simple procedure usage.

Rev 0

Question 40 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 98355 User-Defined ID: 98355

Reference:

AP 29B-003 Topic: 40 RO containment spray surveillance test data pass/fail RO Importance Rating: 3.7 SRO Importance Rating: 4.1 K/A Number: 026 2.2.12 Comments: Handout provided NEW Lesson Plan Objective: LO1733214 R8, DEMONSTRATE proper application of the Definitions, Responsibilities and Procedural Requirements associated with AP29B-003, Surveillance Testing.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Containment spray - Equipment control - Knowledge of surveillance procedures Modification History:

Rev 0

41 ID: 98356 Points: 1.00 An event has occurred in the plant and 'B' S/G pressure has risen to 1217 psig.

The total number of safety valves that will be open at this pressure is...

(ignore the effects of blowdown/accumulation)

A. 2 B. 3 C. 4 D. 5 Answer: B Answer Explanation:

SG safety valves open at 1185, 1197, 1210, 1222, 1234 psig respectively.

Correct - with current pressure at 1217 psig there should only be 3 safety valves open Incorrect - 2. Plausible if the student confuses setpoints or only recalls one and assumes the others.

Incorrect - 4. Plausible if the student confuses setpoints or only recalls one and assumes the others.

Incorrect - 5. Plausible if the student confuses setpoints or only recalls one and assumes the others.

Meets KA asks to monitor pressure to ensure main steam system design limits are not exceeded RO knowledge system design and setpoints Rev 0

Question 41 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 100318 User-Defined ID: 98356

Reference:

SY1503900 Topic: 41 RO main steam safety valve setpoints RO Importance Rating: 3.0 SRO Importance Rating: 3.1 K/A Number: 039 A 1.06 Comments: NEW Lesson Plan Objective: SY1503900 R4, Discuss the operation of the Steam Generator Atmospheric Relief Valves.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 7 KA - Main and reheat steam system - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the main and reheat steam system controls including - Main steam pressure Modification History:

Rev 0

42 ID: 98357 Points: 1.00 What design features protect the containment spray nozzles from plugging during recirculation phase?

Containment Sump Strainers Debris barriers at bio shield wall Self cleaning strainer between the RWST discharge and spray pump suction A. 1, 2, & 3 B. 1&3 C. 1&2 D. 2&3 Answer: C Answer Explanation:

Correct - strainers are part of the filtering process but the installed barriers from the loops also stop this debris.

Incorrect - 1 & 3. First part is correct, second part is a device that is used on systems at Wolf Creek to filter the water and then self clean. Plausible if the student confuses where the installed self cleaning strainers are located.

Incorrect - 2 & 3. First part is correct, second part is a device that is used on systems at Wolf Creek to filter the water and then self clean. Plausible if the student confuses where the installed self cleaning strainers are located.

Incorrect - 1, 2, & 3. The first two are correct, last part is a device that is used on systems at Wolf Creek to filter the water and then self clean. Plausible if the student confuses where the installed self cleaning strainers are located.

Meets KA asks how the nozzles are protected from debris during recirc RO knowledge system design and components Rev 0

Question 42 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98357 User-Defined ID: 98357

Reference:

SY1302600 Topic: 42 RO CTMT sump strainer function RO Importance Rating: 2.8 SRO Importance Rating: 3.2 K/A Number: 026 K 4.05 Comments: NEW Lesson Plan Objective: SY1302600 R3, Describe the function of major Containment Spray System components and controls.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 8 KA - Containment Spray System - Knowledge of CSS design feature and or interlocks which provide for the following: - Prevention of material from clogging nozzles during recirculation Modification History:

Rev 0

43 ID: 98358 Points: 1.00 With the plant operating at full power which of the following failures of AE PT-508, Main Feed Header Pressure Channel, would cause an INITIAL RISE in feedwater flow to all S/G's AND what procedural action will the crew take to mitigate the failure?

A. Fails LOW. Take manual control of Main Feedwater Regulating valves.

B. Fails LOW. Take manual control of Main Feedwater pump speed.

C. Fails HIGH. Take manual control of Main Feedwater pump speed.

D. Fails HIGH. Take manual control of Main Feedwater Regulating valves.

Answer: B Answer Explanation:

Correct - this is an input to the feed pump speed control circuit so when this fails low the MFP will speed up causing an initial SG level rise. To fix this per OFN SB-008 take manual control of the MFP speed controller.

Incorrect - AE PT-508 fails high take manual control of main feed reg valves.

This failure will cause a lowering of main feed pump to restore program differential pressure for the main feed pump. The OFN will have the BOP take manual control of the pump not each feed reg valve. Plausible if the student confuses what inputs to the main feed pump speed and what affect it would have on the system as well as this action is required for other failures within the feed system.

Incorrect - AE PT-508 fails high take manual control of main feed pump speed.

Correct procedure action to take but the failure will cause a lowering of main feed pump to restore program differential pressure for the main feed pump. Plausible if the student confuses what inputs to the main feed pump speed and what affect it would have on the system.

Incorrect - AE PT-508 fails low take manual control of main feed reg valves.

Correct failure to cause the initial rise in main feed pump speed to raise flow to all SG but the OFN will have the BOP take manual control of the pump not each feed reg valve. Plausible since this is an action for different failures within the main feed pump and feed reg valves.

Meets KA because asks what is the malfunction that would cause an effect to SG from MFW RO knowledge since it is system knowledge Rev 0

Question 43 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98358 User-Defined ID: 98358

Reference:

SY1505900 Topic: 43 RO SG level effects of AE PT 508 failure RO Importance Rating: 3.5 SRO Importance Rating: 3.7 K/A Number: 059 K 3.03 Comments: MODIFIED - 58915 Lesson Plan Objective: SY1505900 R11, Discuss the instrumentation and controls of the Feedwater System, including trips and automatic actions of a Feedwater Isolation Signal (FWIS).

Tier # 2 Group # 1 Last Used - 2007 Callaway Comprehension 55.41 part 5 KA - Main feedwater - Knowledge of the effect that a loss or malfunction of the MFW will have on the following - S/G Modification History:

Rev 0

44 ID: 98359 Points: 1.00 Which ONE of the following is the correct power supply to AL HV-34, CST to MD AFP B?

A. NG01 B. NG02 C. NG03 D. NG04 Answer: D Answer Explanation:

E-13AL02B MDAFWP's don't have MOVs for discharge valves Correct - per the electrical print NG04 Incorrect - NG01. There are only four safety related power sources for this valve and this is not it. Plausible if the student doesn't know the power supply.

Incorrect - NG02. There are only four safety related power sources for this valve and this is not it. Plausible if the student doesn't know the power supply.

Incorrect - NG03. There are only four safety related power sources for this valve and this is not it. Plausible if the student doesn't know the power supply.

Meets KA because it asks for the power supply to the AFW MOVs RO knowledge system understanding Rev 0

Question 44 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98359 User-Defined ID: 98359

Reference:

E-13AL02C Topic: 44 RO Power supply to AL HV-34 RO Importance Rating: 3.2 SRO Importance Rating: 3.3 K/A Number: 061 K 2.01 Comments: MODIFIED - 59879 Lesson Plan Objective: SY1406100 R3, Explain the operation of the MDAFWP Discharge Valves.

Tier # 2 Group # 1 Last Used - 2009 Callaway Audit Memory 55.41 part 7 KA - AFW - Knowledge of bus power supplies to the following - AFW system MOV's Rev 0

45 ID: 98360 Points: 1.00 Given the following with the unit at 100% power:

Reactor trip

'B' MDAFWP fails to start Based on the information given what is the status of the following MDAFW valves 1 minute later?

X=CLOSED O=OPEN A S/G MD B S/G MD C S/G MD D S/G MD A. X O O X B. X X X X C. O X X X D. O X X O Answer: C Answer Explanation:

Validated on the desktop simulator With the reactor trip from 100% AFW will start on SG low level. The TDAFWP will start and flow to all SGs and discharge valves will not throttle they will stay full open. The B MDAFWP starting allows the signal to be sent to the smart valves to arm and throttle flow. The D SG MD valve works different. When the B MD pump trips the D SG valve will throttle closed base on flow from the TD pump but the A SG valve will stay open since the B MD pump is not running. For the smart valves to work the MD pump must be running but D SG is different.

Correct - With a start or not start of the B MD pump the D SG smart valve will still throttle closed due to flow from the TDAFW pump. But the A SG valve will not close because the B MD pump is not running which is required to make this valve work as design Incorrect - all closed. This is the normal response if no failure happens. All smart valve close based on flow. Plausible if the confuses the fact that since flow is going to all SG that all smart valves should function as normal.

Incorrect - A and D open, B and C closed. This is what would happen if the A MD pump failed to start. Plausible if the student confuses what pump feeds what SG. Correct is A - BC and B - AD Incorrect - A and D closed, B and C open. Plausible if the student understands how the smart valves work but forgets the D SG valve is wired different.

Meets KA asks for effect on AFW if pumps malfunction Rev 0

RO knowledge system understanding Question 45 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98360 User-Defined ID: 98360

Reference:

SY1406100 Topic: 45 RO AFW discharge valve positions with B pump failure RO Importance Rating: 2.6 SRO Importance Rating: 2.7 K/A Number: 061 K 6.02 Comments: NEW Lesson Plan Objective: SY1406100 R3, Explain the operation of the MDAFWP Discharge Valves.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7, 8 KA - AFW - Knowledge of the effect of a loss or malfunction of the following will have on the AFW components - Pumps Modification History:

Rev 0

46 ID: 98361 Points: 1.00 Which of the following is an acceptable power lineup per T.S. to an NN bus with the plant in MODE 4?

A. NG04 NK26 NK04 NN16 NN04 B. PG20 NK26 NK02 NN12 NN02 C. NG03 NK21 NK01 NN11 NN01 D. NG01 NK21 NK01 NN15 NN03 Answer: A Answer Explanation:

Correct - This line up is from a safeguards source through the correct path to the NN bus.

Incorrect - PG20. This power supply is a backup for the safeguards supply and is not allowed without entering the TS for loss of power to the battery charger.

Plausible as this is a correct line up to the NN bus.

Incorrect - NG03. This line up connects the two safeguards buses not allowing train separation. Plausible if the student confuses power supplies for this system.

Incorrect - NG01. This line up connects the two safeguards buses not allowing train separation. Plausible if the student confuses power supplies for this system.

Meets KA asks for AC to DC to AC physical connection for the class 1E instrument bus RO knowledge system design and above the double line TS Rev 0

Question 46 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98361 User-Defined ID: 98361

Reference:

SY1506300 Topic: 46 RO AC power to NN bus per TS RO Importance Rating: 3.5 SRO Importance Rating: 4.0 K/A Number: 062 K 1.03 Comments: NEW Lesson Plan Objective: SY1506300 R5, Explain the relationship between Technical Specifications and the Class 1E 125V DC and Class 1E 120V AC power systems at the level of detail expected for the job position.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - AC electrical distribution - Knowledge of the physical connections and or cause effect relationships between the AC distribution system and the following systems - DC distribution Modification History:

Rev 0

47 ID: 98362 Points: 1.00 A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run of NE01 is in progress using STS KJ-011A, EDG NE01 24 HOUR RUN.

The EDG has been in parallel with offsite power for 90 minutes when the following indications are noted:

AC Megawatts 6801 kW and stable NB01 bus voltage 4.25 kV and stable Which of the following is correct with regard to this surveillance?

A. Lower bus voltage to prevent damage to the bus.

B. Lower bus voltage to prevent damage to the EDG.

C. Limit total run time at this kW level to prevent damage to the EDG.

D. Limit total run time at this kW level to prevent damage to the bus.

Answer: C Answer Explanation:

Correct - continuous loading of the EDG is limited to 6.2 Mw, any loading over that is limited by time. Loading over 6.2 to 6.8 Mw is limited to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by procedure. Operation with this high of loading can cause overheating and insulation breakdown over time.

Incorrect - limit total time to prevent damage to the bus. The bus can handle more load than the EDG so this is a correct limit for the EDG but not the bus.

Plausible if the student confuses the voltage and the MW limit.

Incorrect - lower voltage to prevent damage to the bus. This voltage is a little high but within normal limits. Plausible if the student confuses the voltage limit and the MW limit and believes that overall bus damage would occur at this voltage level.

Incorrect - lower voltage to prevent damage to the EDG. This voltage is a little high but within normal limits. Plausible if the student confuses the voltage limit and the MW limit.

Meets KA asks for ability to predict and or monitor parameters to prevent exceeding design limits with regards to overall EDG loading RO knowledge based on system precautions and limitations.

Rev 0

Question 47 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98362 User-Defined ID: 98362

Reference:

SYS KJ-011A Topic: 47 RO EDG overloaded rating for how long RO Importance Rating: 3.4 SRO Importance Rating: 3.8 K/A Number: 062 A 1.01 Comments: NEW Lesson Plan Objective: SY1406401 R3, Assess the functional interrelationship with the AC Distribution System.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - AC electrical distribution - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the AC distribution system controls including - Significance of DG load limits Modification History:

0 - modified based on feedback from Scott 4/25/15 Rev 0

48 ID: 98363 Points: 1.00 Given the following:

The unit is at full power DC Breaker NK0111, INVERTER NN11 input from NK01, fails and opens Which of the following describes the Main Control Board annunciator response AND the operation of the NN11 transfer switch?

A. 025A, NN01 INST BUS UV, is received The transfer switch automatically transfers to the bypass source B. 025A, NN01 INST BUS UV, is received The transfer switch must be manually transferred to the bypass source C. 025B, NN11 INV TRBL/XFR, is received The transfer switch automatically transfers to the bypass source D. 025B, NN11 INV TRBL/XFR, is received The transfer switch must be manually transferred to the bypass source Answer: C Answer Explanation:

Correct - With the loss of the input DC from the battery to the NN inverter the inverter will auto swap to the bypass source (static transfer switch). This will cause only the trouble alarm 025B to come into the control room. The NN bus is still energized and working just with a different power supply.

Incorrect - 025B and manually transferred. The alarm is correct but the operation of the transfer switch is auto not manual. Plausible since this piece of equipment was replaced with a new one in spring of 2015.

Incorrect - 025A and auto transferred. The NN bus will not lose voltage and will operate normally just on the bypass source. The second part is correct.

Plausible if the student thinks this DC input loss will cause the NN bus to become de-energized.

Incorrect - 025A and manually transferred. The NN bus will not lose voltage and will operate normally just on the bypass source. The operation of the transfer switch is auto not manual. Plausible if the student thinks this DC input loss will cause the NN bus to become de-energized and since this equipment was replace in spring of 2015.

Meets KA because student must know the expected response of the DC electrical system in order to monitor that response. As the DC system feeds the NN inverters and when there is a problem with the DC input the inverter will swap to an alternate source. This indicates that there is a problem with the DC source.

RO knowledge system understanding Rev 0

Question 48 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98363 User-Defined ID: 98363

Reference:

ALR 025B REV 9 48 RO MCB alarms with loss of DC supply to NN01 and Topic:

transfer switch RO Importance Rating: 2.7 SRO Importance Rating: 3.1 K/A Number: 063 A 3.01 Comments: BANK - Ginna Lesson Plan Objective: SY1506300 R7, Integrate system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 2 Group # 1 Last Used - 2012 Ginna #47 Comprehension 55.41 part 7 KA - DC electrical distribution - Ability to monitor auto operation of the DC electrical system including - Meters annunciators dials recorders and indicating lights Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

49 ID: 98364 Points: 1.00 The unit was operating at full power when a LOCA occurred.

20 seconds AFTER the LOCA sequencer activated a CSAS signal was generated.

At what time will the Containment Spray pumps start AND why?

A. Time 60 seconds, to prevent a trip of the EDG output breaker.

B. Time 40 seconds, to prevent a trip of the EDG output breaker.

C. Time 40 seconds, to ensure all Containment fans are running prior to spray starting.

D. Time 60 seconds, to ensure all Containment fans are running prior to spray starting.

Answer: B Answer Explanation:

E-12NF01; SY 1406401 and Figure 7.

If a CSAS is not present at the 15 second step on the LOCA sequencer circuit, the Containment Spray Pumps are prevented from automatically starting until the 25 second timer is completed. This timer starts at time 15 seconds on the LOCA sequencer if the spray signal does not exist at that time. The earliest it could start in this case would be 40 seconds (15 + 25 = 40). The reason is to allow the starting current to decay down prior to starting the next required load to prevent the EDG output breaker from tripping on over current.

Correct - based on the block of the CSAS until the LOCA sequencer is at 40 seconds.

Incorrect - time 40, ensure all containment fans are running. Time is correct.

The H2 fans shift to slow speed at time 60 seconds (separate timer than the LOCA sequencer). Plausible if the student knows the time but not the reason since fans do start all the way out to the 60 second mark.

Incorrect - time 60, prevent a trip of the EDG output breaker. The H2 fans shift to slow speed at time 60 seconds (separate timer than the LOCA sequencer).

Plausible if the student confuses which loads sequenced on the safeguards bus since the H2 fans do shift at this point.

Incorrect - time 60, ensure all containment fans are running. The H2 fans shift to slow speed at time 60 seconds (separate timer than the LOCA sequencer).

Plausible if the student confuses the loads sequenced on the safeguards bus and the reason for the delay in the start for the spray pump.

Meets KA asks system design to provide for auto load sequencer actuation with respect to the EDG RO knowledge system understanding Rev 0

Question 49 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98364 User-Defined ID: 98364

Reference:

E-12NF01 Topic: 49 RO CSAS and LOCA sequencer activation RO Importance Rating: 3.5 SRO Importance Rating: 4.0 K/A Number: 064 K 4.11 Comments: NEW Lesson Plan Objective: SY1406401 R5, Discuss the Sequencer System purpose.

Tier # 2 Group # 1 Last Used - N/A Memory 55.41 part 7, 8 KA - EDG - Knowledge of EDG system design features and or interlocks which provide for the following - Auto load sequencer, safeguards Modification History:

Rev 0

50 ID: 98365 Points: 1.00 At 10:00 Wolf Creek experienced a loss of off-site power.

At 10:30 Off-site power has been restored and is ready to be re-aligned to the safety related busses.

The RO is performing SYS NB-201, TRANSFERRING NB01 POWER SOURCES, to restore normal off-site power to NB01.

The synchroscope is placed in the Main Feeder Position and is rotating slowly in the SLOW direction (counter clockwise).

Incoming voltage is 4185 VAC NB01 bus voltage is 4130 VAC What actions must the RO take to parallel the EDG with the off-site source?

A. Raise EDG speed, Raise EDG voltage B. Lower EDG speed, Raise EDG voltage C. Raise EDG speed, Lower EDG voltage D. Lower EDG speed, Lower EDG voltage Answer: B Answer Explanation:

This starts in OFN NB-035, LOSS OF OFFSITE POWER RESTORATION, and then that sends you to the SYS procedure to complete the swap.

Correct - by lowering the speed of the diesel then the offsite source will be faster (sync scope is rotating with the incoming source not the diesel) and the sync permissives will be met. The voltage must be raised for the diesel so the breaker can be closed and because the procedure has the operator match voltage. This will allow all the permissives to be met and the normal feeder breaker to be closed.

Incorrect - Lower and lower. Lower speed is correct. Lowering voltage will cause the diesel to be greater than 50 volts lower than offsite power. The procedure says to match voltage with offsite. Plausible if the student confuses which source is on the bus and which source is incoming.

Incorrect - Raise and lower. If speed is raised then the sync scope will rotate fast in the slow direction since it is looking and the incoming source compared to the running source. Lowering voltage will cause the diesel to be greater than 50 volts lower than offsite power. The procedure says to match voltage with offsite.

Plausible if the student confuses which source is on the bus and which source is incoming.

Incorrect - Raise and raise. If speed is raised then the sync scope will rotate fast in the slow direction since it is looking and the incoming source compared to the running source. Raise voltage is correct.

Rev 0

Meets KA asks for ability to operate from the control room parallel operation of the diesel to the grid with load on the diesel.

RO knowledge system understanding of parallel operations for matching voltage and speed for sync Question 50 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98365 User-Defined ID: 98365

Reference:

SYS NB-201 50 RO parameters checked as the EDG is synced with Topic:

offsite power RO Importance Rating: 3.4 SRO Importance Rating: 3.4 K/A Number: 064 A 4.07 Comments: BANK - Callaway Lesson Plan Objective: SY1406401 R3, Assess the functional interrelationship with the AC Distribution System.

Tier # 2 Group # 1 Last Used - 2011 Callaway #47 Comprehension 55.41 part 7 KA - EDG - Ability to manually operate and or monitor in the control room - Transfer EDG with load to the grid Modification History:

0 - replaced based on Scotts comments 4/25/15 Rev 0

51 ID: 98366 Points: 1.00 The RO is performing a source check of a detector on RM-11R (SP056A).

What is the reason for performing this function prior to an effluent release?

A. Cause the ALERT alarm to come in to check its setpoint.

B. Calibrate the detector prior to performing a radioactive release.

C. Prove the monitor is functional to ensure monitoring of the release.

D. Check that the display colors change due to increased rad levels from the detector.

Answer: C Answer Explanation:

SYS SP-121, ALR 062C, SY1407200. All discuss the way to perform a check source and the proper outcome of the test to ensure the monitor is operable. It also discusses what is seen if the check source fails. Precaution of the SYS states that if the monitor check source is energized and the detector reaches the source limit that it is a SAT check. The source check is only about a third decade per minute so no alarms will come in due to the check. The check is to determine if the monitor can detect radiation only, not a calibration check of alarm setpoints or accuracy.

Correct - per the SYS the monitor check is to determine if the monitor is operable or not.

Incorrect - Calibrate the detector. The check source is only a response test not a calibration of the detector. Plausible as the detector does need to be calibrated prior to use but this is not the correct way to perform that.

Incorrect - Cause the alert alarm. This will only indicate that the detector will show a response not raise it to the level of an alarm. Plausible since all the detectors do have alarm setpoint that should be checked prior to a release but this check will not set them off.

Incorrect - Display color change. While this will cause the display color to change from green to half intensity cyan for the duration of the test it will go back to green when completed sat. The color change is not why the test is ran just an indication that the test is running. Plausible since this does take place but not to just see the color change its for response check.

Meets KA asks ability to operate in the control room a check source for a detector RO knowledge system understanding of the rad monitor Rev 0

Question 51 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98366 User-Defined ID: 98366

Reference:

SYS SP-121 Topic: 51 RO RM-11 check source reason RO Importance Rating: 3.1 SRO Importance Rating: 3.2 K/A Number: 073 A 4.03 Comments: NEW Lesson Plan Objective: SY1407200 R3, EXPLAIN the basic operation of the Area Radiation Monitoring System components.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 11 KA - Process Rad Monitor - Ability to manually operate and or monitor in the control room - Check source for operability demonstration Modification History:

Rev 0

52 ID: 98367 Points: 1.00 Given the following plant conditions:

The unit is at 100% power.

Annunciator 009B, SERV WTR PMP TRIP, alarms.

Investigation reveals that ALL Service Water pumps have tripped and CANNOT be started.

Which ONE of the following describes actions required by the crew?

A. Start both ESW Trains, with one train supplying Service Water. Trip the Turbine if any Turbine trip setpoint is reached.

B. Start both ESW Trains, with one train supplying Service Water. Trip the Reactor if any Turbine trip setpoint is reached.

C. Place both ESW Trains in service. Trip the Turbine if any Turbine trip setpoint is reached.

D. Place both ESW Trains in service. Trip the Reactor if any Turbine trip setpoint is reached.

Answer: D Answer Explanation:

Technical

References:

ALR 00-009B AP 15C-003 SYS EF-200 CORRECT - Based on system response the reactor is tripped to ensure fuel integrity.

Incorrect - Place both ESW Trains in service. Trip the Turbine if any trip setpoint is reached. INCORRECT, plausible since a Turbine trip setpoint may be challenged, however to ensure adequate core cooling the reactor is tripped first and the P-4 signal will trip the turbine. Tripping the turbine will not necessarily trip the reactor if a load runback occurred due to stator water temperature increasing.

Incorrect - Start both ESW Trains, with one train supplying Service Water. Trip the Reactor if any Turbine trip setpoint is reached. INCORRECT, plausible since the loss of heat sink to the secondary components is the concern.

Incorrect - Start both ESW Trains, with one train supplying Service Water. Trip the Turbine if any trip setpoint is reached. INCORRECT, plausible since the loss of heat sink to the secondary components is the concern.

Meets KA asks for ability to predict impact of loss of service water has on plant and use procedures to correct RO knowledge overall understanding of system interrelations Rev 0

Question 52 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98367 User-Defined ID: 98367

Reference:

ALR 009B Topic: 52 RO loss of service water with the reactor at 100%

RO Importance Rating: 3.5 SRO Importance Rating: 3.7 K/A Number: 076 A 2.01 Comments: BANK - 46137 Lesson Plan Objective: LO1733203 R14, DISCUSS procedure implementation IAW AP 15C-003.

Tier # 2 Group # 1 Last Used - 2012 Palo Verde #52, 2009 Callaway #51 Comprehension 55.41 part 10 KA - Service water - Ability to predict the impacts of the following malfunction or operations on the SWS and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Loss of SWS Modification History:

0-Rev 0

53 ID: 98368 Points: 1.00 Given the following:

The plant is in MODE 4 RHR Train "A" is in service RHR Heat Exchanger Bypass Valve EJ FCV-618 is set to maintain 3400 GPM RHR Heat Exchanger Outlet Valve EJ HCV-606 demand position set at 30%

The Instrument Air supply line to RHR Heat Exchanger Bypass Valve EJ FCV-618 becomes severed and is completely detached No other air operated valves are impacted by the failure Which ONE of the following describes the plant parameter changes from the initial steady state conditions?

RCS Temperature Total RHR flow A. Higher Higher B. Lower Higher C. Lower Lower D. Higher Lower Answer: C Answer Explanation:

Correct - FCV-618 fails closed, so there is less bypass flow mixing with more HX flow, resulting in a lower temperature on the HX outlet.

Incorrect - higher/higher. Total RHR flow is controlled by FCV-618 and would lower, forcing more water through the RHR HX for cooling.

Incorrect - higher/lower. FCV 618 failing closed will result in full cooling through the RHR HX and the HX outlet temperature will lower along with the total RHR flow lowering. Plausible because the applicant may confuse valves for total flow versus HX flow.

Incorrect - lower/higher. Temperature effect is correct and plausible because the applicant may confuse valves for total flow versus HX flow Meets KA asks system response to a loss of air RO knowledge system design Rev 0

Question 53 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98368 User-Defined ID: 98368

Reference:

M-12EJ01 Topic: 53 RO EJ RHR loss of EJ FCV-618 RO Importance Rating: 3.4 SRO Importance Rating: 3.6 K/A Number: 078 K 3.02 Comments: BANK - 58852 Lesson Plan Objective: SY1300500 R7, INTEGRATE system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 2 Group # 1 Last Used - 2007 Callaway Comprehension 55.41 part 7 KA - Instrument air - Knowledge of the effect that a loss or malfunction of the air system will have on the following -

Systems having pneumatic valves and controls Modification History:

Rev 0

54 ID: 98369 Points: 1.00 The following plant conditions exist:

Compressor Sequencer Selector Switch, KA HSS-310, is selected to the C-A-B position All three air compressors are selected to AUTOMATIC

'A' Air Compressor (CKA01A) is running unloaded

'B' Air Compressor (CKA01B) is NOT running

'C' Air Compressor (CKA01C) is running loaded KA PV-11, Service Air Isolation Valve, is open What is the automatic system response to an air leak that results in air system pressure decreasing to 105 psig?

A. All three air compressors are running, but only two are loaded. KA PV-11 remains OPEN.

B. Only CKA01A and CKA01C are running and both are loaded. KA PV-11 remains OPEN.

C. All three air compressors are running and all three are loaded, KA PV-11 CLOSES.

D. Only CKA01A and CKA01C are running and both are loaded, KA PV-11 CLOSES.

Answer: C Answer Explanation:

Lead cycles between 116 and 125 psig Lag cycles between 114 and 123 psig Standby cycles between 112 and 121 psig KA PV-11 closes at 110 psig.

With pressure at 105 psig, all three compressors are running loaded and KA PV-11 valve is closed.

Correct - at 105 psig all the compressors should be loaded and KA PV-11 will have closed at 110 psig.

Incorrect - only A and C running with KA PV-11 open. With pressure this low all are running and 11 is closed. Plausible if the student confuses values of start stop and or sequencer switch position.

Incorrect - all three running two loaded with KA PV-11 open. With pressure this low all are running and 11 is closed. Plausible if the student confuses values of start stop and or sequencer switch position.

Incorrect - A and C running with KA PV-11 open. With pressure this low all are running and 11 is closed. Plausible if the student confuses values of start stop and or sequencer switch position.

Meets KA asks auto operation of the instrument air system based on pressure Rev 0

RO knowledge system level of understanding of system Question 54 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98369 User-Defined ID: 98369

Reference:

SY1407800 Topic: 54 RO KA Air Compressor loading RO Importance Rating: 3.1 SRO Importance Rating: 3.2 K/A Number: 078 A 3.01 Comments: BANK -18505 Lesson Plan Objective: SY1407800 R5, Explain the alarms, controls, indications, and interlocks associated with the system.

Tier # 2 Group # 1 Last Used - 2015 ILO systems Fundamental 55.41 part 7 KA - Instrument Air - Ability to monitor automatic operation of the IAS including: Air pressure Modification History:

Rev 0

Q55, Rev 1 ID: 98370 Points: 1.00 Given the following:

• Core off load is in progress

• Equipment Hatch is open

• ALL MSIVs are closed

• ALL S/G secondary manways are removed

• Both doors in the Personnel Air Lock are OPEN but OPERABLE

• ONE door is closed in the Emergency Air Lock

• Control Room was just now informed that last night maintenance removed ONE 'B' S/G safety valve and shipped it offsite for calibration

• NO compensatory actions are in place Which of the following actions (if any) will be required at this time?

A. NO action is required at this time B. IMMEDIATELY suspend fuel movement in Fuel Building C. IMMEDIATELY suspend fuel movement in Containment D. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> install a blank flange over the removed S/G safety opening Answer: C Answer Explanation:

T.S. 3.9.4 Correct - This is an immediate action in TS 3.9.4 with fuel movement in containment.

Incorrect - No action required. Plausible if the student misunderstands containment integrity with respect to the SG manways and the safety being removed.

Incorrect - Immediately stop fuel movement in fuel building. TS only discusses in containment. Plausible if the student misunderstands the relationship of the fuel building and containment with regards to integrity.

Incorrect - Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> install blank flange. This will restore containment integrity but the TS action with fuel movement in progress is immediately not 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible if the student forgets the TS requirement.

Meets KA by asking what to do if CTMT integrity is lost during fuel movement RO knowledge this is TS above the line and less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q55, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98370 User-Defined ID: 98370

Reference:

T.S. 3.9.4 Topic: 55 RO containment closure during fuel movement RO Importance Rating: 3.7 SRO Importance Rating: 4.1 K/A Number: 103 K 3.03 Comments: NEW Lesson Plan Objective: LO1732701 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 9, 10 KA - Containment - Knowledge of the effect that a loss or malfunction of the containment system will have on the following - Loss of containment integrity under refueling operations Modification History:

0 - modified based on Scotts feedback 4/25/15 1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

56 ID: 98371 Points: 1.00 Given the following with the unit at 100% power:

OFN BB-007, RCS LEAKAGE HIGH, was entered based on PZR level lowering

'B' CCP is running BG FK-121, CCP DISCH FLOW CTRL, is in manual with 100% output demand ALL letdown orifice isolation valves are CLOSED PZR level is 48% and down at 1% per minute RCS pressure is stable at 2231 psig As the crew performs the appropriate actions which of the following describes the injection path for the CCPs FIVE minutes later?

A. Both CCPs are injecting into ALL four cold legs in ECCS mode B. 'A' CCP is injecting into loop 1 & 2 cold legs in ECCS mode ONLY

'B' CCP is injecting into loop 3 & 4 cold legs in ECCS mode ONLY C. 'A' CCP is secured

'B' CCP is injecting into the loop 1 cold leg in charging mode D. Both CCPs are injecting into the loop 1 cold leg in charging mode Answer: A Answer Explanation:

Correct - OFN BB-007 foldout page states that if charging is maximized from one pump with letdown isolated and PZR level lowering then trip the Rx and actuate SI (ECCS mode injection). Five minutes later that crew will have performed this and still be in EMG E-0 working through it. The CCPs will be both running, due to the SI, and injecting through the BIT header which is both pumps discharge coming together and flowing to all four cold legs in the ECCS mode.

Incorrect - A CCP secured and B injecting. OFN BB-007 will have the crew actuate SI and this will realign the pumps to ECCS injection mode. Plausible if the student forgets the OFN foldout page criteria for SI and thinks that the crew will continue to work through the OFN since PZR level is lowering very slow.

Incorrect - Both pumps injecting in charging mode. Both pumps will be running due to the SI but will not be using the normal charging flowpath. Plausible if the student remembers the SI off of foldout page but confuses where the injection for the CCPs is for the ECCS mode vs normal charging mode.

Incorrect - A CCP injecting into 1 and 2 B CCP injecting into 3 and 4. The SI realign both pumps to be running but not to the loops given. They will both be injecting to all loops. Plausible if the student confuses the RCS tap points for the CCPs between the normal charging and ECCS mode.

Meets KA asks for physical connection from the RCS to ECCS pumps (CCPs not used in charging mode)

RO knowledge system connection understanding Rev 0

Question 56 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98371 User-Defined ID: 98371

Reference:

M-12BB01 REV 32 56 RO RCS leak from OFN BB-007 to ECCS injection path Topic:

for CCPs RO Importance Rating: 4.5 SRO Importance Rating: 4.6 K/A Number: 002 K 1.08 Comments: NEW Lesson Plan Objective: SY1300600 R7, Determine the flow path(s), including major valve positions and pump alignments, during each phase of ECCS operation.

Tier # 2 Group # 2 Last Used - N/A Memory 55.41 part 7 KA - RCS - Knowledge of the physical connections and or cause effect relationships between the RCS and the following system - ECCS Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

57 ID: 98372 Points: 1.00 The plant is operating at 100% when the upper selected PZR level channel fails to 0%.

Which of the following is an expected plant response to this failure AND why?

(assume NO operator action)

A. Actual PZR level lowers due to charging flow lowering.

B. PZR spray valves will open due to actual PZR pressure rising.

C. ALR 033E, PZR HTR GROUP LOCKOUT, will alarm due to heaters being tripped off.

D. PZR backup heaters will energize due to actual PZR level being greater than 5% above program.

Answer: B Answer Explanation:

M-744-0028 Correct - With this failure letdown isolates and charging goes to max so no water out and more water in will raise pressure in the PZR. Heaters are off due to letdown isolation at 17% (failed channel low) so this is not adding to the pressure rise. Sprays open to lower pressure but will not lower pressure back to 2235 psig.

Incorrect - ALR 033E. This alarm will come in if any PZR heater breaker is tripped open. Since the failed channel failed low the breakers are not tripped but locked out from coming on. Plausible if student confuses how the PZR heaters operate under these conditions.

Incorrect - PZR backup heaters energize. These heaters are controlled by the upper selected channel so when it fails low they backup heaters are lock out from coming on. Plausible since actual level will rise and be greater than 5% above program which should have turned them on baring the failure.

Incorrect - PZR level lowers. This would happen if the level channel had failed HI. Plausible if the student confuses the different failure modes, when letdown isolates then charging should lower to keep from over filling the PZR.

Meets KA by asking how PZR level failure affects PZR pressure control system RO knowledge system understanding Rev 0

Question 57 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98372 User-Defined ID: 98372

Reference:

SY1301000 Topic: 57 RO PZR level channel failure effect on pressure RO Importance Rating: 3.2 SRO Importance Rating: 3.7 K/A Number: 011 K 3.03 Comments: NEW Lesson Plan Objective: SY1301000 R10, Predict the impact of a given instrument failure, heater failure, or spray valve failure on the pressurizer pressure and/or level control system.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.41 part 5, 7 KA - PZR level control - Knowledge of the effect that a loss or malfunction of the PZR level control system will have on the following - PZR PCS Modification History:

0 - modified based on Scotts feedback 4/25/15 Rev 0

Q58, Rev 1 ID: 98373 Points: 1.00 The plant is operating at 100% power when the following takes place:

• Reactor trip occurs

• 'A' train reactor trip breaker is closed

• 'B' train reactor trip breaker is open With NO operator action what affect (if any) will this failure have on the operation of the steam dump system?

A. Steam dumps will ARM and RCS temperature will go to 557°F.

B. Steam dumps will ARM and RCS temperature will go to 559°F.

C. Steam dumps will NOT ARM and RCS temperature will go to 557°F.

D. Steam dumps will NOT ARM and RCS temperature will go to 561°F.

Answer: A Answer Explanation:

Correct - The A train reactor trip breaker controls steam dump arming but arming is also controlled by loss of load from C-7 so steam dumps are armed. B train reactor trip breaker controls which controller the steam dumps work off of. With this breaker open the steam dumps will be controlled by the plant trip controller.

Incorrect - armed and 559F. Plausible if the student confuses which reactor trip breaker does what. This statement is if they reversed the correct breaker inputs.

Incorrect - not armed and 557F. Plausible since the A train reactor trip breaker still being closed means no arming from it but C-7 does arm the steam dumps. Since the B train breaker opened the plant trip controller is controlling the temperature at 557F.

Incorrect - not armed and 561F. Plausible since the A train reactor trip breaker still being closed means no arming from it but C-7 does arm the steam dumps. But if the student doesn't think the steam dumps are armed then he would think the temperature is being controlled by the ARVs which are set at 561F Meets KA asks if student can monitor the auto selection of NNIS inputs with regards to steam dump controller used during a reactor trip with a failure RO knowledge basic system design and interlocks OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q58, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98373 User-Defined ID: 98373

Reference:

SY1504100 58 RO steam dump controlling temp with reactor trip breaker Topic:

closed RO Importance Rating: 2.9 SRO Importance Rating: 2.9 K/A Number: 016 A 3.01 Comments: NEW Lesson Plan Objective: SY1504100 R3, Explain the various modes of system operation.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.41 part 7 KA - Non nuclear instrumentation - Ability to monitor auto operation of the NNIS including - Auto selection of NNIS inputs to control systems.

Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q59, Rev 1 ID: 98374 Points: 1.00 The unit has experienced an earthquake, loss of Off Site power, and a LOCA. The crew has transitioned to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT. The following indications are noted:

• RCS pressure 1790 psig

• CETC temperature 1210 °F

• RVLIS 55% down slow

• 'A' CCP OOS due to planned maintenance

• 'B' CCP Breaker tripped and will NOT reset

• Total AFW flow 150,000 lbm/hr stable

• All S/G NR levels 2% thru 8%

Based on the indications given the crew will...

A. transition to EMG FR-C1, RESPONSE TO INADEQUATE CORE COOLING B. transition to EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING C. transition to EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK D. continue with EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT Answer: A Answer Explanation:

With a loss of offsite power RCP's are off. Subcooling for the 1790 psig is around 621°F and with CETC at 1150°F subcooling is not met. Also with the power loss the NCP is unavailable.

FR procedures are monitored after the transition out of E-0 is made.

Correct - Based on CETCs over 1200F the entry condition has been made for red path to C1.

Incorrect - FR-C2 orange. Plausible if the student fails to use the 1200F as a entry condition and moves on down the F-0 to the orange path for C2 based on RCPs stopped, RVLIS, and CETC.

Incorrect - FR-H1 red path. AFW flow is lower than the 270,000 lbm/hr requirement but since SG levels are above 6% NR this can be lowered. Plausible if the student mistakes the low flow with a requirement from F-0 to go here.

Incorrect - E-1. RCS pressure is used in multiple locations throughout this procedure to determine if ECCS flow should be reduced and with pressure staying high this is a concern with RHR pumps. Plausible if the student mistakes high RCS pressure with the need to secure RHR pumps and reduce ECCS flow over the red path entry.

Meets KA asks ability to monitor core exit temperature to prevent exceeding a design limit so a procedure change is needed RO knowledge this is knowing the entry conditions for FR red or orange paths OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question 59, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98374 User-Defined ID: 98374

Reference:

EMG F-0 Topic: 59 RO C1 red path entry RO Importance Rating: 3.7 SRO Importance Rating: 3.9 K/A Number: 017 A 1.01 Comments: NEW Lesson Plan Objective: LO1732341 R6, DISCUSS the entry conditions for procedure EMG FR-C2.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.41 part 5, 7, 10 KA - In core temp monitor - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the in core temp monitor system controls including - Core exit temperature Modification History:

Rev 1: ??

Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q60, Rev 1 ID: 98375 Points: 1.00 The unit is operating at 100% power at end of life when the following takes place:

• RO reports SFP level is -25" and lowering rapidly

• OFN EC-046, FUEL POOL COOLING AND CLEANUP MALFUNCTIONS, has been entered

• Aux Building operator confirms RO report using the local level indicator

• The leak location is currently UNKNOWN Which of the following areas is required to be evacuated AND what makeup source is preferred?

A. Fuel Building Area ONLY, RWST B. Fuel Building Area ONLY, RMWST (water only)

C. Fuel Building Area AND the Aux Building, RWST D. Fuel Building Area AND the Aux Building, RMWST (water only)

Answer: A Answer Explanation:

Correct - Foldout page of the OFN directs to Attachment A which evacuates the fuel building and makeup is from a borated water source first.

Incorrect - fuel building and makeup from RMWST. The first part is correct but makeup will be from a borated water source before an unborated sources is used and since the stem says nothing about an issue with the RWST it will be used. Plausible if the student forgets what type of water to use to fill the pool.

Incorrect - fuel and aux building makeup from RWST. Makeup source is correct but the procedure only has the fuel building evacuated. Foldout page items are required knowledge for RO's. Plausible if the student thinks the two buildings have the same ventilation system.

Incorrect - fuel and aux building makeup from RMWST. Procedure only has the fuel building evacuated. Makeup will be from a borated source first. Plausible if the student confuses the ventilation systems or where makeup will come from first.

Meets KA asks ability to use procedures to correct low SFP level issues.

RO knowledge this is foldout page item and the makeup is system understanding of which source is the correct one for the SFP.

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98375 User-Defined ID: 98375

Reference:

OFN EC-046 Topic: 60 RO loss of SFP level actions per OFN RO Importance Rating: 3.1 SRO Importance Rating: 3.5 K/A Number: 033 A 2.03 Comments: NEW Lesson Plan Objective: LO1732454 R3, Given a procedure flow path, EXAMINE the available options for procedure actions.

Tier # 2 Group # 2 Last Used - N/A Fundamental 55.41 part 10 KA - SFP cooling - Ability to predict the impacts of the following malfunctions or operations on the SFP cooling system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Abnormal spent fuel pool water level or loss of water level Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

61 ID: 98376 Points: 1.00 Spent Fuel assemblies are being moved in preparation for an upcoming refueling.

Fuel Building ventilation is in normal lineup.

Fuel Building Rad Monitor GG RE-28 loses power.

Which of the following automatic actuations occur, if any?

A. Fuel Building Supply Fan stops, both trains of Emergency Exhaust start.

B. Fuel Building Supply Fan stops, ONLY 'B' train of Emergency Exhaust starts.

C. Fuel Building Supply Fan remains running, both trains of Emergency Exhaust start.

D. Fuel Building Supply Fan remains running, ONLY 'B' train of Emergency Exhaust starts.

Answer: A Answer Explanation:

The loss or malfunction of a rad monitor will have NO effect on fuel handling system only a procedure action if this happens.

Correct - FBIS isolates the fuel building and sends a signal to isolate the control room as well.

Incorrect - fuel building supply fan stops, one train emergency exhaust starts.

Plausible if the student misunderstands that the rad monitors in the fuel building cross trip to both trains.

Incorrect - fuel building supply fan remains running, both trains of emergency exhaust starts. Plausible if the student remembers that the emergency system will start but forgets that the normal system stops.

Incorrect - fuel building supply fan remains running, one train emergency exhaust starts. Plausible if the student misunderstands that the rad monitors in the fuel building cross trip to both trains and that the normal system will stop when this occurs Meets KA by showing how a loss of rad monitors in the fuel building will affect the ventilation system which is part of the fuel handling system isolation.

RO knowledge system knowledge of ESFAS actuation signals Rev 0

Question 61 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98376 User-Defined ID: 98376

Reference:

SY1408803 Topic: 61 RO what gives a FBIS RO Importance Rating: 2.6 SRO Importance Rating: 3.3 K/A Number: 034 K 6.02 Comments: BANK - 47936 Lesson Plan Objective: SY1408803 R9, DISCUSS the function of major Fuel Building HVAC System components and controls.

Tier # 2 Group # 2 Last Used - Commanche Peak 2007 Comprehension 55.41 part 11 KA - Fuel handling equipment - Knowledge of the effect of a loss or malfunction of the following will have on the fuel handling system - Radiation monitoring systems Modification History:

Rev 0

62 ID: 98377 Points: 1.00 What is ONE purpose of the S/G Flow Restrictor?

A. Restricts flow of steam to the Main Turbine.

B. Provides a measuring point for steam flow rate.

C. Provides for the upper tap of the S/G WR level detector.

D. Provides a tap for the Main Steam Header pressure detector.

Answer: B Answer Explanation:

Correct - One of the purposes of the flow restrictor.

Incorrect - provides tap for steam pressure. Plausible since this does supply a point for the steam flow which uses pressure to density compensation.

Incorrect - restricts flow to main turbine. While it does provide a flow restriction on a steam line break (SG pressure to atmosphere) it doesn't restrict flow to the main turbine. The flow rate to the main turbine is lower than the flow would be on a steam line break. Plausible if the student confuses what flow is being restricted.

Incorrect - provides tap for SG WR level. Since there is a tap for steam flow here it is plausible that the upper tap for SG level could be at this same location.

Meets the KA asks purpose of SG component RO knowledge system understanding Rev 0

Question 62 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98377 User-Defined ID: 98377

Reference:

SY1503900 Topic: 62 RO purpose of the SG flow restrictor RO Importance Rating: 3.9 SRO Importance Rating: 4.0 K/A Number: 035 2.1.27 Comments: NEW Lesson Plan Objective: SY1503900 R1, Organize into a flow path the major components of the main and reheat steam system.

Tier # 2 Group # 2 Last Used - N/A Fundamental 55.41 part 8 KA - S/G - Conduct of ops - Knowledge of system purpose and or function Rev 0

63 ID: 98378 Points: 1.00 During a plant heat up RCS temperature is to be maintained automatically at 500°F using Condenser Steam Dumps.

What Steam Pressure Controller (AB PK-507) setting is required?

A. 3.33 B. 4.44 C. 4.54 D. 4.64 Answer: B Answer Explanation:

This controller is a 10 turn pot with 0-1500 psig Steam Tables:

500°F = 680.86 psia 680.86 psia - 14.7 psia = 666.16 psig Correct - 666.16 psig/150psig/turn = 4.44 turns Incorrect - 4.54, plausible since 680.86/150 Incorrect - 4.64, plausible since 680.86 + 14.7/150 Incorrect - 3.33, plausible since 500/150 Meets KA asks for ability to operate the steams dumps in steam pressure mode RO knowledge system understanding of the steam dumps in steam pressure mode Rev 0

Question 63 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98378 User-Defined ID: 98378

Reference:

STEAM TABLES 63 RO steam dump setpoint for maintaining RCS temp in Topic:

pressure mode RO Importance Rating: 2.7 SRO Importance Rating: 2.7 K/A Number: 041 A 4.04 Comments: BANK - 17644 Lesson Plan Objective: SY1504100 R3, Explain the various modes of system operation.

Tier # 2 Group # 2 Last Used - 2015 Systems exam #15 Comprehension 55.41 part 5, 7 KA - Steam dump - Ability to manually operate and or monitor in the control room - Pressure mode Rev 0

64 ID: 98379 Points: 1.00 The operators are raising power from 35% to full power. Current RCS boron concentration is 630 ppm.

Which of the following will be used to control reactivity during the power increase?

A. Use control rods to maintain I on the 100% power target.

B. Allow rods to move automatically to follow turbine load without diluting.

C. Use rods and normal dilution to maintain I and Tavg / Tref within band.

D. Maintain the control rods at the 'parked position' and use normal dilution to change power.

Answer: C Answer Explanation:

Correct - since a combination of rods and dilution will raise power, delta flux has to be within limits above 50% power per TS 3.2.3 and QPTR is valid at or above 50% TS 3.2.4.

Incorrect - Use control rods to maintain delta-I on 100% power target. Axial Offset (AO) is maintained as close to 100% value as possible, however, delta-I can be far from the 100% target since rods start out at a very low position causing delta-I to be much more negative than AO. Plausible if the student misunderstands the relationship between rods and boron as power changes.

Incorrect - Maintain the control rods at the 'parked position' and use normal makeup to change power. If power were higher to start with this could be true but with power crossing the 50% value rods must be moved to maintain delta-I.

Plausible if the student doesn't recognize the power change encompasses TS items.

Incorrect - Allow rods to move automatically to follow turbine load without diluting.

If power were to be lowered this would be correct rods are used in auto to help follow the turbine. Also rods will reach their full out position prior to the unit reaching full power so dilution will be required. Plausible if the student mistakes raising and lowering power rod control or misunderstands that rods will be full out prior to full power.

Meets KA asks knowledge of operations impacts of rods and boration/dilution for turbine load changes RO knowledge system interrelations of rod control and boron (CVCS) for turbine load changes Rev 0

Question 64 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98379 User-Defined ID: 98379

Reference:

GEN 004 64 RO raise power from 35% to 100% rods and dilution Topic:

relation RO Importance Rating: 2.7 SRO Importance Rating: 2.8 K/A Number: 045 K 5.23 Comments: BANK - 19533 Lesson Plan Objective: LO1732104 R5, EXPLAIN the major steps of procedure GEN 00-004.

Tier # 2 Group # 2 Last Used - Normal ops #2, WC 2001 Memory 55.41 part 1 KA - Main turbine generator - Knowledge of the operational implications of the following concepts as they apply to the main turbine generator - Relationship between rod control and RCS boron concentration during TG load increases Rev 0

65 ID: 98380 Points: 1.00 What type of fire detection device detects the presence of abnormal heat AND where are they most widely used at Wolf Creek?

A. Infrared Flame detector, in the EDG rooms.

B. Ionization detectors, in the Turbine Building.

C. Protectowire Linear Heat Detectors, in Containment.

D. Photoelectric Smoke detector, in areas of higher radiation levels.

Answer: C Answer Explanation:

Correct - Protectowire. Detects temperature in a given location based on melting of the plastic coating separating the wires. This is used only in containment.

Incorrect - Photoelectric. Detects smoke (something to block light). Plausible as this is a device used in plant to detect fire but it doesn't look for hi temperature Incorrect - Infrared flame. Detects light emitted by a flame. Plausible as this is a device used in plant to detect fire but it doesn't look for hi temperature Incorrect - Ionization. This detector that detects combustion products in the air (particles from material that have burned). This detects fire before smoke and flame is present. Plausible as this is a device used in the plant to detect fire but it doesn't look for hi temperature.

Meets KA asks about detection devices RO knowledge system design understanding Rev 0

Question 65 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98380 User-Defined ID: 98380

Reference:

SY1408600 Topic: 65 RO Protecowire detector operation and where used RO Importance Rating: 3.1 SRO Importance Rating: 3.7 K/A Number: 086 K 4.03 Comments: NEW Lesson Plan Objective: SY1408600 R3, Explain the characteristics of the system major components.

Tier # 2 Group # 2 Last Used - N/A Fundamental 55.41 part 7 KA - Fire protection - Knowledge of design features and or interlocks which provide for the following - Detection and location of fires Rev 0

66 ID: 98381 Points: 1.00 The plant has experienced an event and entered the EMG procedure network. The CRS has assigned you as the RO to monitor two Continuous Action steps from EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT. The crew now transitions to EMG FR-C1, RESPONSE TO INADEQUATE CORE COOLING The Continuous Actions from EMG E-1...

A. become optional upon entry into EMG FR-C1.

B. are NOT applicable upon entering EMG FR-C1.

C. remain applicable throughout the performance of EMG FR-C1.

D. remain applicable until superseded by directed actions of EMG FR-C1.

Answer: B Answer Explanation:

AP 15C-003 6.6.6 and 7 After transitioning to another procedure, continuous action steps are applicable unless superseded by alternate guidance in the new procedure or stated to be inapplicable.

If a Red or Orange path ERG is entered, any continuous action steps from the previous procedures should not be performed. Entry into a Red or Orange path ERG indicates that plant conditions have severely degraded and the strategy of the suspended procedures is not effective.

Correct - since a C-1 is a red path procedure any continuous actions prior to entry are no longer performed.

Incorrect - remain applicable until superseded by directed actions of EMG FR-C1. Plausible if the student misunderstands the requirement for continuous action steps in FRs.

Incorrect - are NOT applicable upon transitioning from EMG E-1. Plausible if the student confuses the transition to FRs and other EMGs since normally they do apply.

Incorrect - become optional upon entry into EMG FR-C1. Plausible since the continuous actions are applicable if transitions are made to other EMG procedures that are not red or orange path until superseded.

Meets KA asks for a conduct of ops knowledge item which procedure use is one of RO knowledge procedure rules of usage is an admin requirement for RO's to understand and be able to adhere to without the procedure in hand Rev 0

Question 66 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98381 User-Defined ID: 98381

Reference:

AP 21C-003 Topic: 66 RO admin AP continuous action steps RO Importance Rating: 3.8 SRO Importance Rating: 4.2 K/A Number: 2.1.1 Comments: BANK - 17424 Lesson Plan Objective: LO1732312 R1, DISCUSS the EMG rules of usage in accordance with procedure AP 15C-003.

Tier # 3 Group #

Last Used - WC 2001 Memory 55.41 part 10 KA - Conduct of ops - Knowledge of conduct of operations requirements Rev 0

67 ID: 98382 Points: 1.00 Which of the following conditions will have the crew:

Trip the reactor Enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION A. PZR level is 91%

B. 'A' S/G NR level is 77%

C. RCS pressure is 2185 psig D. 'B' RCP Frame vibration is 8 mils Answer: D Answer Explanation:

AP 21-001, CONDUCT OF OPERATIONS, if any trip setpoint has been reached and no auto reactor trip has worked after you verify the validity of the alarm then the reactor must be placed in a safe condition.

Correct - The RCP has exceeded the rapid shutdown criteria for the pump which will require a manual reactor trip and a stop of the affected RCP.

Incorrect - PZR level. PZR level has not exceeded the auto trip setpoint.

Plausible if the student confuses the trip setpoint with action contained in OFNs for this condition.

Incorrect - RCS pressure is low. RCS pressure has not lowered enough to cause an auto reactor trip and one is not warranted at this point. Plausible if the student thinks they should try and 'beat' the reactor trip before any setpoint is reached.

Incorrect - SG MFRV. At 77% SG level a FWIS has not been generated.

Plausible if the student confuses the trip setpoint with action contained in OFNs for this condition.

Meets KA asks what should the operator do if system limits or precautions are exceeded (conduct of ops)

RO knowledge as they are required to know when an auto trip should have occurred or if they need to take action to protect the reactor.

Rev 0

Question 67 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98382 User-Defined ID: 98382

Reference:

AP 21-001 Topic: 67 RO conduct of ops when to trip if no auto trip works RO Importance Rating: 3.8 SRO Importance Rating: 4.0 K/A Number: 2.1.32 Comments: NEW Lesson Plan Objective: LO1733211, R7, DISCUSS the purpose / scope and selected knowledge requirements of procedure AP 21-001, Conduct Of Operations.

Tier # 3 Group #

Last Used - N/A Memory 55.41 part 10 KA - Conduct of ops - Ability to explain and apply system limits and precautions Rev 0

68 ID: 98383 Points: 1.00 Which of the following actions REQUIRES a peer check prior to performance?

A. Changing rod position for rod parking.

B. Tripping of the main turbine during an ATWS event.

C. Hanging a clearance order tag in a High Radiation area.

D. Placing a MFRV in manual after a S/G controlling level fails.

Answer: A Answer Explanation:

Correct - This is required since rod parking is a planned event that will change reactivity.

Incorrect - tripping the main turbine during an ATWS. This is one of the first steps in EMG E-0 which peer checks are suspended for so it will not slow down the mitigation strategy. Plausible is the student thinks of this as a reactivity change and peer checks are required for it except during EMGs Incorrect - placing the main feed reg valve in manual. During OFNs peer checks are suspended until the plant is stable. Plausible if the student confuses this with EMG guidance.

Incorrect - hanging a clearance order tag in a high rad area. Plausible since the procedure directs this as a shall except in a high rad area.

Meets KA because a peer check is required for all planned reactivity changes which is a guideline for reactivity management at WC RO knowledge conduct of ops peer checks when required Rev 0

Question 68 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98383 User-Defined ID: 98383

Reference:

AP 21-001 Topic: 68 RO peer check required for planned reactivity changes RO Importance Rating: 4.3 SRO Importance Rating: 4.6 K/A Number: 2.1.37 Comments: NEW Lesson Plan Objective:

Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 10 KA - Conduct of ops - Knowledge of procedures, guidelines, or limitations associated with reactivity management Modification History:

0 - replaced based on Scotts feedback 4/25/15 1 - replaced based on validation comments 8/19/15 Rev 0

69 ID: 98384 Points: 1.00 Given the following RCS leak rate data at 100% power:

Total RCS leak rate is 6.9 gpm Leakage into the PRT is 3.2 gpm Leakage into the RCDT is 0.2 gpm

'A' SG tube leakage is 0.08 gpm

'B' SG tube leakage is 0.03 gpm Which one of the following Technical Specification RCS leakage limits has been exceeded?

A. Unidentified Leakage B. Total S/G Leakage C. Primary to Secondary Leakage in 'A' S/G D. Identified Leakage Answer: A Answer Explanation:

TS 3.4.13 Correct - 6.9 - (3.2 + .2 + .08 + .03) = 3.39 gpm unidentified leakage. Leakage to the PRT and RCDT is identified leakage since it can only come from a given source.

Incorrect - identified. All the identified leakage adds up to 3.51 gpm so this is less than the 10 gpm allowed by TS. Plausible if the student confuses TS values.

Incorrect - total SG. There is no limit for total SG leakage just 150 gpd for each SG but the two SG are leaking 158.4 gpd total. Plausible if the student confuses TS values.

Incorrect - primary to sec. .08 gpm is 115.2 gpd less than the limit. .03 is 43.2 gpd less than the limit. Plausible if the student confuses TS values.

Meets KA asks knowledge of TS (conditions of license) for equipment control RO knowledge TS above the double line Rev 0

Question 69 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98384 User-Defined ID: 98384

Reference:

T.S. 3.4.13 Topic: 69 RO TS operational leakage RO Importance Rating: 3.6 SRO Importance Rating: 4.5 K/A Number: 2.2.38 Comments: BANK - ANO Lesson Plan Objective: LO1732701 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 3 Group #

Last Used - 2011 ANO Comprehension 55.41 part 10 KA - Equipment control - Knowledge of conditions and limitations in the facility license Rev 0

70 ID: 98385 Points: 1.00 A NPIS computer point is coming in alarm. The RO has investigated and determined that the point is NOT in alarm and the NPIS point has malfunctioned.

Which of the following actions will the crew perform for removing / tracking this malfunctioning NPIS computer point?

A. Pull the annunciator card, place an OOS sticker on the NPIS terminal B. Pull the annunciator card, delete the analog point from alarm processing C. Make an Equipment Out-Of-Service log entry, place an OOS sticker on the NPIS terminal D. Make an Equipment Out-Of-Service log entry, delete the analog point from alarm processing Answer: D Answer Explanation:

Correct - Per AP 21F-001, EQUIPMENT OUT OF SERVICE CONTROL, since this is a NPIS point only a log entry and the deletion of the alarm from processing is all that is required.

Incorrect - Make an out of service entry and place an OOS sticker on the NPIS terminal. The procedure specifically states NOT to put a sticker on alarm points.

Plausible if the student confuses the type of alarm that is being deleted.

Incorrect - Pull the card and delete the point. This is a NPIS point and cards are associated with annunciators. Plausible if the student confuses the type of alarm that is being deleted.

Incorrect - Pull the card and place an OOS sticker on the NPIS terminal. This is a NPIS point and cards are associated with annunciators. The procedure states NOT to place an OOS sticker on alarm points. Plausible if the student confuses the type of alarm that is being deleted.

Meets KA asks for knowledge of the process of tracking inop alarms per procedure RO knowledge admin for how to document an inoperable alarm Rev 0

Question 70 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98385 User-Defined ID: 98385

Reference:

AP 21F-001 Topic: 70 RO tracking inoperable alarms RO Importance Rating: 3.0 SRO Importance Rating: 3.3 K/A Number: 2.2.43 Comments: NEW Lesson Plan Objective: LO1733213 R6, DISCUSS the operator responsibilities assigned by procedure AP 21F-001.

Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 10 KA - Equipment control - Knowledge of the process used to track inoperable alarms Rev 0

71 ID: 98386 Points: 1.00 A Wolf Creek employee with a current Form NRC-4 record needs to perform work in an area with general radiation levels of 75 mR/hr. The worker has NOT received any exposure today.

The worker's exposure history is:

Lifetime: 24.5 Rem Year to date: 1400 mR Current quarter 225 mR Which one of the following is the MAXIMUM time that the worker can work in the area without exceeding any Administrative exposure limits? (Assume NO special authorization.)

A. 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> B. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C. 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> and 20 minutes D. 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> and 40 minutes Answer: B Answer Explanation:

Correct - 2000 mR - 1400 mR = 600 mR / 75 mR = 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Incorrect - 3000 mR - 1400 mR = 1600 mR / 75 mR = 21.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. AP 25B-100 allows for up to 3000 mR if some is from another site. Plausible if the student confuses limits.

Incorrect - 2000 mR - 1400 mR - 225 mR = 375 mR / 75 mR = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The AP allows for up to 4000 mR with approval. Plausible if the student confuses limits.

Incorrect - 5000 mR -1400 mR = 2600 mR / 75 mR = 34.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. The 10 CFR 20 limits are higher. Plausible if the student confuses limits.

Meets KA asks knowledge of exposure limits under normal conditions.

RO knowledge admin procedure exposure limits is every rad workers responsibility Rev 0

Question 71 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98386 User-Defined ID: 98386

Reference:

AP 25B-100 Topic: 71 RO exposure limits under normal conditions RO Importance Rating: 3.2 SRO Importance Rating: 3.7 K/A Number: 2.3.4 Comments: BANK - 22235 Lesson Plan Objective: LO1733204 R1, DISCUSS the requirements of procedure AP 25B-100, Radiation Worker Guidelines as pertaining to the responsibilities of rad.

workers, exposure limits, and contamination controls.

Tier # 3 Group #

Last Used - N/A Comprehension 55.41 part 12 KA - Rad control - Knowledge of radiation exposure limits under normal or emergency conditions Rev 0

72 ID: 98387 Points: 1.00 Which of the following evolutions will raise radiation levels in the local area?

A. Starting up the S/G Blowdown system B. Batching a new batch of boric acid C. Adding hydrogen peroxide to the RCS D. Makeup to the BL tank (TBL01, RMWST)

Answer: C Answer Explanation:

Correct - SYS BG-207 discusses that HP must be aware that this is being done so room can be high rad.

Incorrect - SG blowdown. This water is not contaminated so this will not change rad levels. Plausible if the student confuses what water is contaminated and which is not.

Incorrect - makeup to BL tank. This will only add pure water to a tank that makes up to the RCS water. Plausible if the student confuses what water is contaminated and which is not.

Incorrect - batching boric acid. This is adding water from the RMWST to the batch add tank and adding boric acid bags. There is no caution in SYS BG-206 for changing rad levels for this evolution. Plausible if the student doesn't understand how to batch acid.

Meets KA asks for knowledge of changing rad levels while performing normal activities RO knowledge system understanding Rev 0

Question 72 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98387 User-Defined ID: 98387

Reference:

SYS EJ-110A Topic: 72 RO Changing rad levels for various evolutions RO Importance Rating: 3.4 SRO Importance Rating: 3.8 K/A Number: 2.3.14 Comments: NEW Lesson Plan Objective: LO1733204 R1, DISCUSS the requirements of procedure AP 25B-100, Radiation Worker Guidelines as pertaining to the responsibilities of rad.

workers, exposure limits, and contamination controls Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 12 KA - Rad control - Knowledge of rad or contamination hazards that may arise during normal abnormal or emergency conditions or activities Modification History:

Rev 0

73 ID: 98388 Points: 1.00 An earthquake has impacted the Wolf Creek plant. Extensive damage is observed. The operating crew notes the following conditions / indications:

Offsite power Unavailable Time since trip 40 minutes RCS pressure 2345 psig CETCs 715°F PZR level 0%

ECCS Flow indicated Rod bottom lights All lit IR SUR + 0.1 DPM RVLIS NC range 42%

S/G NR levels A 7%

B 3%

C 3%

D 5%

AFW flow 272,000 lbm/hr Containment normal sump 2004' 2" Which of the following CSFSTs is the highest priority?

A. Subcriticality B. Core Cooling C. Heat Sink D. Containment Answer: B Answer Explanation:

Current conditions show:

subcriticality - orange due to + SUR on IR core cooling - red due to low RVLIS and CETC heat sink - yellow due to not all levels greater than 6%

integrity - green due to a less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cooldown but all temps greater than 270°F containment - orange due to sump level being higher than 2003' 11" inventory - yellow due to PZR level lower than 17%

Correct - This is the only red path based on EMG F-0 Incorrect - subcriticality. This path is orange. Plausible if the student misdiagnoses the event.

Incorrect - containment. This path is orange. Plausible if the student misdiagnoses the event.

Incorrect - heat sink. This path is yellow. Plausible if the student misdiagnoses the event.

Rev 0

Meets KA asks if they can use indications, determine if entry conditions are met, and then prioritize which procedure takes priority over others (knowledge of procedure organization for emergency)

RO knowledge entry into red and orange path is RO and priority of them Question 73 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98388 User-Defined ID: 98388

Reference:

EMG F-0 REV 17 Topic: 73 RO status tree determination RO Importance Rating: 3.7 SRO Importance Rating: 4.3 K/A Number: 2.4.5 Comments: NEW Lesson Plan Objective: LO1732338 R3, DISCUSS the major action steps of procedure EMG F-0.

Tier # 3 Group #

Last Used - N/A Comprehension 55.41 part 10 KA - Emergency procedures - Knowledge of the organization of the operating procedures network for normal abnormal and emergency evolutions Modification History:

Rev 0

74 ID: 98389 Points: 1.00 Given the following plant conditions:

Plant is in MODE 4 RCS pressure is 340 psig RHR pump flow begins oscillating between 2200 and 3500 gpm The crew has entered OFN EJ-015, LOSS OF RHR COOLING Which of the following describes the required action for the crew if this condition continues?

A. Open BN HV-8812A, RWST outlet to 'A' RHR pump B. Stop the affected RHR pump C. Open FCV-618, RHR HX A BYPASS CTRL valve D. Throttle open EJ-V033, 'A' train CCW Heat Exchanger outlet valve Answer: B Answer Explanation:

Correct - per OFN EJ-015 foldout page if RHR pump flow is cycling over 1000 gpm then stop the RHR pump Incorrect - open EJ-V033. This action is discussed in this procedure but not for this condition. Plausible if the student thinks lowering RHR outlet temperature will correct the cavitation.

Incorrect - open BN HV-8812A RWST suction. Plausible as the student thinks this would raise suction pressure but since RCS pressure is 340 psig this would cause water to go from the RCS to the RWST Incorrect - open FCV-618. Plausible if the student confuses how this valve works and thinks this will lower flow, but this will raise flow.

Meets KA knowledge of RHR shutdown implications with loss of RHR and mitigation strategies. Even though this is a Tier 3 KA question, the KA is specific to the conditions presented in the stem.

RO knowledge foldout page item Rev 0

Question 74 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98389 User-Defined ID: 98389

Reference:

OFN EJ-015 Topic: 74 RO actions for RHR pumps on cavitation RO Importance Rating: 3.8 SRO Importance Rating: 4.2 K/A Number: 2.4.9 Comments: NEW Lesson Plan Objective: LO1732425 R4, Given a procedure flow path, EXAMINE the available options for procedure OFN EJ-015 actions.

Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 10 KA - Emergency procedures - Knowledge of low power shutdown implications in accident, LOCA or loss of RHR, mitigation strategies Rev 0

75 ID: 98390 Points: 1.00 Wolf Creek has declared a Site Area Emergency. You have been assigned the duties of the ENS communicator.

Which of the following items is required to be reported to the NRC over the ENS phone line per EPP 06-001, CONTROL ROOM OPERATIONS?

A. Terminating the event.

B. Dispatching a repair team into Containment C. Stopping ECCS pumps D. S/G ARV lifting at setpoint Answer: A Answer Explanation:

7.4.3 Provide the following additional information to the NRC:

1. Any further degradation in the level of safety of the plant or other worsening plant conditions

2. Any change from one emergency class to another

3. Termination of an emergency class

4. The results of ensuing evaluations or assessments of plant conditions

5. The effectiveness of response or protective measures taken

6. Any information related to plant behavior that is not understood by the NRC Correct - This is stated in the EPP as an item to inform the NRC of.

Incorrect - stopping of ECCS pumps. This is per procedure and not a degradation of the plant. Plausible if the student don't understand the required items from the EPP to report.

Incorrect - SG ARV lifting. This is not a degradation in the level of safety since it is lifting at setpoint. Plausible if the student don't understand the required items from the EPP to report.

Incorrect - dispatching of a repair team. Plausible if the student don't understand the required items from the EPP to report.

Rev 0

Question 75 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98390 User-Defined ID: 98390

Reference:

EPP 06-001 Topic: 75 RO ENS communicator reporting requirements RO Importance Rating: 3.9 SRO Importance Rating: 3.8 K/A Number: 2.4.39 Comments: NEW Lesson Plan Objective: LO1733201 R4, EXPLAIN the duties and responsibilities of the Reactor Operator IAW AP 17C-007 Tier # 3 Group #

Last Used - N/A Comprehension 55.41 part 10 KA - Emergency procedures - Knowledge of RO responsibilities in emergency plan implementation Rev 0

76 ID: 98391 Points: 1.00 The unit was at 100% power when the following indications are observed:

Power Stable RCS pressure Down fast SI occurs Tave Stable PZR level Down fast Containment pressure Up slow Containment humidity Up slow S/G pressure Stable Using the attached reference, based on the given information on what tree will the classification be made on AND why?

A. EAL-4, MAIN STEAM LINE BREAK based on SI actuated ONLY.

B. EAL-3, LOSS OF REACTOR COOLANT BOUNDARY based on SI actuated ONLY.

C. EAL-4, MAIN STEAM LINE BREAK based on Containment pressure and humidity AND SI actuated.

D. EAL-3, LOSS OF REACTOR COOLANT BOUNDARY based on apparent RCS leakage AND SI actuated.

Answer: D Answer Explanation:

Correct - EAL-3, 1,2,3,5,6,7 Alert. A LOCA pressure and level will lower fast. A steam break removes more heat from the coolant and steals some from the main turbine. The extra steam flow also raises reactor power. In this case power is constant and pressure is lowering so this is a LOCA Incorrect - EAL-4, SI actuated ONLY. A steam break will cause an SI but with power and Tave stable this is not the case here. Plausible if the student misdiagnoses this event.

Incorrect - EAL-3, SI actuated ONLY. A steam break will cause an SI based on the cooldown alone. Plausible if the student misdiagnoses this event.

Incorrect - EAL-4, Containment pressure and humidity AND SI actuated. Steam break will lower RCS pressure but it also lowers Tave. Plausible if the student misdiagnoses this event.

Meets KA asks for ability to interpret differences in overcooling (steam break) and LOCA (loss of coolant only)

SRO knowledge classify an event Rev 0

Question 76 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98391 User-Defined ID: 98391

Reference:

APF 06-002-01 Topic: 76 SRO EAL classification tree usage and why RO Importance Rating: 3.7 SRO Importance Rating: 3.7 K/A Number: 011 EA 2.13 Comments: Handout provided NEW Lesson Plan Objective: LO1733215, R1, DISCUSS how to classify an event IAW EPP 06-005.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Large break LOCA - Ability to determine or interpret the following as they apply to a large break LOCA -

Difference between overcooling and LOCA indications -

Safety function Modification History:

Rev 0

77 ID: 98392 Points: 1.00 Given the following:

Unit is shutdown for a refueling outage RCS temperature 195°F down slow

'B' RHR pump is providing shutdown cooling Containment equipment hatch is open for equipment load in

'A' RHR pump breaker has been removed from service for clean and inspect Fire in the 'D' CCW pump causes significant damage to that pump AND the 'B' CCW pump Fire brigade has the fire OUT Which of the following procedures will the SRO direct based on priority?

A. OFN BB-031, SHUTDOWN LOCA B. OFN EJ-015, LOSS OF RHR COOLING C. OFN EG-004, CCW SYSTEM MALFUNCTIONS D. OFN EC-046, FUEL POOL COOLING AND CLEANUP MALFUNCTIONS Answer: B Answer Explanation:

Correct - OFN EJ-015. This procedure contains action that will protect the RHR pumps and establish containment closure. This procedure is the highest priority based on conditions given even though other entry conditions are met for other procedures they will not correct the issue.

Incorrect - OFN BB-031. Plausible if the student confuses the loss of CCW (thermal barrier cooling) with loss of seal injection which would lead to a LOCA through the seals. Also this procedure can only be entered if the plant is in mode 3 after accumulators are isolated or mode 4.

Incorrect - OFN EG-004. Plausible if the student thinks the loss of CCW is higher priority than the loss of RHR. This procedure will not correct the loss of RHR.

Incorrect - OFN EC-046. Plausible if the student understands that the loss of CCW will lead to a loss of fuel pool cooling and thinks this is higher priority than loss of RHR.

Meets KA asks procedure usage in mode 5 with a loss of RHR cooling SRO knowledge asks for specific procedure mitigation and has the SRO prioritize which procedure will need to be used to correct the larger issue not just symptoms of the issue (loss of RHR)

Rev 0

Question 77 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98392 User-Defined ID: 98392

Reference:

OFN EJ-015 Topic: 77 SRO loss CCW leads to a loss of RHR RO Importance Rating: 4.3 SRO Importance Rating: 4.4 K/A Number: 025 2.1.23 Comments: NEW Lesson Plan Objective: LO1732425 R3, Given a procedure flow path, EXAMINE the available options for procedure OFN EJ-015 actions.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Loss of RHR - Conduct of ops - Ability to perform specific system and integrated plant procedures during all modes of plant operation - Safety function 4 Modification History:

0 - replaced based on Scotts feedback 4/25/15 1- replace based on Scotts feedback 6/18/15 Rev 0

78 ID: 98393 Points: 1.00 What is the T.S. limit for RCS DOSE EQUIVALENT I-131 AND what design base accident is this limit based on?

A. 60 uCi/gm AND LOCA B. 60 uCi/gm AND SGTR C. 500 uCi/gm AND SGTR D. 500 uCi/gm AND LOCA Answer: B Answer Explanation:

Correct - This limit is based on a SGTR event and release of steam through a SG ARV.

Incorrect - less than 500 uCi/gm and SGTR. Correct event wrong limit. Plausible as this limit is in the same TS as the iodine limit but this is for XE-133.

Incorrect - less than 60 uCi/gm and LOCA. Correct limit but wrong accident.

Plausible as a LOCA will release activity into the containment and you would want this to be as small as possible.

Incorrect - less than 500 and LOCA. Plausible based on LOCA will release activity into containment and the value is that of XE-133 from same TS.

Meets KA because asks for LCO and limits with regard to SGTR event as the design bases accident SRO because asks for TS bases knowledge and limits below the double line in TS Rev 0

Question 78 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98393 User-Defined ID: 98393

Reference:

TS 3.4.16 78 SRO dose equivalent I-131 limit and what this is based Topic:

on RO Importance Rating: 4.0 SRO Importance Rating: 4.7 K/A Number: 038 2.2.22 Comments: NEW Lesson Plan Objective: LO1732708 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 1 Group # 1 Last Used - N/A Fundamental 55.43 part 2 KA - SGTR - Equipment control - Knowledge of limiting conditions for operations and safety limits - Safety function 3

Rev 0

Q79, Rev 1 ID: 98394 Points: 1.00 Given the following with the plant at 100% power:

• Fuse disconnect NK0311, Feeder to NN13 from NK03 opened

• A loss of bus NN03 occured

• The crew stabilized the plant

• NN03 is now energized from NN15 on the bypass transformer via the static switch Based on the information given:

1. What procedure will the crew perform to correct this issue?

2. What is the OPERABILITY status of NN03?

A. 1. OFN NN-021, LOSS OF VITAL 120 VAC INSTRUMENT BUS

2. INOPERABLE B. 1. OFN NK-020, LOSS OF VITAL 125 VDC BUS NK01, NK02, NK03 AND NK04

2. INOPERABLE C. 1. OFN NN-021, LOSS OF VITAL 120 VAC INSTRUMENT BUS

2. OPERABLE D. 1. OFN NK-020, LOSS OF VITAL 125 VDC BUS NK01, NK02, NK03 AND NK04

2. OPERABLE Answer: A Answer Explanation:

We have new instrument inverters. The swing inverters can be powered from an NK (DC) source or NG (AC) source and per TS any inverter suppling an NN bus must be powered from an NK source.

Correct - Since the statement in the stem for the power source to NN15 (bypass transformer) which is NG01, this is not considered an operable line up for the instrument buses in this mode Incorrect - OFN NK-020, inoperable. This procedure is plausible since it does cover issues dealing with the NK, safeguards batteries. The stem states that a loss of power from the NK source happened so the NN was lost due to the loss of the NK. The OFN NK only deals with actual battery issues not the fuse disconnects on the busses. The inoperable part is correct.

Incorrect - OFN NN-021, operable. Correct procedure. NN15 on the bypass transformer is not operable as the power supply currently is NG01 and not an NK source per TS.

Plausible if the student is not familiar with the power supplies to the swing NN inverters and thinks this lineup is OK per TS.

Incorrect - OFN NK-020, operable.This procedure is plausible since it does cover issues dealing with the NK, safeguards batteries. The stem states that a loss of power from the NK source happened so the NN was lost due to the loss of the NK. The OFN NK only deals with actual battery issues not the fuse disconnects on the busses. Also the stem states NN15 is powered from the bypass transformer. Plausible if the student is not familiar with the power supplies to the swing NN inverters and thinks this lineup is OK per TS. Since there is two power supplies to the swing inverters the student must know which one is correct to satisfy TS OPS INITIAL NRC Page: 1 of 2 14 October 2015

Meets the K/A asks for entry conditions for abnormal operating procedures for the instrument buses SRO knowledge operability calls are SRO job function Question Q79, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98394 User-Defined ID: 98394

Reference:

OFN NN-021 79 SRO operabliity of NN03 with NN15 suppling on the Topic:

bypass RO Importance Rating: 4.5 SRO Importance Rating: 4.7 K/A Number: 057 2.4.4 Comments: NEW Lesson Plan Objective: SY1506300 R5, Explain the relationship between Technical Specifications and the Class 1E 125V DC and Class 1E 120V AC power systems at the level of detail expected for the job position.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 2, 5 KA - Loss of vital AC instrument bus - Emergency procedures - Ability to recognize abnormal indications for system operating parameters that are entry level conditions for emergency and abnormal operating procedures - Safety function 6 Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 1: ??

Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q80, Rev 1 ID: 98395 Points: 1.00 The plant is operating at 62% power while maintenance repairs the 'B' MFP. The following indications are observed by the crew: (items listed in a time line format)

• Letdown flow Lowers to 0 gpm

• VCT level 35% and lowering

• Charging flow 75 gpm and lowering

• PZR level 51.3 % and rising

• PZR pressure 2285 psig and rising

• PZR sprays Closed

• Alarm 032D, PZR LEV DEV HTRS ON, is in

• Alarm 064C, RCS WR PRESS HI, is in

• Alarm 033B, PZR HI PRESS DEV is in

• Alarm 035B, PORV OPEN, is in What event is in progress AND what procedure actions will the crew take?

A. Select away from all RED train channels B. Place excess letdown in service C. Select alternate PZR level channel for control D. Place PZR master pressure controller in manual and control pressure Answer: B Answer Explanation:

If the line downstream of KA HIS-29 which is the instrument air supply line to all of containment fails then a loss of air to containment will happen. Every air operated valve in containment will go to is failure position over time. RCS pressure will rise because all letdown is isolated and PZR sprays fail closed so PZR level will be rising and when level is 5% over program the backup heaters will turn on due to a possible insurge into the PZR.

This will cause pressure to rise to the PORV setpoint and they will open to control pressure.

Correct - Loss of air to containment so place excess letdown in service. This procedure has steps to place excess letdown in service and control seal injection control PZR level.

Incorrect - PZR level channel failure select alternate level channel for control. Plausible as the student could interpret the indications given as a level channel failure which the procedure would direct the crew to select away from the failed channel.

Incorrect - PZR pressure channel failure place master controller in manual and control pressure. Plausible as the student could interpret the indications given as a pressure channel failure and the procedure would direct the crew to take manual control of the master PZR pressure controller and control pressure.

Incorrect - Loss of NN bus select away from all red train channels. Plausible as the student could interpret the indications given as a loss of the red train instrument bus and the procedure would direct the crew to select away from all red train equipment.

Meets KA because asks for loss of air and failure understanding of components SRO only because asks to identify the correct event in progress (RO) and then what steps in the procedure to mitigate the event (SRO)

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q80, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98395 User-Defined ID: 98395

Reference:

OFN KA-019 80 SRO loss of air to containment affects and procedure Topic:

actions RO Importance Rating: 2.9 SRO Importance Rating: 3.3 K/A Number: 065 AA 2.08 Comments: NEW Lesson Plan Objective: LO1732429 R2, RECOGNIZE the available situations which are addressed by OFN KA-019.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Loss of IA - Ability to determine and interpret the following as they apply to the loss of IA - Failure modes of air operated equipment - Safety function 8 Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

81 ID: 98396 Points: 1.00 The switchyard is split East and West due to grid instability which has been reported to Wolf Creek by the TSO. As the grid instability continues to get worse indicated voltage on NB01 is now 3740 and down slow.

If the above indications continue for the next 2 minutes which of the following statements is correct with regards to NE01 AND offsite circuit operability?

A. NE01 is carrying NB01 All offsite circuits are OPERABLE B. NE01 is carrying NB01 One offsite circuit is INOPERABLE C. NE01 is in STBY All offsite circuits are OPERABLE D. NE01 is in STBY One offsite circuit is INOPERABLE Answer: B Answer Explanation:

With the east and west buses split NB01 is being powered from the east bus and NB02 is being supplied by the west (startup transformer). NB01 is a 4160 V bus and 90% of that is 3744 V so anything less than that will start the degraded voltage timers to start counting down and after a total of 119 seconds will trip both the normal and alternate feeder breakers to NB01 in this case. This starts NE01 and it closes onto the bus. This doesn't make NE01 inoperable since it is performing its safety function. The offsite circuit is inoperable since the voltage supplied to the site is now too low per TS 3.8.1 and bases.

Correct - When less than 3744 V is reached the 119 second timer starts and NE01 will then start on NB01 undervoltage and take the bus. The offsite circuit is inoperable because it can't carry the NB bus. The diesel is operable because it is performing its safety function.

Incorrect - NE01 running and offsite operable. The first part is correct. The offsite circuit is not operable since it indicates low voltage by NE01 starting on degraded voltage. Plausible if the student doesn't understand the switchyard and NB01 relationship with operability.

Incorrect - NE01 in STBY offsite inoperable. The diesel is running at this point on undervoltage on NB01 due to the degraded voltage signal opening the normal and alternate feeder breakers. Plausible if the student forgets the 90% degraded voltage setpoint or time relay.

Incorrect - NE01 in STBY and offsite operable. The diesel is running at this point on undervoltage on NB01 due to the degraded voltage signal opening the normal and alternate feeder breakers. The offsite circuit is not operable since it indicates low voltage by NE01 starting on degraded voltage. Plausible if the student doesn't understand the switchyard to NB01 relationship with operability and forgets the 90% degraded voltage setpoint or time relay.

Rev 0

Meets KA because interrelates the offsite sources with the ESF switchgear SRO because asks for an operability determination for the offsite source Question 81 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98396 User-Defined ID: 98396

Reference:

OFN AF-025 81 SRO degraded grid voltage and NE01 and offsite Topic:

operability RO Importance Rating: 3.6 SRO Importance Rating: 4.0 K/A Number: 077 AA 2.07 Comments: NEW Lesson Plan Objective: SY1406401 R3, Assess the functional interrelationship with the AC Distribution System.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 2 KA - Generator voltage and electric grid disturbances -

Ability to determine and interpret the following as they apply to generator voltage and electric grid disturbances -

Operational status of engineered safety features - Safety function 6 Rev 0

82 ID: 98397 Points: 1.00 The unit is in MODE 3 heating up after a refueling outage. SR NIs are indicating:

SE NI-31B 5 X 101 CPS SE NI-32B 5 X 101 CPS

'A' bank shutdown rods are pulled in preparation for a reactor startup. SR NIs are now indicating:

SE NI-31B 6 X 101 CPS SE NI-32B 6 X 101 CPS Five minutes pass with NO other operator action. SR NIs are now indicating:

SE NI-31B 7 X 101 CPS SE NI-32B 8 X 102 CPS Which of the following statements is correct with regard to SR NIs AND what actions are required?

A. SE NI-31B is INOPERABLE Fully insert all rods and then open reactor trip breakers B. SE NI-32B is INOPERABLE Fully insert all rods and then open reactor trip breakers C. SE NI-31B is INOPERABLE Open reactor trip breakers IMMEDIATELY D. SE NI-32B is INOPERABLE Open reactor trip breakers IMMEDIATELY Answer: B Answer Explanation:

Correct - Per TS 3.3.1 table 2 SR NI are required in mode 3, 4, or 5. With one inoperable then action must be taken to restore both to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OR rods must be inserted and reactor trip breakers opened within 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br />. Channel checks that they both agree within 1 decade. Count rate should not have changed by a full decade with just a single bank pull.

Incorrect - SE NI-32B inoperable open trip breakers immediately. First part is correct. The procedure has the rods inserted and then trip breakers opened.

Plausible if the student confuses TS and the procedure for required actions.

Incorrect - SE NI-31B inoperable open trip breakers immediately. Indication of SR NIs should be close to the same given rods were withdrawn all over in an equal pattern. Plausible if the student thinks that this detector should have raised higher (after any startup rate has decayed off) and it did not. Also the TS do not require the breakers open immediately but after rods have been manually inserted.

Rev 0

Incorrect - SE NI-31B inoperable fully insert all rods. Indication of SR NIs should be close to the same given rods were withdrawn all over in an equal pattern.

Plausible if the student thinks that this detector should have raised higher (after any startup rate has decayed off) and it did not. The second part is correct.

Meets KA asks for ability to determine expected SR count rate change when rods are moved SRO knowledge of procedure actions for the loss of a single SR NI and operability of the detector Question 82 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98397 User-Defined ID: 98397

Reference:

T.S. 3.3.1 Topic: 82 SRO operability of SR NI RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: 032 AA 2.02 Comments: NEW Lesson Plan Objective: LO1732701 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.43 part 2 KA - Loss of SR NI - Ability to determine and interpret the following as they apply to the loss of SR NI - Expected change in source range count rate when rods are moved -

Safety function 7 Modification History:

Rev 0

Q83, Rev 1 ID: 98398 Points: 1.00 Given the following:

• The unit was operating at 100% power

• SGTR in 'A' S/G occurs and the unit is tripped

• SI is actuated

• S/G 'C' is faulted into Containment

• AFW flow CAN NOT be established to any S/G

• Both Containment Mini Purge Exhaust isolation outlet dampers, remain open A _____(1)________ is required to be declared. If an operator is later successful in manually closing at least one Mini Purge damper Containment integrity will be verified by ____(2)______.

(Reference attached)

A. 1. Site Area Emergency

2. CTMT PURGE ISO SYS status light illuminating B. 1. Site Area Emergency

2. Header for CTMT ISO SYS PHASE A, lit solid white C. 1. General Emergency

2. Header for CTMT ISO SYS PHASE A, lit solid white D. 1. General Emergency

2. CTMT PURGE ISO SYS status light illuminating Answer: A Answer Explanation:

Correct - The flow path for the classification is 1, 2, 3, 4, 6, 7, 5, 8 which is a SAE. The evidence that the manual action was successful is the status lights on SA066-X and Y illuminating when the damper is closed.

Incorrect - SAE and Header for CTMT ISO SYS PHASE A, lit solid white. The classification is correct. Plausible if the student forgets that these status lights on the ESFAS panels will come on to show isolation was successful. Also closing the mini purge valve doesn't bring in this white header it brings in the CPIS header Incorrect - GE and Header for CTMT ISO SYS PHASE A, lit solid white. The classification is wrong but if the student doesn't follow the correct path from box 7 to 5 then 8 and just goes to 5 this will get a GE. Plausible if the student forgets that these status lights on the ESFAS panels will come on to show isolation was successful or gets on the wrong path in the trees. Also closing the mini purge valve doesn't bring in this white header it brings in the CPIS header Incorrect - GE and containment status lights on. The classification is wrong but if the student doesn't follow the correct path from box 7 to 5 then 8 and just goes to 5 this will get a GE. The second part is correct. Plausible if the student gets on the wrong path on the trees.

Meets KA because asks to classify based on containment integrity and then after it is achieve what indication will prove you have it SRO because classification job is SRO OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q83, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98398 User-Defined ID: 98398

Reference:

APF 06-002-01 83 SRO site area emergency and containment mini purge Topic:

valves open RO Importance Rating: 3.9 SRO Importance Rating: 4.4 K/A Number: 069 AA 2.02 Comments: Provided reference - EAL sheets MODIFIED - Prairie Island Lesson Plan Objective: LO1733215 R1, DISCUSS how to classify an event IAW EPP 06-005.

Tier # 1 Group # 2 Last Used - 2010 Prairie Island #88 Comprehension 55.43 part 1, 5 KA - Loss of containment integrity - Ability to determine and interpret the following as they apply to the loss of containment integrity - Verification of auto and manual means of restoring integrity - Safety function 5 Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

84 ID: 98399 Points: 1.00 The crew has entered EMG ES-01, REDIAGNOSIS.

Given the following indications prior to and after a reactor trip and SI:

Parameter Just prior to Rx trip After Rx trip and SI RCS pressure Down fast Down slow Tave Stable Down slow PZR level Down fast Up fast SI Standby/Armed Actuated Rx power Down slow Tripped Main generator power Down slow Tripped CTMT pressure Stable Up slow S/G levels ALL stable ALL up slow S/G pressures ALL stable ALL down slow NB01 Energized Locked out NB02 Energized Energized MSIVs ALL open ALL closed What procedure will the SRO transition to from EMG ES-01?

A. EMG C-0, LOSS OF ALL AC POWER.

B. EMG E-3, STEAM GENERATOR TUBE RUPTURE.

C. EMG E-2, FAULTED STEAM GENERATOR ISOLATION.

D. EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

Answer: D Answer Explanation:

The parameters given were taken from the desktop simulator for a large PZR steam space leak. The general nature of what is given is all that is needed in ES-01 to make decisions as to where to transition to so no specific values are given.

Correct - Since E-0 must have been exited prior to coming to ES-01 the first transition out is not used. E-1 is correct since PZR level is rising and no faulted SG was found. This is the last transition out of ES-01 in the RNO column Incorrect - E-3. With levels rising in all SGs this could be thought of as a rupture.

This is from the normal AFW flow that will actuate on the reactor trip and subsequent shrink of the SGs. Plausible if the student doesn't understand all the auto actuation that will take place on the trip.

Incorrect - C-0. With a loss of one emergency bus and having and OFN and an EMG for loss of power this transition could be made. Plausible if the student fails to see that only one bus is locked out and it only takes one bus to complete mitigation actions in EMG procedures.

Rev 0

Incorrect - E-2. With SG pressure slowly lowering the student could mistake the cooldown from the ECCS pumps injecting and SG pressure lowering with a fault.

Plausible if the student fails to think about the cooldown from ECCS.

Meets KA because asks for understanding of steps in rediagnosis SRO because procedure selection based on indications and involves a procedure transition in the EOPs Question 84 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98399 User-Defined ID: 98399

Reference:

EMG ES-01 84 SRO rediagnosis and transition to E-1 from PZR steam Topic:

space leak RO Importance Rating: 4.6 SRO Importance Rating: 4.6 K/A Number: E01 2.1.20 Comments: NEW Lesson Plan Objective: LO1732314 R4, EXPLAIN the bases and knowledge requirements for selected procedure steps.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.43 part 5 KA - Rediagnosis / SI termination - Conduct of ops - Ability to interpret and execute procedure steps - Safety function 3 Rev 0

85 ID: 98400 Points: 1.00 Given a LOCA inside Containment, which of the following is the correct procedure to enter AND why?

(use the attached indications for decision basis)

A. EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE, Red path.

Based on containment pressure.

B. EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE, Orange path.

Based on only one containment spray pump running.

C. EMG FR-Z2, RESPONSE TO CONTAINMENT FLOODING, Orange path.

Based on containment sump level.

D. EMG FR-Z3, RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL, Yellow path.

Based on containment rad levels.

Answer: C Answer Explanation:

Correct - With containment adverse and the sump level high this is the correct procedure to enter. There is only ONE procedure at Wolf Creek that deals with containment flooding and it is an orange path FR-Z2.

This procedure contains 4 steps and one is return to procedure step in effect. Indications given for this question make the student determine the status of the containment spray pumps and the fact that the CS pumps were started by the LOCA sequencer and not by the crew. This interpretation is needed to justify distractors. Then the student must determine that containment is adverse to get to the correct procedure entry.

Incorrect - FR-Z3 yellow. Conditions are met to enter this procedure on a yellow path but since the orange FR-Z2 is met procedure use and adherence has the higher level procedure enter first. Plausible since this is a correct entry but not with the orange path also met.

Incorrect - FR-Z1 red. Plausible since containment pressure is well above normal but the red path entry would be over 60 psig.

Incorrect - FR-Z1 orange. Plausible since there is one spray pump not running but this entry would be based on two spray pumps stopped or 'as least one running' per procedure.

Meets KA The student has to evaluate/determine if indications are correct for the current plant condition from the attached graphic depicting MCB switch positions, equipment status, and meter values and goes on to ask what procedure needs to be entered based on those indications SRO knowledge procedure selection based on conditions given Rev 0

Question 85 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98400 User-Defined ID: 98400

Reference:

EMG F-0 Topic: 85 SRO FR-Z entry for flooding RO Importance Rating: 4.6 SRO Importance Rating: 4.3 K/A Number: E15 2.1.31 Comments: NEW Lesson Plan Objective:

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.43 part 5 KA - Containment Flooding - Ability to locate control room switches controls and indications and to determine that they correctly reflect the desired plant lineup - Safety function 5 Rev 0

Rev 0 86 ID: 98401 Points: 1.00 Given the following:

Unit is recovering from a refueling outage The RCP seal injection throttle valves were replaced ALL valves have been throttled (set)

RCS temperature is 410°F Using the attached reference and the given pressures, which of the following is an acceptable seal injection flow to each RCPs AND the basis for that flow restriction?

A. Charging header pressure 2350 psig RCS pressure 2235 psig 9.5 gpm Limit the MAXIMUM flow to the seal to prevent premature seal failure.

B. Charging header pressure 2550 psig RCS pressure 2235 psig 13.8 gpm Limit the MAXIMUM flow to the seal to prevent premature seal failure.

C. Charging header pressure 2550 psig RCS pressure 2235 psig 13.8 gpm Limit the amount of flow that is diverted from the normal injection path in accident conditions.

D. Charging header pressure 2350 psig RCS pressure 2235 psig 9.5 gpm Limit the amount of flow that is diverted from the normal injection path in accident conditions.

Answer: C Answer Explanation:

Correct - The DP here is 315 and the seal injection flow given is within the acceptable region of the provided graph. The reason is out of the basis which is to limit flow diverted away from the injection.

Incorrect - 13.8 gpm, limit seal water flow. The flow number is correct. The reason is not per the TS basis flow diverted away from the normal injection path is the concern. The spec does discuss the flow to the seals to prevent damage but this is not a limit on max flow to the seal but min flow to the seal to prevent failure. Plausible if the student can use the graph but mistakes the reason.

Incorrect - 9.5 gpm, limit amount NOT going to injection. The DP here is 115 so the max would be close to 8.5 gpm. The reason is correct. Plausible if the student mis-uses the graph and understands the basis.

Rev 0

Incorrect - 9.5 gpm, limit seal water flow. The DP here is 115 so the max would be close to 8.5 gpm. The spec does discuss the flow to the seals to prevent damage but this is not a limit on max flow to the seal but min flow to the seal to prevent failure. Plausible if the student mis-uses the graph and mistakes the reason.

Meets KA asks ECCS TS basis SRO knowledge TS basis Question 86 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98401 User-Defined ID: 98401

Reference:

BD TS 3.5.5 Topic: 86 SRO seal flow DP and reason in TS basis RO Importance Rating: 3.2 SRO Importance Rating: 4.2 K/A Number: 006 2.2.25 Comments: Handout provided NEW Lesson Plan Objective: SY1300300 R8, EXPLAIN Technical Specifications associated with the RCPs at the level of detail expected for the job position.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 2 KA - ECCS - Equipment control - Knowledge of the bases in TS for limiting condition for operations and safety limits Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

Q87, Rev 1 ID: 98402 Points: 1.00 Of the following reportable events, which one has a time limit maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />?

A. Wolf Creek deviated from T.S.

B. Wolf Creek declared a Site Area Emergency.

C. Wolf Creek experienced a reactor trip from 10% power.

D. Wolf Creek sent a contaminated / injured person to the Coffey County Hospital for medical treatment.

Answer: C Answer Explanation:

Correct - Per 10 CFR 50.72 and 73 and AP 26A-001 any event that results in a reactor trip when the reactor is critical is reportable within four hours.

Incorrect - SAE. Plausible as this is reportable but within one hour not four.

Incorrect - contaminated / injured person offsite. Plausible as this is reportable within eight hours not four.

Incorrect - Deviation of TS. Plausible since this is reportable within one hour not four.

Meets KA asks by when an event is reportable to the NRC and is associated with an ESFAS signal SRO knowledge reportability is SRO function OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q87, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98402 User-Defined ID: 98402

Reference:

10 CFR 50.72 Topic: 87 SRO reportablity within four hours RO Importance Rating: 2.7 SRO Importance Rating: 4.1 K/A Number: 013 2.4.30 Comments: NEW Lesson Plan Objective: LO1734021 R5, Given initial conditions, determine the reportability requirements to the NRC.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 1 KA - ESFAS - Emergency procedures - Knowledge of events related to system operation status that must be reported to internal organizations or external agencies such as the state the NRC or the transmission system operator Modification History:

0 - modified based on Scotts feedback 4/25/15 1 - revised based on NRC comment 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q88, Rev 1 ID: 98403 Points: 1.00 Given the following:

• Reactor power was 100%

• LOCA occurred

• SI auto actuated

• CTMT pressure peaked at 32 psig

• Crew entered the appropriate procedures

• Annunciator 047D, RWST LEV LOLO 1 AUTO XFR, is received

• Crew transitioned to EMG ES-12, TRANSFER TO COLD LEG RECIRCULATION

• Crew has completed EMG ES-12

• Annunciator 047C, RWST LEVEL LOLO2, is received

• Both RHR pumps are now indicating fluctuating flow and amps

1. What is the current condition of the Containment Spray pumps?

2. What procedure will the crew transition to?

A. 1. Both spray pumps are running

2. EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE B. 1. Both spray pumps are stopped

2. EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE C. 1. Both spray pumps are stopped

2. EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION D. 1. Both spray pumps are running

2. EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION Answer: A Answer Explanation:

Correct - ES-12 has no step to stop containment spray pumps with RWST level greater than 6% (foldout page). It does have the spray pumps aligned to the sump if RWST level is less than 12%. C-13 will be entered based on RHR pump flow.

Incorrect - ES-13 and stopped. ES-12 has no step to stop containment spray pumps with RWST level greater than 6%. It does have the spray pumps aligned to the sump if RWST level is less than 12%. Plausible if the student uses foldout page of ES-12 to stop pumps.

Incorrect - C-11 and running. The running is correct while in ES-12 but a transition to C-11 would only be made if the valves to RHR could not be placed in the correct alignment.

Plausible if the student sees this as a valve problem and not a loss of suction due to blockage.

Incorrect - C-11 and stopped. ES-12 has no step to stop spray pumps with RWST level greater than 6%. Plausible if the student sees this as a valve problem and not a loss of suction due to blockage.

Meets KA asks for procedure usage to correct a loss of containment spray while in recirc mode due to a loss of suction SRO knowledge since this question sets up a time line of events and after the given procedure is complete it asks the status of equipment (SRO knowledge of procedure steps) and then asks what procedure the crew will transition to next (SRO knowledge based on assessment of conditions given).

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q88, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 4.00 System ID: 98403 User-Defined ID: 98403

Reference:

EMG C-13 Topic: 88 SRO transition to C-13 and status of spray pumps RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: 026 A 2.07 Comments: NEW Lesson Plan Objective: LO1732332, R3, SUMMARIZE the major action categories and the bases for the steps that accomplish each category.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Containment spray - Ability to predict the impacts of the following malfunction or operations on the CSS and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Loss of containment spray pump suction when in recirc mode possibly caused by clogged sump screen pump inlet high temperature exceeded cavitation voiding or sump level below cutoff interlock limit Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

89 ID: 98404 Points: 1.00 The unit is operating at 100% power when a reactor trip occurs on S/G low level. 30 seconds later auto SI occurs on containment pressure. The RO then throttles AFW flow per procedure. The following indications are now observed:

Parameter Trip plus 2 minutes Trip plus 5 minutes RCS pressure 1965 psig down fast 1915 psig down slow RCS temperature 544°F down fast 520°F down slow PZR level 23.5% down fast 15.2% down slow CTMT pressure 12 psig up slow 18 psig stable CTMT Rad Normal Normal S/G pressure A 940 psig down slow 800 psig down slow B 936 psig down slow 783 psig down slow C 938 psig down slow 617 psig down fast D 941 psig down slow 798 psig down slow S/G WR level A 42.5% down slow 35.9% down slow B 42.3% down slow 35.3% down slow C 23.4% down slow 6.9% down slow D 41.8% down slow 33.7% down slow S/G steam flow 0.5 X 106 MPPH 0 MPPH A

B 0.5 X 106 MPPH 0 MPPH C 0 MPPH 0 MPPH D 0.5 X 106 MPPH 0 MPPH As the crew works through EMG E-0, REACTOR TRIP OR SAFETY INJECTION:

1) What event is in progress?

2) What procedure will mitigate the event?

3) What actions will be taken during the performance of this procedure?

A. 1) Main Feed Line break in Containment

2) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT

3) 'C' S/G steam flow will be isolated AND ECCS flow WILL be reduced B. 1) Main Steam Line break in Containment

2) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT

3) 'C' S/G AFW flow will be isolated to AND ECCS flow WILL be reduced C. 1) Main Steam Line break in Containment

2) EMG E-2, FAULTED STEAM GENERATOR ISOLATION

3) 'C' S/G AFW flow will be isolated to AND ECCS flow WILL NOT be reduced D. 1) Main Feed Line break in Containment

2) EMG E-2, FAULTED STEAM GENERATOR ISOLATION

3) 'C' S/G steam flow will be isolated AND ECCS flow WILL NOT be reduced Answer: D Rev 0

Answer Explanation:

All data was gathered from running this on the desk top simulator with AFW flow throttled at event initiation, a full feedline break inside containment, and no other operator actions. This would be consistent with working through E-0 since all actions up to diagnostics have no effect on the parameters used here.

Correct - Per E-0 C SG is faulted at step 16 a transition is made to E-2. The diagnosis key is early steam flow from all but C SG and C SG WR level being lower than all the rest with no indicated steam flow which points at the feed line break. At first a feed line break will look like a steam break until the feed ring is uncovered and then the SG pressure will drop rapidly and then the diagnosis can be made. E-2 will isolate all the flow from and feed to the SG. At the end of E-2 the operator is asked if ECCS flow should be reduced based on indications.

Since the RCS pressure is still lowering a transition is made to E-1 and when the SG blows dry and the pressure stabilizes out then ECCS flow can be reduced.

This is the last step of E-2.

Incorrect - steam line break, E-1, AFW iso. The steam line break in containment would drop SG pressure immediately. E-1 would not be the correct procedure entry if there was a main steam line break in containment. AFW will be isolated in E-0. Plausible if the student misdiagnosis this event.

Incorrect - feed line break, E-1, SG C iso. Correct diagnosis but E-1 is not the correct transition. Also ECCS flow will not be reduced since RCS pressure is still lowering. Plausible if the student misdiagnosis the event.

Incorrect - steam line break, E-2, AFW iso. The steam line break in containment would drop SG pressure immediately. E-2 is the correct procedure. AFW will be isolated in E-0 not E-2, only the steam flow from the C SG to the TDAFWP.

Plausible if the student misdiagnosis the event.

Meets KA asks for ability to predict the impact of a main feed line break inside containment and then select the correct procedure to mitigate the event and control the consequences.

SRO knowledge since it asks for specific procedure step actions i.e. not reducing ECCS flow.

Rev 0

Question 89 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 5 Difficulty: 4.00 System ID: 98404 User-Defined ID: 98404

Reference:

EMG E-2 Topic: 89 SRO diagnose feedline break in containment RO Importance Rating: 3.1 SRO Importance Rating: 3.4 K/A Number: 059 A 2.05 Comments: MODIFIED - 16504 Lesson Plan Objective: LO1732324 R4, EXPLAIN the bases and any knowledge requirements for selected procedure steps.

Tier # 2 Group # 1 Last Used - STP 2001 Comprehension 55.43 part 5 KA - Main feedwater - Ability to predict the impacts of the following malfunctions or operations on the MFW system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Rupture in MFW suction or discharge line Modification History:

Rev 0

90 ID: 98405 Points: 1.00 Given the following:

The crew is performing STS KJ-005A, MANUAL/AUTO START, SYNC AND LOADING OF EDG NE01 The RO is ready to close the output breaker for NE01 in order to parallel it with NB01 (off site)

NB01 voltage is indicating 4160 VAC and frequency is indication 60.1 Hz NE01 voltage is indicating 4080 VAC and frequency is indicating 60.3 Hz What action is required to complete the paralleling operation of NE01 to NB01?

Under these conditions, a copy of which procedure is required to be kept with the RO and NSO to keep NE01 OPERABLE during this surveillance run?

A. 1. Lower NE01 frequency.

2. OFN NB-042, LOSS OF OFFSITE POWER TO NB01(NB02) WITH EDG PARALLELED.

B. 1. Lower NE01 frequency.

2. SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR 'A'.

C. 1. Raise NE01 voltage.

2. SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR 'A'.

D. 1. Raise NE01 voltage.

2. OFN NB-042, LOSS OF OFFSITE POWER TO NB01(NB02) WITH EDG PARALLELED.

Answer: D Answer Explanation:

Correct - Per SYS KJ-123 the incoming voltage needs to be within +/- 50 volts to parallel the diesel. STS KJ-005A 4.11.1 states that if individuals and a copy of OFN NB-042 are present during the test the EDG remains operable.

Incorrect - raise voltage copy of SYS. The SYS is only the test procedure after maintenance and the OFN is required at the location and in the control room for the EDG to remain operable. Plausible if the student mistakes the contents of the SYS with the requirement to have the OFN.

Incorrect - lower frequency copy of OFN. Plausible if the student confuses the starting parameter requirements for the EDG. The OFN is correct to have at the locations.

Incorrect - lower frequency copy of SYS. Plausible if the student confuses the starting parameter requirements for the EDG. Also the SYS is only the test procedure after maintenance and the OFN is required at the location and in the control room for the EDG to remain operable. Plausible if the student mistakes the contents of the SYS with the requirement to have the OFN.

Rev 0

Meets KA asks about using procedures to operate the EDG when sync'ing it to other sources SRO only because asks about what it takes to maintain operability of the EDG during the test run Question 90 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98405 User-Defined ID: 98405

Reference:

SYS KJ-123 90 SRO EDG operability when paralleled and initial Topic:

conditions to sync RO Importance Rating: 3.1 SRO Importance Rating: 3.3 K/A Number: 064 A 2.09 Comments: NEW Lesson Plan Objective:

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 2 KA - EDG - Ability to predict the impacts of the following malfunctions or operations on the EDG system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Sync of EDG with other electric power supplies Rev 0

91 ID: 98406 Points: 1.00 Given the following:

Reactor power is 85%

80B, RPI NON URG ALARM, is received DRPI indicates Data A failure 1,2,3 AND GW for ALL rods What is the status of the DRPI system and what is the current accuracy of the system?

A. OPERABLE, +10/-4 B. OPERABLE, -10/+4 C. INOPERABLE, +10/-4 D. INOPERABLE, -10/+4 Answer: A Answer Explanation:

Correct - With the loss of only one power supply per TS the DRPI system goes to half accuracy which is still within the +/- 12 steps the TS asks for. The DRPI system remains operable.

Incorrect - OPERABLE -10/+4. Correct TS call. The accuracy is +10/-4 (opposite common mistake). Plausible if the student confuses the accuracy but knows the TS.

Incorrect - INOPERABLE +10/-4. Accuracy is correct and with a power failure of one DRPI then the panel will show alarms and lights. Plausible if the student knows the accuracy for the power failure but applies TS wrong (common misconception) for only having one power supply available.

Incorrect - INOPERABLE -10/+4. Common misconception for accuracy.

Plausible if the student applies the TS wrong with the loss of power.

Meets KA asks ability to use plant procedures (TS) to control the loss of RPIS power SRO knowledge operability determinations are SRO only job function Rev 0

Question 91 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98406 User-Defined ID: 98406

Reference:

SY1301400, TS 3.1.7 Topic: 91 SRO loss of DRPI A TS operability and accuracy RO Importance Rating: 3.1 SRO Importance Rating: 3.6 K/A Number: 014 A 2.02 Comments: NEW Lesson Plan Objective: SY1301400 R4, Describe how the Digital Rod Position Indication System control board display unit indicates a half accuracy condition and State the rod position accuracy while in this condition.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.43 part 2 KA - 014 A 2.02 - Rod position indication system - Ability to predict the impacts of the following malfunctions or operations on the RPIS and based on those on those predications use procedures to correct control or mitigate the consequences of those malfunctions or operations -

loss of power to the RPIS Rev 0

92 ID: 98407 Points: 1.00 Given the following information with the plant in MODE 1:

1200 A fire has occurred in the Control Room 1203 Spurious equipment has actuated 1205 Control Room evacuation has taken place 1205 Halon has NOT activated for the cable trenches 1225 CRS places all the Isolate/Transfer switches in isolate What procedure will mitigate this event and what will be the MINIMUM classification that will be made?

A. OFN RP-013, CONTROL ROOM NOT HABITABLE Alert B. OFN RP-017, CONTROL ROOM EVACUATION Site Area Emergency C. OFN RP-017, CONTROL ROOM EVACUATION Alert D. OFN RP-013, CONTROL ROOM NOT HABITABLE Site Area Emergency Answer: B Answer Explanation:

Correct - OFN RP-017 is the only procedure that will mitigate this event since this is a fire in the control room. Since the notes for the procedure and the emergency plant state that the isolate switches must be in iso within 15 minutes and the time line shows more than that the SAE is called.

Incorrect - OFN RP-017 and Alert. Correct procedure but since the switches were not placed correctly within 15 minutes the alert is to low. Plausible if the student fails to see the time the switches were moved and just classifies off the notes and the control room evacuation Incorrect - OFN RP-013 and SAE. This procedure will be entered if the control room is evacuated for reasons OTHER than fire so this procedure will not mitigate this event. SAE is correct. Plausible if the student knows the correct classification but misses that a fire has occurred.

Incorrect - OFN RP-013 and Alert. This procedure will be entered if the control room is evacuated for reasons OTHER than fire so this procedure will not mitigate this event. Since the switches were not placed correctly within 15 minutes the alert is too low. Plausible if the student misses the fire and the switches.

Meets KA assumes failure of auto fire protection system and then asks for a procedure to mitigate consequences of this failure.

SRO knowledge asks for classification if control room switches are not isolated within the 15 minute clock.

Rev 0

Question 92 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98407 User-Defined ID: 98407

Reference:

OFN RP-017 92 SRO correct procedure to use with control room evac Topic:

and a classification RO Importance Rating: 3.3 SRO Importance Rating: 3.9 K/A Number: 086 A 2.04 Comments: NEW Lesson Plan Objective: LO1732426 R4, EXPLAIN the basis and any knowledge requirement for selected procedure steps.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.43 part 5 KA - Fire protection - Ability to predict the impacts of the following malfunctions or operations on the fire protection system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Failure to actuate the FPS when required, resulting in fire damage Modification History:

Rev 0

93 ID: 98408 Points: 1.00 The plant is in a refueling outage with the following:

Core off load is in progress with just over half of the assemblies remaining in the vessel The Upender is moving to the vertical position on the Refuel Pool side A fuel assembly is on the hoist and has just been withdrawn from the core Refuel Pool level suddenly begins to lower rapidly The Control Room reports that the 'A' S/G Nozzle Dam has failed Which of the following actions will the Fuel Handling SRO will direct?

A. Lower the fuel assembly into the vessel Send the Fuel Transfer Cart to the SFP B. Place the fuel assembly in the Upender Send the Fuel Transfer Cart to the SFP C. Lower the fuel assembly into the vessel Leave the Fuel Transfer Cart in the Refueling Pool D. Place the fuel assembly in the Upender Leave the Fuel Transfer Cart in the Refueling Pool Answer: A Answer Explanation:

Fuel handling is an SRO only function at Wolf Creek.

Correct - Per the OFN KE-018, FUEL HANDLING ACCIDENT, there are multiple actions the fuel handling SRO will perform. This list is all correct for what is required to completed. The fuel transfer cart must be on the SFP side or the fuel transfer gate valve cannot be closed. The fuel assembly is placed back in the only safe location on the refuel pool side. The transfer tube gate is closed to prevent losing SFP level. The equipment hatch is closed to regain containment integrity for this event.

Incorrect - Place the fuel assembly in the upender, send the cart to the SFP.

Plausible if the student does not understand the safe locations to place the fuel assembly, the rest is correct.

Incorrect - Leave the cart in the refuel pool, ensure the gate is closed. Plausible if the student does not understand that the cart has a cable connected to it that goes back to the SFP side stopping the gate valve from being closed.

Incorrect - leave the cart in the refuel pool, place the fuel assembly in the upender. Plausible if the student does not understand the safe locations to place the fuel assembly or where to put the transfer cart.

Rev 0

Meets KA asks if the student understands how the refueling process goes as far as sequence of events (if the fuel element is on the hoist and just been raised it is still over the vessel, upender is just coming to the upright position must understand where it was and were it is going to know what is next). Knowing what to do in this case is based on the prediction that level will drop far enough to force the action specified so that design limits for fuel cooling and radiation shielding are not exceeded.

SRO knowledge asks for procedure step recall. The actions taken are performed by the refuel SRO inside containment (as directed by the CRS).

Question 93 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98408 User-Defined ID: 98408

Reference:

OFN KE-018 Topic: 93 SRO refuel pool inventory loss fuel handling accident RO Importance Rating: 2.9 SRO Importance Rating: 3.7 K/A Number: 034 A 1.02 Comments: NEW Lesson Plan Objective: LO1732428 R3, Given a procedural flow path, EXAMINE the available options for procedure actions.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.43 part 7 KA - Fuel handling equip - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the fuel handling system controls including - water level in the refueling canal Rev 0

94 ID: 98409 Points: 1.00 Which of the following states the EDG and support systems design mission time?

Which system design limits are being protected by the EDGs per T.S. basis?

A. 1. 7 days

2. Fuel and Containment B. 1. 7 days

2. S/G (secondary side) and SFP C. 1. 10 days

2. Fuel and Containment D. 1. 10 days

2. S/G (secondary side) and SFP Answer: A Answer Explanation:

TS BD 3.8.3 Correct - Total support system time is design 7 days per TS. The systems is protected are the fission product boundary (fuel, RCS, containment)

Incorrect - 7 days S/G and SFP. First part is correct as well as RCS but SG are not part of the systems the EDG is protecting. Yes they are half in the RCS but the secondary side is not used in the design bases accident. Plausible if the student knows the time but not what the EDG function is during a DBA.

Incorrect - 10 days Fuel and containment. The systems are correct but the time is not. TS do discuss a 10 supply but not as the design of the system. Plausible if the student knows the systems but not the design run time.

Incorrect - 10 days SG and SFP. Time is wrong and the SG are wrong. The RCS is correct. Plausible if the student is not familiar with TS bases for the EDG or the mission time.

Meets KA asks ability to explain system design limits. This question asks for the EDG limits as it applies to the fission product boundaries.

SRO knowledge TS bases for EDG support systems Rev 0

Question 94 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98409 User-Defined ID: 98409

Reference:

BD TS 3.8.3 Topic: 94 SRO EDG mission time and systems its protecting RO Importance Rating: 3.8 SRO Importance Rating: 4.0 K/A Number: 2.1.32 Comments: NEW Lesson Plan Objective: LO1732700R7, DESCRIBE the bases of Technical Specifications and how the bases relate to the various sections.

Tier # 3 Group #

Last Used - N/A Fundamental 55.43 part 2 KA - Conduct of ops - Ability to explain and apply system limits and precautions.

Rev 0

95 ID: 98410 Points: 1.00 Given the following with the plant shutdown and in a refueling outage:

Core offload is in progress A single head stud was left in the vessel flange (stuck)

The area was programed into the PLC of the Refueling Machine A need has arose which will require operating the Refueling Machine in Bypass mode Based on the conditions given can the Bypass operation be allowed?

Why and or how?

A. 1. NO

2. Bypass operation is NOT allowed with fuel in the vessel.

B. 1. YES

2. The Manipulator Crane will still prevent movement in the affected area.

C. 1. NO

2. The Manipulator Crane will NOT prevent movement inside of the affected area.

D. 1. YES

2. The SRO and crane operator can maintain heightened awareness around the stud area.

Answer: D Answer Explanation:

Correct - per FHP 03-001 the refuel machine may be used in bypass mode as long as the SRO and the crane operator maintain heightened awareness around the area.

Incorrect - NO, the Manipulator Crane will not prevent hitting the stud. Plausible as the reason is correct the PLC will not prevent the crane from hitting the stud in bypass mode but this operation is allowed by FHP procedure.

Incorrect - NO, bypass operation is not allowed with fuel in the vessel. Plausible as bypass operation will allow the crane to move in areas it normally would not and with fuel in the vessel this is a conservative choice.

Incorrect - YES, the Manipulator Crane will still maintain the area of the stud off limits. Plausible if the student does not understand the bypass operation of the refuel machine.

Meets KA asks about knowledge of the refuel process with respect to the refueling machine SRO knowledge the refueling process with regards to the refuel machine Rev 0

Question 95 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98410 User-Defined ID: 98410

Reference:

FHP 001 Topic: 95 SRO refuel in bypass mode with a stuck stud RO Importance Rating: 2.8 SRO Importance Rating: 3.7 K/A Number: 2.1.41 Comments: NEW Lesson Plan Objective:

Tier # 3 Group #

Last Used - N/A Fundamental 55.43 part 7 KA - Conduct of ops - Knowledge of the refueling process Rev 0

96 ID: 98411 Points: 1.00 The unit has been shut down in preparation for a refueling outage. The Operations Manager is conducting a brief for GEN 00-008, RCS LEVEL LESS THAN REACTOR VESSEL FLANGE OPERATIONS.

During the performance of this 'Infrequently Performed and Potentially Degrading Evolution' brief which of the following is required to be discussed?

A. Contact information of field participants.

B. The need for managing breaks during the test.

C. Responsibilities of management for oversight of the test.

D. The importance of completing the test as quickly as possible.

Answer: C Answer Explanation:

SRO level question based on applicant having to have the knowledge of administrative requirements concerning the conduct of Infrequently Performed Test or Evolution (IPTE) which could affect the Margin of Safety for the plant.

This brief is required to be provided by management ONLY so ROs would not perform this.

Correct - as stated in the AP Incorrect - contact info. While important this would be discussed in the pre job brief not the IPTE brief by management. Plausible if the student is unfamiliar with the IPTE briefing requirements.

Incorrect - completing as quickly as possible. This is always part of every outage time pressure. For this case it would be important to have the test move in an efficient manor to minimize time in reduced inventory but not at the expense of safety or correctness. Plausible if the student is unfamiliar with the IPTE briefing requirements.

Incorrect - managing breaks. This could be discussed in the pre job brief if the test would take a long time but not applicable for the management oversight brief. Plausible if the student is unfamiliar with the IPTE briefing requirements.

Meets KA asks knowledge of process for conducting a special test brief SRO knowledge SRO job function Rev 0

Question 96 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98411 User-Defined ID: 98411

Reference:

AI 15C-006 96 SRO requirements during an IPTE brief for Topic:

management RO Importance Rating: 2.9 SRO Importance Rating: 3.6 K/A Number: 2.2.7 Comments: MODIFIED - Callaway Lesson Plan Objective: LO1733201 R3, EXPLAIN the duties and responsibilities of the Control Room Supervisor IAW AP 17C-006.

Tier # 3 Group #

Last Used - 2013 Callaway Fundamental 55.43 part 5, 6 KA - Equipment control - Knowledge of the process for conducting special or infrequent test Rev 0

97 ID: 98412 Points: 1.00 Maintenance activities are being performed during full power operations that have resulted in additional work needing to be added to the original job Work Scope and it now requires contingency measures be put into place to perform the activity.

What procedure would be used to screen, assess, and manage the addition to the original Work Scope?

A. AP 22C-002, WORK CONTROLS B. AP 22A-001, SCREENING, PRIORITIZATION, AND PRE-APPROVAL C. AP 16C-006, MPAC WORK REQUEST / WORK ORDER PROCESS CONTROLS D. AP 22C-003, ON-LINE NUCLEAR SAFETY AND GENERATION RISK ASSESSMENT Answer: D Answer Explanation:

Correct - This procedure assesses the added risk for the scope growth.

Incorrect - AP 22C-002. This procedure will not cover the work scope change.

Plausible since this procedure deals with work in general whether online or not.

Incorrect - AP 22A-001. This procedure will not cover the increase in risk.

Plausible since this procedure discusses the screening of work which for this case would seem to be correct.

Incorrect - AP 16C-006. This procedure will not cover the increase in risk.

Plausible since this procedure discusses the generation of the new work order or instructions.

Meets KA asks which procedure will assess the correct risk due to added work scope while online.

SRO knowledge due to job function Rev 0

Question 97 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98412 User-Defined ID: 98412

Reference:

AP 22C-003 Topic: 97 SRO AP 22C-003 scope and risk RO Importance Rating: 2.6 SRO Importance Rating: 3.8 K/A Number: 2.2.17 Comments: BANK - 16910 Lesson Plan Objective: LO1734018 R17, DESCRIBE the purpose, scope and operator responsibilities of procedure AP 22C-003, ON-LINE NUCLEAR SAFETY AND GENERATION RISK ASSESSMENT.

Tier # 3 Group #

Last Used - N/A Fundamental 55.43 part 5 KA - Equipment control - Knowledge of the process for managing maintenance activities during power operations such as risk assessments work prioritization and coordination with the transmission system operator Rev 0

98 ID: 98413 Points: 1.00 Given the following plant conditions:

A LOCA has occurred A General Emergency has been declared and the TSC has been activated A two man repair team was dispatched from the TSC into the Aux Building for emergency repairs on a valve to stop the release Due to an accident, both repair team members have been injured, one with a life threatening head injury

1. What is the maximum allowed dose (TEDE) that can be authorized for a rescue team in accordance with EPP 06-013, EXPOSURE CONTROL AND PERSONNEL PROTECTION?

AND

2. Under the conditions given, who approves this dose?

1 2 A. 25 rem Shift Manager B. 10 rem Shift Manager C. 10 rem Site Emergency Manager D. 25 rem Site Emergency Manager Answer: D Answer Explanation:

Correct - This is a non delegatable duty of the site emergency manager and since the repair team was sent from the TSC the shift manager is no longer the emergency manager. The 25 rem TEDE is per procedure.

Incorrect - 25 rem shift manager. Dose is correct but since the TSC is active the shift manager cannot make this determination, only the site emergency manager.

Plausible if the student doesn't know the non delegatable duties.

Incorrect - 10 rem site emergency manager. Approval is correct but the dose is less than allowed by procedure. Plausible if the student doesn't know the dose limits for life saving.

Incorrect - 10 rem shift manager. Dose is less than allowed by procedure and approval is non delegatable to shift manager since the TSC is active. Plausible if the student doesn't know the live saving dose limits or the non delegatable approvals.

Meets KA asks what rad dose is the limit for emergencies. Knowing this limit and who can authorize it is knowledge of the rad hazards that can arise during emergencies.

SRO knowledge job function Rev 0

Question 98 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98413 User-Defined ID: 98413

Reference:

EPP 06-013 Topic: 98 SRO life saving exposure limits and who can approve RO Importance Rating: 3.4 SRO Importance Rating: 3.8 K/A Number: 2.3.14 Comments: BANK - Callaway Lesson Plan Objective: LO1734020 R1, Determine exposure limits and posting requirements in the ALARA program.

Tier # 3 Group #

Last Used - 2013 Callaway #98 Fundamental 55.43 part 4 KA - Rad control - Knowledge of rad or contamination hazards that may arise during normal abnormal or emergency conditions or activities

(#98 2011 ANO2 - least exposure)

Rev 0

Q99, Rev 1 ID: 98414 Points: 1.00 The crew is recovering from a reactor trip that occurred 5 minutes ago due to an I&C technician inadvertently manipulating incorrect switches in the back of the Control Room. The following indications are noted:

• Crew is performing EMG ES-02, REACTOR TRIP RESPONSE

• RCS pressure 2235 psig stable

• RCS temperature 555°F lowering at 2°F per minute (reason unknown)

• RCS T ~1 to 2 °F

• PZR level 26% stable

• AFW flow 100,000 lbm/hr - throttled per fold out page

• MSIVs Closed Assuming the above conditions do NOT change over the next 35 minutes which of the following actions will be taken?

A. Direct an SI and transition to EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

B. Transition to EMG FR H-1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.

C. Transition to OFN BG-009, EMERGENCY BORATION, then return to EMG ES-02 step in effect.

D. Direct the RO to perform OFN BG-009, EMERGENCY BORATION, while the CRS and the BOP continue with EMG ES-02.

Answer: D Answer Explanation:

SRO because to answer you have to know the procedure content (steps) to understand that the emergency boration will take place concurrently with the EMG in effect. Comes from AP 15C-003 Correct - Per step 2, continuous action step, if temperature lowers to less than 530°F and no SI (that is why we are in ES-01) then borate so you don't lose SDM. This is how OFN's and EMG's are performed together.

Incorrect - transition to OFN BG-009. This procedure is needed due to the cooldown but ES-02 directs in step 2 RNO that both procedures are to be completed concurrently.

Plausible if the student doesn't know about the temperature requirement AND the use of concurrent OFN and EMG.

Incorrect - direct an SI and transition to E-0. With temperature lowering a jump in logic could be made that will also lower pressure and PZR level which would require an SI. No other items are broke so PZR heaters will maintain pressure and charging will maintain level. Plausible if the student assumes other things are broke outside of what the stem states.

Incorrect - transition to FR H-1. With AFW flow less than 270,000 lbm/hr an assumption could be made that more flow is needed but since flow was throttled per fold out page then SG levels, at least one, must be over 6% NR. Plausible if the student doesn't understand why AFW flow was throttled.

Meets KA asks how to use OFNs with EOPs OPS INITIAL NRC Page: 1 of 2 14 October 2015

SRO knowledge procedure selection based on indications Question Q99, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98414 User-Defined ID: 98414

Reference:

EMG ES-02 Topic: 99 SRO OFN procedure use with EMG's RO Importance Rating: 3.8 SRO Importance Rating: 4.5 K/A Number: 2.4.8 Comments: BANK - DC Cook Lesson Plan Objective: LO1733203 R6, DISCUSS procedure implementation IAW AP 15C-002.

Tier # 3 Group #

Last Used - 2011 DC Cook #99 Comprehension 55.43 part 5 KA - Emergency procedures - Knowledge of how abnormal operating procedures are used in conjunction with EOP's Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

100 ID: 98415 Points: 1.00 The unit is operating at 100% power when a RED first out for OTT comes in.

The alarm is verified correct but the reactor does NOT trip.

The CRS directs entry into EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION / ATWS.

The crew performs their immediate actions.

The crew is now verifying the reactor and turbine trip when an SI occurs.

Which of the following statements is correct with regards to what action(s) the CRS will direct next AND what is the basis for this action?

A. Direct the RO to return to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, and perform steps 1-7, while the BOP and the CRS continue with EMG FR-S1.

This is to verify the emergency borate flow remains above 30 gpm until SDM is determined.

B. Direct the RO to return to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, and perform steps 1-7, while the BOP and the CRS continue with EMG FR-S1.

This action is to verify that all of the SI actuated equipment has actuated.

C. Continue with the EMG FR-S1 until a transition is made back to EMG E-0 step

1. This action is to verify that all of the SI actuated equipment has actuated.

D. Continue with EMG FR-S1 until a transition is made back to EMG E-0 step 1.

This is to verify the emergency borate flow remains above 30 gpm until SDM is determined.

Answer: B Answer Explanation:

Correct - At the step the crew is in they have passed the continuous action step to check for SI. This step sends the crew to perform the first 7 steps of E-0 and then when the CRS completes FR-S1 by moving out to step 25 and 27 a transition will be made to E-0 step 8 since the RO has completed the auto SI verification steps. The bases is there for the SI verification, all the equipment functioned properly and the SI is valid.

Incorrect - Direct the RO back to E-0 and verify emergency boration flow is above 30 gpm. The first part is correct the crew will send an operator back to E-0 to verify the SI. The bases is wrong because even if the emergency boration was started and flow verified prior to the SI when the SI happens the BA transfer pumps are load shed, the emergency borate valve is load shed, and the emergency borate flow goes to 0. This is because the SI is now injecting not the emergency flow path. Plausible if the student doesn't understand what actions take place on the SI.

Incorrect - Continue with FR-S1 and to verify SI. If the crew does not send an operator back to E-0 and then makes it to step 27 they could transition to E-0 at step 1 or step 8, the distinction is whether SI has occurred or not. The verification of the SI must be completed by either the operator sent back to perform it or the crew. The second part is correct to verify the SI. Plausible if the student understands the verification part but doesn't catch the fact that the first part is sending them back to step 1 so the verification will not be made.

Incorrect - Continue with FR-S1 and verify emergency borate flow above 30 gpm.

If the crew does not send an operator back to E-0 and then makes it to step 27 they will transition to E-0 at step 1 or step 8. The verification of the SI must be completed by either the operator sent back to perform it or the crew. The bases is wrong because even if the emergency boration was started and flow verified prior to the SI when the SI happens the BA transfer pumps are load shed, the emergency borate valve is load shed, and the emergency borate flow goes to 0.

This is because the SI is now injecting not the emergency flow path. Plausible if the student doesn't understand what actions take place on the SI and forgets that the SI verification steps must be completed.

Meets KA asks SRO knowledge because it asks about knowledge of the content of the procedure, the procedure steps and sequence must be known to understand what has been completed and what has not.

Question 100 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 4.00 System ID: 98415 User-Defined ID: 98415

Reference:

BD EMG FR S-1 100 SRO knowledge of bases for FR-S1 procedure steps Topic:

for E-0 return RO Importance Rating: 3.4 SRO Importance Rating: 4.4 K/A Number: 2.4.23 Comments: NEW Lesson Plan Objective: LO1732339 R3, Given an EMG FR-S1 procedure flow path, EXAMINE the available options for procedure actions.

Tier # 3 Group #

Last Used - N/A Comprehension 55.43 part 5 KA - Emergency procedures - Knowledge of the basis for prioritizing emergency procedure implementation during emergency operations

1 ID: 98316 Points: 1.00 Step 5 in EMG E-0, REACTOR TRIP OR SAFETY INJECTION, states 'Check if SI is required'.

What is the purpose of this step?

A. Ensure BIT valves are closed so flow can be directed to the normal flow path if indications warrant.

B. To determine which signal caused the SI for future use in the procedure.

C. To ensure BOTH trains of SI have actuated and if NOT to actuate BOTH trains.

D. Maximize the time available to prevent a possible PZR over pressurization caused by going water solid.

Answer: D Answer Explanation:

Correct - per BD E-0 step 5 checks for if this SI is inadvertent and if so it stops all but one charging pump then closes the ECCS flow path valves to the RCS from the CCPs. This action maximizes the time the crew will have to restore normal letdown to prevent the PZR from going water solid and lifting a PORV.

Incorrect - ensure bit valves are closed. The RNO for this step will close the BIT valves IF the SI is inadvertent. Plausible if the student remembers that this step RNO will close the valves but doesn't understand why.

Incorrect - determine which signal caused the SI. The step asks for each signal that could have caused the SI. Plausible if the student uses the check step as a need to know what caused the SI as to if the SI is needed.

Incorrect - ensure both trains of SI have actuated. Step 4 of the procedure checks for this. Plausible as this step deals with the SI but only to determine if its actuated not if its required.

Meets the K/A because it asks for the student to have knowledge of the reasons for steps in EMG E-0 This question is RO level because it asks for bases or knowledge of EMG procedure steps Rev 0

Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98316 User-Defined ID: 98316

Reference:

BD EMG E-0 Topic: 1 RO Reason for steps in E-0 RO Importance Rating: 4.0 SRO Importance Rating: 4.6 K/A Number: 007 EK 3.01 Comments: NEW Lesson Plan Objective: LO1732313 R4, EXPLAIN the bases and any knowledge requirements for selected procedure steps.

Tier #1 Group #1 Last Used - N/A Memory 55.41 part 10 KA - Reactor trip - Knowledge of the reasons for the following as they apply to a reactor trip - Actions contained in EOP for reactor trip - Safety function 1 Modification History:

0 - Revised based on Scott's comments from 4/25/15 1 - revised based on Robs comments from 8/6/15 Rev 0

2 ID: 98317 Points: 1.00 Given the following plant conditions:

• Plant is operating at 100%

• The crew is responding to a reactor trip and SI actuation

• EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, is in progress at step 2

• PZR is 79% and up slow

• RCS pressure is 1780 psig and down slow Based on these conditions, the failure is a leak through the...

A. RCS hot leg.

B. PZR safety valve(s).

C. PZR liquid space sample valve.

D. charging header connection to the RCS loop.

Answer: B Answer Explanation:

Correct - A steam space leak will lower pressure in the RCS but PZR level will raise.

Incorrect - PZR liquid space. This leak location will cause RCS pressure and PZR level to lower. Plausible if the student doesn't understand the leak location.

Incorrect - charging header connection. This leak location will cause RCS pressure and PZR level to lower. Plausible if the student doesn't understand the leak location.

Incorrect - RCS hot leg. This leak location will cause RCS pressure and PZR level to lower. Plausible if the student doesn't understand the leak location.

Meets the K/A because it asks the student to evaluate the indications given and determine the leak is a vapor space leak via the safety valve which would be an interrelation between the vapor space leak phenomena and the valve.

RO level because it asks only for what could cause the indications given not what or where to go.

Rev 0

Question 2 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98317 User-Defined ID: 98317

Reference:

USAR 15.6.1.1 Topic: 2 RO PZR steam space leak indications RO Importance Rating: 2.7 SRO Importance Rating: 2.7 K/A Number: 008 AK 2.01 Comments: BANK - Braidwood Lesson Plan Objective: LO1610722, R8, DISCUSS the effects of the inadvertent opening of a Pressurizer safety valve transient.

Tier #1 Group #1 Last Used - 2013 Braidwood # 40 Comprehension 55.41 part 14 KA - PZR vapor space accident - Knowledge of the interrelations between the PZR vapor space accident and the following - Valves - Safety function 3 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

3 ID: 98318 Points: 1.00 Given the following:

• BB PS-455F, PZR Master Pressure Controller, is selected to P457/P456

• Then BB PI-457 suddenly fails to 2500 psig.

Which, if any, PORV Block valve(s) MUST to be closed to stop the pressure from lowering?

A. BB HV-8000A ONLY B. BB HV-8000B ONLY C. Neither PORV Block valve D. Both PORV Block valves Answer: A Answer Explanation:

Validated on desktop simulator BB HIS-455A PORV is controlled by the output of PZR master pressure controller BB PK-455. Since the selected channel failed hi the associated PORV will open but the other PORV will not. BB HV-456A is controlled by 2/4 pressure channels above setpoint only.

Correct - selected controlling pressure channel fails hi 455A will open and 456A will not. If the block valve for the open PORV were closed then RCS pressure will stabilize out until the problems can be corrected.

Incorrect - Both PORV block valve. Plausible if the student thinks this failure will affect both PORVs in the same way that the controller will operate both valves.

Incorrect - BB HV-8000B. Plausible if the student confuses which PORV the controlling pressure channel controls. If they think of this in reverse then this would be the correct response.

Incorrect - Neither PORV block valve. Plausible because if the actual pressure lowered to less than 2185 psig then the associated PORV block valve will close and since the PORV is a pilot operated valve it will lose its opening force and reclose. The non-affected PORV would be closed before the block valves could close.

Meets the K/A because the opening of the PZR PORV is a SBLOCA and it asks the student to understand how the PORVs function.

RO knowledge due to asking how the controller works and responds to inputs.

Rev 0

Question 3 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98318 User-Defined ID: 98318

Reference:

BD EMG E-1 3 RO PORV valve positions during a pressure transient Topic:

caused by a pressure channel failure RO Importance Rating: 3.9 SRO Importance Rating: 4.1 K/A Number: 009 EA 1.15 Comments: New Lesson Plan Objective: SY1301000 R3, Realize how a Pressurizer Pressure Control Channel regulates the operation of the master pressure controller, heaters, spray valves, PORV and PORV block valves.

Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 7 KA - Small break LOCA - Ability to operate and monitor the following as they apply to a small break LOCA - PORV and PORV block valve - Safety function 3 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

4 ID: 98319 Points: 1.00 Using the attached NPIS indication, determine what indication AND procedure action that is required by the crew.

A. 'A' RCP stator winding temperature is too high, trip the reactor, trip 'A' RCP B. 'B' RCP motor bearing temperature is too high, trip the reactor, trip 'B' RCP C. 'C' RCP #1 seal and bearing water temperature is too high, perform a controlled shutdown to remove 'C' RCP from service D. 'D' RCP thrust bearing temperature is too high, perform a controlled shutdown to remove 'D' RCP from service Answer: B Answer Explanation:

Correct - The value for an immediate RCP shutdown has been exceeded (bearing temp of 195F) per foldout page of OFN BB-005, RCP MALFUNCTIONS, the RCP must be shutdown immediately.

Incorrect - A RCP and trip. The value given is high but for this to cause an immediate shutdown this would have to exceed 299F. Plausible if the student confuses which values go with which component (common misconception).

Incorrect - C RCP and controlled shutdown. The value give for seal and bearing temperature is high but it would have to exceed 230F to be and immediate shutdown and this also only calls for a controlled shutdown. Plausible if the student confuses the values for immediate shutdown and controlled shutdown.

Incorrect - D RCP and controlled shutdown. The thrust bearing temperature is high but would need to be over 195F for this to cause any action for the crew. Plausible if the student confuses the values for immediate and controlled shutdown (common misconception)

Meets the K/A because asks student to interpret plant computer screen to evaluate system status and make a decision as to actions taken based on that evaluation.

RO knowledge because foldout page items for off normal Rev 0

Question 4 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98319 User-Defined ID: 98319

Reference:

OFN BB-005 4 RO RCP malfunctions that require action per OFN BB-Topic:

005 RO Importance Rating: 3.9 SRO Importance Rating: 3.8 K/A Number: 015 2.1.19 Comments: New Lesson Plan Objective: LO1732415 R2, RECOGNIZE the available situations which are addressed by procedure OFN BB-005.

Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 10 KA - Reactor coolant pump malfunctions - Conduct of ops -

Ability to use plant computers to evaluate system or component status - Safety function 4 Modification History:

0 - Revised based on Scott's comments from 4/25/15 1 - revised based on Robs comments 8/6/15 2 - replaced based on incorrect KA selection Rev 0

Rev 0 5 ID: 98320 Points: 1.00 Given the following:

• The plant is being cooled down for a refueling outage

• RCS pressure is 345 psig stable

• RCS temperature is 333°F stable

• PZR level 25% stable

• 'A' RHR pump in cooldown mode of operation

• 'B' RHR pump in STBY ECCS lineup

• Then: 'A' RHR pump TRIPS With NO operator action, RCS pressure/temperature will _______________, RHR hot leg suction isolation valve will ____________, and the RHR suction relief will ____________.

A. lower, close, remain closed B. lower, remain open, remain closed C. rise, close, lift if pressure rises to 475 psig D. rise, remain open, lift if pressure rises to 475 psig Answer: D Answer Explanation:

Wolf Creek has no auto isolation of RHR piping if pressure rises.

Correct - If RHR is lost during a plant cooldown then RCS pressure and temperature will rise. The RHR suction valves have an interlock to not open unless pressure is lower than 360 psig. If pressure rises after the valve is open there is no effect due to no interlock to close. The suction relief valve lifts at 450 psig so the valve should be open at the given pressure.

Incorrect - Lower, open, and relief closed. The suction valve will be open. The suction relief being closed is plausible if the student confuses the lift setpoint with the RHR discharge relief.

Incorrect - rise, closed and relief open. Plausible if the student recalls the interlock for opening the RHR suction valve and misunderstands that it will not reclose if pressure rises. The relief will be open at this pressure.

Incorrect - lower, closed and relief closed. Plausible if the student recalls the interlock for opening the RHR suction valve and misunderstands that it will not reclose if pressure rises. The suction relief being closed is plausible if the student confuses the lift setpoint with the RHR discharge relief.

Meets the K/A because the question poses a loss of RHR and asks if the system will isolate due to pressure rise. Since there is no auto response for the RHR isolation the only response is that of RCS pressure and its effect on the reliefs.

Rev 0

RO knowledge system interlock and design understanding Question 5 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98320 User-Defined ID: 98320

Reference:

SY1300500 5 RO RHR hot leg suction valve and suction relief valve Topic:

operations RO Importance Rating: 3.3 SRO Importance Rating: 3.7 K/A Number: 025 AK 3.02 Comments: NEW Lesson Plan Objective: SY1300500 R6, Explain the operation of the system during normal, off normal, and emergency operating modes.

Tier #1 Group #1 Last Used - N/A Fundamental 55.41 part 7 KA - Loss of RHR - Knowledge of the reasons for the following responses as they apply to the loss of RHR -

isolation of RHR low pressure piping prior to pressure increase above specified level.

Modification History:

Rev 0

6 ID: 98321 Points: 1.00 Given the following:

• Reactor power is 100%

• RO notes 053D, CCW SRG TK B LEV HILO, is lit

• BOP notes 'B' CCW surge tank level is 18% and down slow

• 'B' CCW loop is supplying all cooling Per procedure the operators will...

A. place the 'B' train CCW pumps in P-T-L, trip the reactor, trip RCP's.

B. place the 'B' train CCW pumps in P-T-L, place ALL safety related pumps in P-T-L.

C. manually align BL water to the 'B' CCW surge tank prior to level lowering to less than 15% while continuing to locate the leak.

D. manually align ESW to the 'B' CCW surge tank prior to level lowering to less than 15% while continuing to locate the leak.

Answer: D Answer Explanation:

OFN EG-004, ALR 53D 053D comes in at CCW surge tank level of 19%

The OFN will have the operators check if auto normal makeup is aligned and then if level is still lowering to manually lineup ESW and continue to look for the leak. The 15% is based on foldout page that states if surge tank level drops to less than 15% then to trip the reactor and continue with the procedure.

Correct - manually align ESW before 15% and continue to look for leak Incorrect - place B train CCW pumps in PTL and trip reactor, plausible since foldout page will direct this AFTER surge tank level lowers to less than 15%.

Plausible if the student doesn't remember when the foldout page directs this action.

Incorrect - place B train CCW pumps PTL, place all safety related pumps in P-T-L plausible since foldout page will direct this if safety loop is aligned to train being affected if level lowers to less than 15%. Plausible if the student doesn't remember when the foldout page directs this action.

Incorrect - manually align BL water plausible if student confuses normal auto makeup and ESW manual makeup.

Rev 0

Meets the K/A because it asks the operator to operate the CCW and related system for a loss of CCW which will happen if surge tank level continues to lower. It contains surge tank level, level control. At Wolf Creek the low level alarm for the surge tank doesn't come in until 19% so auto makeup should have refilled the surge tank without the alarm from coming in. The radiation alarm will not come in on a leak out of the system only a radioactive leak into the system.

For this question the loss of surge tank level is the cause of the loss of CCW since a radiation alarm would indicate a leak into the system and not be a loss of CCW only that a leak is into the system.

RO knowledge since it asks for the operator to understand the purpose and overall mitigative strategy of the procedure.

Question 6 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98321 User-Defined ID: 98321

Reference:

OFN EG-004 6 RO manual ESW makeup for CCW surge tank on Topic:

lowering level RO Importance Rating: 3.1 SRO Importance Rating: 3.1 K/A Number: 026 AA 1.05 Comments: NEW Lesson Plan Objective: LO1732414 R3, Given a procedure flow path, EXAMINE the available options for procedure actions.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Loss of CCW - Ability to operate and or monitor the following as they apply to the loss of CCW - The CCWS surge tank, including level control and level alarms and radiation alarm - Safety function 8 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

7 ID: 98322 Points: 1.00 Given the following:

• Reactor power is 100%

• PZR Master Pressure Controller output drifts low (indicator needle down scale)

Five minutes after the failure RCS subcooling will be ______ AND PZR subcooling will be A. higher, higher B. same, higher C. same, same D. higher, same Answer: D Answer Explanation:

Correct - output lowering will cause heaters to turn on to raise pressure and sprays would not respond to control pressure since the whole system is contorted by the output of the master pressure controller. A PORV will open after pressure gets to setpoint to stop the pressure rise. Since pressure is now higher and temperature is the same RCS subcooling is higher and the PZR pressure and temperature is higher so subcooling would be same.

Incorrect - higher, higher. Plausible if the student confuses pressure and temperature in the PZR since that system will stay at saturated.

Incorrect - same, same. Plausible if the student confuses the fact that the PZR will not change subcooling and thinks the RCS will follow what the PZR will do.

Incorrect - same, higher. Plausible if the student reverses the change in pressure and temperature between the RCS and PZR.

Meets the K/A because ask for understanding of subcooling with respect to the PZR.

RO because it asks about how the system works.

Rev 0

Question 7 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98322 User-Defined ID: 98322

Reference:

SY1301000 Topic: 7 RO Failure of a PZR pressure controller RO Importance Rating: 3.1 SRO Importance Rating: 3.4 K/A Number: 027 AK 1.01 Comments: NEW Lesson Plan Objective: SY1301000 R3, Realize how a Pressurizer Pressure Control Channel regulates the operation of the master pressure controller, heaters, spray valves, PORV and PORV block valves.

Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 14 KA - PZR pressure control system malfunction - Knowledge of the operational implication of the following concepts as they apply to PZR pressure control malfunctions - Definition of saturation temperature - Safety function 3 Modification History:

0 - Revised based on Scott's comments from 4/25/15 Rev 0

8 ID: 98323 Points: 1.00 Given the following:

• 087F, TURB TRIP & P9 RX TRIP, first out is lit

• BOP reports all main stop valves have closed

• RO reports that NO rod bottom lights are lit

• RO reports RX power is 55% down slow

• RO operates SB HS-1, REACTOR TRIP MAN ACTUATION, with NO change in indications

• BOP operates SB HS-42, REACTOR MAN TRIP, with NO change in indications For this condition which of the following are the breaker indications?

A.

B.

C.

D.

Answer: D Answer Explanation:

Correct - For an ATWS the normal plant lineup would be the trip breakers closed and the bypass breakers racked out.

Rev 0

Incorrect - all red lights lit. Plausible if the student thinks that the normal lineup would be all trip and bypass breakers closed for the ATWS.

Incorrect - all green lights lit. Plausible if the student thinks that after the trip switches have been actuated the breakers should be open even for an ATWS.

Incorrect - trip breakers green. Plausible if the student thinks that after the trip switches have been actuated the breakers should be open even for an ATWS.

Meets the K/A because it has the student identify the breaker positions from what is expected for an ATWS condition.

RO knowledge because it is system knowledge to determine breaker position.

Question 8 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98323 User-Defined ID: 98323

Reference:

EMG E-0 Topic: 8 RO ATWS and reactor trip breaker indicating lights RO Importance Rating: 4.2 SRO Importance Rating: 4.3 K/A Number: 029 EA 2.07 Comments: NEW Lesson Plan Objective: SY1300100 R5, EXPLAIN operation of the Rod Control System motor generators and reactor trip breakers.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - ATWS - Ability to determine or interpret the following as they apply to a ATWS - Reactor trip breaker indicating lights - Safety function 1 Modification History:

0 - changed based on Scotts feedback 4/25/15 Rev 0

9 ID: 98324 Points: 1.00 During the performance of EMG C-21, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, the following conditions exist:

• RCS cooldown rate is determined to be 125°F/hr

• All S/G NR levels are off-scale low

• RCS hot leg temperatures are lowering

• Containment pressure is normal

1. Which of the following describes how the crew is directed to control AFW flow?

2. What is the basis for this action?

A. 1. Flow is reduced and maintained at 30,000 lbm/hr to any S/G with less than 6% NR level

2. Help stabilize hot leg temperatures to establish conditions for SI termination B. 1. Flow is reduced and maintained at 30,000 lbm/hr to any S/G with less than 6% NR level

2. Maintain RCS cooldown rate C. 1. Flow is maintained greater than 270,000 lbm/hr until at least ONE S/G NR level is above 6%

2. Help stabilize hot leg temperatures to establish conditions for SI termination D. 1. Flow is maintained greater than 270,000 lbm/hr until at least ONE S/G NR level is above 6%

2. Maintain RCS cooldown rate Answer: A Answer Explanation:

Correct - Flow is reduced per step 5 but the caution above the step discusses minimum flow to any SG with NR level less than 6%. The basis discusses that after feed flow is lowered hot leg temperature will rise and step 5 is a continuous action step so flow and steam dumps will control hot leg temperature to establish conditions for SI termination.

Incorrect - flow reduced to 30,000 and maintain RCS cooldown rate. First part is correct. If the student knows cooldown needs to be slowed down then lower AFW flow would help that. Plausible if the student doesn't understand that stabilizing RCS temperature and pressure will allow for SI termination.

Incorrect - flow maintained at 270,000 and maintain RCS cooldown rate. Flow at 270,000 is the number used for all accidents until at least one SG has 6% NR level. This is not made for this case since the feed flow is helping to cause the cooldown. If the student believes this cooldown rate is appropriate. Plausible if the student doesn't understand that stabilizing RCS temperature and pressure will allow for SI termination.

Rev 0

Incorrect - flow maintained at 270,000 and establish conditions for SI termination.

Flow at 270,000 is the number used for all accidents until at least one SG has 6% NR level. This is not made for this case since the feed flow is helping to cause the cooldown. SI termination part is correct. Plausible if the student forgets that for this specific accident AFW flow is not desirable.

Meets the K/A because it asks for interrelations between the uncontrolled depressurization of all SGs and AFW flow (safety system instrumentation). This interrelation is what to do with AFW flow if this happened.

RO knowledge because it asks for mitigative strategy of the procedure.

Question 9 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98324 User-Defined ID: 98324

Reference:

BD EMG C-21 Topic: 9 RO What to throttle AFW flow to in C-21 and why RO Importance Rating: 3.4 SRO Importance Rating: 3.7 K/A Number: E12 EK 2.1 Comments: BANK - Vogtle Lesson Plan Objective: LO1732334 R3, Discuss the bases for procedural actions in EMG C-21.

Tier # 1 Group # 1 Last Used - 2011 Vogtle #65 Comprehension 55.41 part 7 KA - Uncontrolled depressurization of all steam generators

- Knowledge of interrelations between the uncontrolled depressurization of all steam generators and the following -

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features - Safety function 4 Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

10 ID: 98325 Points: 1.00 Given the following with the unit at 30% power:

• 'B' Main Feed Reg Valve fails open

• 'B' S/G level rises to the P-14 setpoint

• All S/G NR levels have remained 25% or higher

• NO operator actions What affect does this failure have on the AFW system?

A. All AFW pumps start All SMART valves throttle appropriately B. NO AFW pumps start NO SMART valves throttle C. ONLY the TDAFW pump starts NO SMART valves throttle D. ONLY 'A' and 'B' MDAFW pumps start All SMART valves throttle appropriately Answer: D Answer Explanation:

LER 15003 Wolf Creek manual reactor trip due to high steam generator level at low power Correct - The P-14 causes a reactor trip and a trip of both MFW pumps which is a start signal for the MDAFW pumps only since SG level never lowered to the low setpoint for the TD pump. The SMART valves will throttle to limit total flow to each SG.

Incorrect - all AFW pumps start. If all the AF pumps start all the SMART valves will be closed due to the TD pump running. Plausible as on a normal trip SG levels shrink lower than the low SG level AFW actuation setpoint. Since the stem states all levels are higher than that and the TDAFW pump didn't get a start signal it would not be running Incorrect - no AFW pumps start. The SMART valves would not move if no pumps start. Plausible since the stem states that no SG level is lower than the actuation setpoint. If the student doesn't understand that the P-14 causes a trip of both MFW pumps which is a start for the MDAFW pumps.

Incorrect - only the TD pump starts. The D SMART valve will throttle but the others will stay open. Plausible if the student confuses which signal starts which pump Meets the K/A because the ability to determine the state of the AFW system on a loss of MFW is asked.

RO knowledge because it asks for system understanding.

Rev 0

Question 10 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98325 User-Defined ID: 98325

Reference:

SY1505900 Topic: 10 RO trip of both MFW pumps start of MD pumps RO Importance Rating: 4.2 SRO Importance Rating: 4.3 K/A Number: 054 AA 2.04 Comments: NEW Lesson Plan Objective: SY1505900 R11, Discuss the instrumentation and controls of the Feedwater System, including trips and automatic actions of a Feedwater Isolation Signal (FWIS).

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - Loss of main feedwater - Ability to determine and interpret the following as they apply to the loss of main feedwater - Proper operation of AFW pumps and regulating valves - Safety function 4 Modification History:

0 - changed based on Scotts comments 4/25/15 Rev 0

11 ID: 98326 Points: 1.00 A station black out occurred at 0800. The crew entered EMG C-0, LOSS OF ALL AC POWER. At 0845 the crew had shed non-essential AC and DC loads per procedure from all NK batteries.

Which of the following is correct with regards to NK battery life?

A. NK11 and NK14 are expected to be totally discharged by 1200 AND NK12 and NK13 are expected to be totally discharged by 1400 B. NK11 and NK14 are expected to be totally discharged by 1400 AND NK12 and NK13 are expected to be totally discharged by 1200 C. NK11, NK12, NK13, and NK14 are expected to last until 1600 D. NK11, NK12, NK13, and NK14 are expected to be totally discharged by 1200 Answer: C Answer Explanation:

NK11 and NK14 are 1600 amp/hr batteries and NK12 and NK13 are 864 amp/hr batteries. Per BD C-0 all batteries are designed to last 240 minutes (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

Per AI 21-016, OPERATOR TIME CRITICAL ACTIONS VALIDATION and BD C-0 step 27, as long as the non-essential loads are shed off the batteries by the 60 mark all batteries will last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Correct - Having loads shed it is expected to last a full 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per AI-21-016 Incorrect - be totally discharged by 1200. This is the design capacity of the batteries but since a load shed was completed within the 60 minute window then all batteries are expected to last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Plausible if the student fails to recall why the load shed is performed.

Incorrect - NK11 and NK14 totally discharged by 1400 and NK12 and NK13 totally discharged by 1200. NK11 and NK14 are larger batteries and if the student confuses this with the load shed and time then it would be plausible.

Incorrect - NK11 and NK14 totally discharged by 1200 and NK12 and NK13 totally discharged by 1400. NK11 and NK14 are larger batteries and if the student confuses this with the load shed and time then it would be plausible.

Meets the K/A because it asks for battery discharge rate over time (capacity).

RO knowledge because it asks about system response to transient or how the system works.

Rev 0

Question 11 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98326 User-Defined ID: 98326

Reference:

AI 21-016 11 RO Loss of all AC battery discharge rate vs time with Topic:

load shed RO Importance Rating: 3.3 SRO Importance Rating: 3.7 K/A Number: 055 EK 1.01 Comments: NEW Lesson Plan Objective: LO1120201 R3, Solve for various parameters in simple AC and DC circuits using the following: Power Equations Tier #1 Group #1 Last Used - N/A Comprehension 55.41 part 8 KA - Loss of offsite and onsite power - Knowledge of the operational implications of the following concepts as they apply to the station blackout - Effect of battery discharge rates on capacity - Safety function 6 Modification History:

0-Rev 0

12 ID: 98327 Points: 1.00 A loss of offsite power has occurred.

• The Control room staff responded using the appropriate procedure.

• Four minutes into the event the RO re-scans the alarm panels and notes the following alarms locked in.

Which of the following alarms would be the highest priority for the operators to address AND why?

A. 014D, S/U XFMR TROUBLE, restoring power B. 039A, LTDN HX TEMP HI DIVERT, RCS chemistry concerns C. 055B, ESW PMP B PRESS LO, cooling flow to the EDG D. 077E, SR HI VOLT FAIL, energizing SR detectors Answer: C Answer Explanation:

ALR 055C, OFN EF-033 Correct - with the loss of offsite power the only power left is safety related EDGs which are cooled by ESW. The ALR for 055C entry condition state that the pump has tripped and locked out so with no cooling water the EDG will trip on high temperature Incorrect - 014D comes in on any loss of normal power to it so with a loss of offsite power it will be in until power is restored. Not a priority over the EDG.

Plausible if the student believes that offsite power is more important than safety related power.

Incorrect - 039A comes in on the loss of offsite power due to the loss of CCW for a time and the letdown outlet temperature rising. Plausible if the student misunderstands this is a normal alarm for the loss of offsite power and thinks since there is no cleanup of the RCS water chemistry would be negatively affected.

Incorrect - 077E should be in until the SR detectors energize which should take place about 9 minutes after the reactor trip. Plausible if the student doesn't realize that at four minutes this should still be in.

Meets the K/A because this asks for which alarm is more important than other alarms.

RO knowledge because this is an entry condition into a procedure ALR.

Rev 0

Question 12 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98327 User-Defined ID: 98327

Reference:

ALR 055C 12 RO Priority of alarms on loss of offsite power with ESW Topic:

issues RO Importance Rating: 4.1 SRO Importance Rating: 4.3 K/A Number: 056 2.4.45 Comments: NEW Lesson Plan Objective: LO1732443 R2, RECOGNIZE the available situations which are addressed by OFN EF-033.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Loss of offsite power - Emergency procedures - Ability to prioritize and interpret the significance of each annunciator or alarm - Safety function 6 Modification History:

Rev 0

13 ID: 98328 Points: 1.00 The plant is operating at 75% power when a loss of NN02 occurs.

If NN02 can NOT be restored which of the following sets of instruments will still be available to monitor actual plant parameters?

A. AE LI-553, 'C' S/G level BB LI-461, PZR level B. AB PI-525, 'B' S/G Pressure BG LI-185, VCT level C. BB PI-455, PZR pressure BB PI-405, RCS WR pressure D. AB FI-522, 'B' S/G steam flow SE NI-42, Power Range Answer: C Answer Explanation:

Correct - All instruments given are powered from other NN buses Incorrect - AE LI-553, this is NN02 powered. Plausible if the student confuses instrument power supplies.

Incorrect - SE NI-42, this is NN02 powered. Plausible if the student confuses instrument power supplies.

Incorrect - AB PI-525, this is NN02 powered. Plausible if the student confuses instrument power supplies.

Meets KA asks if student knows what the backup indications are for a loss of NN bus RO knowledge power supplies Rev 0

Question 13 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98328 User-Defined ID: 98328

Reference:

SY1506300 Topic: 13 RO NN02 powered instruments RO Importance Rating: 3.2 SRO Importance Rating: 3.4 K/A Number: 057 AA 1.05 Comments: NEW Lesson Plan Objective: SY1506300 R7, Integrate system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 1 Group # 1 Last Used - N/A Fundamental 55.41 part 7 KA - Loss of vital AC electrical instrument bus - Ability to operate and or monitor the following as they apply to the loss of vital AC instrument bus - backup instrument indications Modification History:

Replaced Rev 0

14 ID: 98329 Points: 1.00 STS KJ-005A, MANUAL/AUTO START, SYNC & LOADING OF EDG NE01, is in progress. NE01 has been started from the Control Room and is running with all operating parameters indicating normal. The crew is about to sync NE01 with NB01 (offsite power) when NE01 output voltage goes to 0.

1. Which of the following will cause the above indications?

2. How must NE01 be shutdown given this failure?

A. 1. Loss of NK04 DC bus

2. By manually closing the fuel racks due to a loss of DC control power.

B. 1. Loss of NK01 DC bus

2. By manually closing the fuel racks due to a loss of DC control power.

C. 1. Loss of NK04 DC bus

2. From NE-107 using the STOP pushbutton because the control room switch has lost power.

D. 1. Loss of NK01 DC bus

2. From NE-107 using the STOP pushbutton because the control room switch has lost power.

Answer: B Answer Explanation:

NK01 supplies DC control power to NE01 for field flashing and control from the control room. If this DC is lost when the diesel is running the output voltage will go to 0 and the engine must be shutdown locally since no power to the control room switch no exists.

Correct - per the OFN notes (and the lesson plan) if NK01 is lost with NE01 operating the output voltage will drop to zero and the diesel cannot be synced with the bus. The diesel cannot be shutdown from the control room either it must me locally stopped using the manual lever on the fuel racks.

Incorrect - loss of NK04 and local fuel rack shutdown. NK04 is the control power to NE02. The second part is correct with this loss the fuel racks are the only way to stop the engine. Plausible if the student forgets which DC power supplies which EDG.

Incorrect - NK01 and shutdown from NE-107. The DC power is correct. The local panel and the control room use the DC power to control the EDG so with this loss the local pushbuttons will not work either. Plausible if the student forgets that the same DC power controls the switches in the control room and the local panel.

Incorrect - NK04 and shutdown from NE-107. NK04 is the control power for NE02. The local panel and the control room use the DC power to control the EDG so with this loss the local pushbuttons will not work either. Plausible if the student forgets that the same DC power controls the switches in the control room and the local panel.

Rev 0

Meets the K/A because it asks for knowledge of the loss of DC control power to the diesels has on the diesels. Student must recognize the reason for closing the fuel racks vice just pushing the stop pushbutton which has lost power.

RO knowledge because it asks for system knowledge of the emergency diesel.

Question 14 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98329 User-Defined ID: 98329

Reference:

OFN NK-020 Topic: 14 RO Loss of NK01 with NE01 running affects RO Importance Rating: 3.4 SRO Importance Rating: 3.7 K/A Number: 058 AK 3.01 Comments: NEW Lesson Plan Objective: SY1406401 R4, Assess the functional interrelationship with the DC Distribution System, including electrical power supplies.

Tier # 1 Group # 1 Last used - N/A Comprehension 55.41 part 5 KA - Loss of DC power - Knowledge of the reasons for the following responses as they apply to the loss of DC power -

Use of dc control power by EDGs - Safety function 6 Modification History:

0 - changed based on Scotts feedback 4/25/15 Rev 0

15 ID: 98330 Points: 1.00 Given the following:

• The unit is operating at 100%

• A failure in the normal feeder breaker for NB01 causes the breaker to open

• As the crew is working through the appropriate procedure the RO notes that the ESW system parameters are NOT as expected Which ONE of the following valves and positions would cause abnormal ESW operating parameters?

A. EF HIS-60, ESW TRN B FROM CCW HEAT EXCHANGER, is CLOSED and needs to be OPEN B. EF HIS-32, ESW TRN B TO CTMT AIR COOLERS, is OPEN and needs to be CLOSED C. EF HIS-37, ESW TRN A DISCHARGE TO UHS, is OPEN and needs to be CLOSED D. EF HIS-41, ESW TRN A TO SERVICE WTR SYS, is OPEN and needs to be CLOSED Answer: D Answer Explanation:

OFN EF-033 attachment B valve positions for actuation After an ESW system actuation signal valves reposition automatically Correct - EF-41 should be closed on an actuation of the ESW system since flow is directed back to the UHS and not to service water Incorrect - EF-32 should be open for an actuation to allow full flow through the containment coolers. Plausible if the student confuses which valves in the ESW system reposition and to what position.

Incorrect - EF-37 should be open to allow for the ESW system to discharge back to the UHS. Plausible if the student mistakes the return back to service water and UHS.

Incorrect - EF-60 should be closed since the bypass valves are throttled for proper flow through these heat exchangers. Plausible if the student confuses the bypass and the normal outlet valve positions.

Meets the K/A because it asks for ability to interpret indications and understand how actions taken can correct these indications.

RO knowledge because it asks for ESW system operations understanding.

Rev 0

Question 15 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98330 User-Defined ID: 98330

Reference:

OFN EF-033 Topic: 15 RO ESW valve mispositioning effect on system RO Importance Rating: 4.2 SRO Importance Rating: 4.4 K/A Number: 062 2.2.44 Comments: MODIFIED - 91855 Lesson Plan Objective: SY1408900 R7, EXPLAIN the instrumentation and controls of the ESW System, including symptoms/failure modes.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 5 KA - Loss of nuclear service water - Equipment control -

Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions - Safety function 4 Modification History:

0 - modified based on feedback from Scott 4/25/15 Rev 0

16 ID: 98331 Points: 1.00 Given the following:

• A LOCA outside containment has occurred

• SI has actuated

• EMG C-12, LOCA OUTSIDE CONTAINMENT, has been entered Which ONE of the following actions will be attempted to isolate the break AND which indication is used to determine if the leak has been isolated?

A. Close RCP seal water return isolation valves (BB HIS-8141A, B, C, D) and monitor PZR level.

B. Close RCP seal water return isolation valves (BB HIS-8141A, B, C, D) and monitor RCS pressure.

C. Isolate RHR to Accumulator Injection Loop (EJ HIS 8809A) and monitor RCS pressure.

D. Isolate RHR to Accumulator Injection Loop (EJ HIS 8809A) and monitor PZR level.

Answer: C Answer Explanation:

EMG C-12 Correct - RHR to accumulator injection loop and RCS pressure. This is the only indication that is looked for in C-12. The RHR cold leg injection valve is specifically called out.

Incorrect - RCP seal return isolation valves and PZR level. C-12 doesn't look at PZR level as an indication that the leak is isolated. RCP common seal return valve is closed not each individual valve. Also the individual valves are inside containment not outside. Plausible if the student doesn't understand what indications to look for in C-12 for leak isolation and if RCP seal return valves are mistaken for being outside of containment.

Incorrect - RCP seal return isolation valves and RCS pressure. RCP wrong for reasons above. RCS pressure is correct. Plausible if the student thinks the RCP seal return valves are outside of containment.

Incorrect - RHR to accumulator injection loop and PZR level. RHR valves are correct but PZR level is not for reasons given above. Plausible if the student doesn't understand what indications to look for in C-12 for leak isolation.

Meets the K/A by having the crew isolate items and then check indications so knowledge of operational implications and indicating signals dealing with LOCA outside containment.

RO knowledge because it asks for system knowledge of the RHR system and an understanding of the isolation indication Rev 0

Question 16 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98331 User-Defined ID: 98331

Reference:

EMG C-12 16 RO What to isolate and what to look for in LOCA outside Topic:

containment C-12 RO Importance Rating: 3.5 SRO Importance Rating: 3.9 K/A Number: E04 EK 1.3 Comments: BANK - Indian Point Lesson Plan Objective: LO1732333 R3, EXPLAIN major actions accomplished by procedure EMG C-12.

Tier # 1 Group # 1 Last Used - 2010 Indian Point NRC exam Comprehension 55.41 part 7, 10 KA - LOCA outside containment - Knowledge of the operational implications of the following concepts as they apply to the LOCA outside containment - Annunciators and conditions indicating signals and remedial actions associated with the LOCA - Safety function 3 Modification History:

Added justification items 4/25/15 Rev 0

17 ID: 98332 Points: 1.00 The plant was tripped from full power due to a LOCA outside of containment. The crew has transitioned to EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION.

Which ONE of the following signals will the crew reset while in this procedure AND why?

A. P-4 (reactor trip) to allow reset / re-arming of SI.

B. Phase A isolation to restore control of valves to operators.

C. ONLY the SI signal to prevent ECCS pumps from restarting when they are placed in Normal-After-Stop.

D. ONLY the SI (RWST) signal to prevent possible loss of RWST inventory to the containment sump and damage to the RHR pumps.

Answer: D Answer Explanation:

System knowledge of an SI signal and the RWST water level coming down makes this question not a procedure knowledge question but understanding how the system works. The SI reset which was performed in the previous procedures that sent you to C-11, is only part of the swapover for the RWST reset logic.

There is a separate switch for the SI (RWST) swapover that now needs to be reset or the swapover will still occur. The original SI reset just allowed ECCS pumps and valves to be operated as normal without the signal changing any operations of the equipment back to its safeguards lineup.

Correct - Since the SI signal has been reset this will allow the reset of the SI (RWST) signal now to stop the auto swapover to the CTMT sump (that is empty) which would allow RWST water to be lost from injection and damage to the RHR pumps due to no suction source.

Incorrect - only the SI signal. This has been reset in other procedures prior to this point but if this were reset (which is needed to reset the SI (RWST) signal then this would only be part of the signals that need reset in this procedure.

Plausible if the student doesn't understand how the SI and the swapover for the RWST signals work together since the LOCA is outside of CTMT there is no water in the sump for the RHR pump to use if an auto swap were to occur.

Incorrect - P-4. This is an active signal since the plant was tripped. Plausible if the student thinks that P-4 needs to be reset to allow for a second SI for this accident.

Incorrect - Phase A. This is an active signal since the plant was tripped from full power. Plausible if the student thinks that the phase A signal is preventing the proper positioning of valves.

Meets the K/A because it asks for the knowledge of interrelations between the loss of recirc and a component (RHR). Since this procedure assumes there is no water in the CTMT sump protecting the RHR pumps is an interrelation for the loss of recirc.

Rev 0

RO knowledge because it does not ask for the procedure step it only puts the crew in the procedure but asks what signals should be reset to protect what equipment. This is system knowledge of the SI signal and the RWST swapover signal.

Question 17 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98332 User-Defined ID: 98332

Reference:

EMG C-11 Topic: 17 RO which signals get reset in C-11 and why RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: E11 EK 2.1 Comments: NEW Lesson Plan Objective: LO1732332 R3, SUMMARIZE the major action categories and the bases for the steps that accomplish each category.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Loss of emergency coolant recirc - Knowledge of the interrelations between the loss of emergency coolant recirc and the following - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features - Safety function 4 Modification History:

0 - revised based on feedback from Scott 4/25/15 Rev 0

18 ID: 98333 Points: 1.00 Given the following plant conditions:

• The plant is at 100% power

• DG 'A' has been paralleled with bus NB01 and is carrying 5.8 MWe of load in accordance with STS KJ-015A, MANUAL/AUTO FAST START SYNCHRONIZATION AND LOADING OF EDG NE01

• ANN 019E, NB01 BUS DGRD VOLT, alarms

• 55 seconds later, BKR 13-48, #7 XFMR OUTPUT BKR, trips open With NO Operator actions, what is the current status of the 'A' Train Safeguards Power system?

A. NB01 Normal Feeder Breaker is CLOSED, NE01 will stop.

B. NB01 Emergency Supply Breaker is OPEN, NE01 will stop.

C. NB01 Normal Feeder Breaker is CLOSED, NE01 will remain running.

D. NB01 Emergency Supply Breaker is OPEN, NE01 will remain running.

Answer: C Answer Explanation:

OFN NB-042, SY1406401 R7 Correct - NB01 Normal Feeder Breaker will remain CLOSED, NE01 will remain running. Once the 13-48 breaker detaches the NB bus from the switchyard, the NB voltage returns to normal values (or a little higher), and the Normal supply breaker remains shut. The EDG remains running despite the load reject.

Incorrect - NB01 Normal Feeder Breaker will remain CLOSED, NE01 will stop.

Right breaker position, D/G will not stop. The loss of load that the EDG will experience when the BKR 13-48 Normal supply breaker opens will not overspeed the EDG.

Incorrect - NB01 Emergency Supply Breaker will OPEN, NE01 will stop. Breaker will remain closed, D/G will not stop. -- There is nothing in the current condition to cause the EDG output breaker to open, and the EDG will not overspeed or stop itself.

Incorrect - NB01 Emergency Supply Breaker will OPEN, NE01 will remain running. Wrong breaker position, right NE01 status. There is nothing in the current condition to cause the EDG output breaker to open.

Comments: ALR 00-019E states that if the EDG is in parallel then to stop parallel operation. OFN NB-042, LOSS OF OFFSITE POWER TO NB01 (NB02)

WITH EDG PARALLELED requires removing power from the switchyard first and then opening the EDG breaker to cause the UV on the bus for emergency operation.

Rev 0

Additional insights: This question is essentially the basis for why OFN NB-042 was created and why we keep a dedicated operator during diesel runs with the engine paralleled. The issue is that an upstream transient may be isolated prior to the bus feeder breaker. If the bus feeder breaker remains closed, the diesel will not shift to emergency mode. Without an SI present, 55 seconds is not long enough for the degraded voltage condition to open the bus feeder breaker. The transformer feeder breaker opens, but there is no transformer lockout. The diesel is tested by surveillance to withstand up to full load reject. The EDG will restore voltage once divorced from the degraded offsite source. The voltage will likely go high, but not excessively due to the power factor limitations in the STS procedure. The operators will perform the actions of OFN NB-042 to place the EDG in emergency mode.

Meets the K/A because it asks for ability to interpret grid disturbances and status of the EDG RO knowledge since this is system understanding of the output breaker of the EDG and NB01 normal power supply breaker.

Rev 0

Question 18 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98333 User-Defined ID: 98333

Reference:

OFN NB-042 18 RO EDG status after loss of offsite power to NB bus with Topic:

EDG on bus RO Importance Rating: 3.9 SRO Importance Rating: 4.3 K/A Number: 077 AA 2.09 Comments: BANK - 45856 Lesson Plan Objective: SY1406401 R7, INTEGRATE system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 1 Group # 1 Last Used - 2011 Wolf Creek NRC Comprehension 55.41 part 7 KA - Generator voltage and electric grid disturbances -

Ability to determine and interpret the following as they apply to generator voltage and electric grid disturbances -

Operational status of emergency diesel generators - Safety function 6 Modification History:

added to pedigree Rev 0

19 ID: 98334 Points: 1.00 Given the following:

• Unit is operating at 100% power

• Main Turbine is operating in Open Loop mode

• Rod Control selector switch is in Auto

• Control Bank 'D' rods are parked at 231 steps

• Control Rod B10 drops fully into the core

• NO operator actions have been taken Which of the following statements is correct with regard to this event AND why?

A. Tave will return to program due to rods stepping OUT.

B. RCS pressure will remain lower than program due to the lower Tave.

C. Main Turbine electrical load lowered due to lower S/G pressures.

D. Tref returns to previous value due to control valves not changing position.

Answer: C Answer Explanation:

SY1511701 4.1.1 Open Loop/Valve Management In OPEN LOOP, the main turbine control system will maintain a steady control valve position without any feedback mechanism. Ovation will position the turbine control valves as needed to establish its interpretation of megawatts based on valve position curves developed by the vendor. Valve curves have been updated and more will be more finely tuned by DCP 14452 (Rev 1) software upgrade.

With a dropped rod at 100% power Tave will lower causing auto rod motion OUT.

Since D bank rods are already at 231 they will only move out to 232 and stop (C-11). With this small rod step and the fact that rods don't have much 'bite' at this height Tave will not return to target and will stay lower than program. This lower Tave will cause the SG pressures to lower thereby lowering overall steam flow to the main turbine. Since the turbine is in open loop mode the control valves will receive no feedback to change valve position so the overall affect is lower electrical output with no operator action.

Correct - main turbine load will lower due to lower SG pressures due to lower Tave Incorrect - Tave will return to program. Tave will remain lower since control rods can only step 4 steps and don't affect temperature much at this height. Tave will stay lower even with the rods stepping out. Plausible if the student thinks that rods will return temperature back to program.

Rev 0

Incorrect - RCS pressure will remain lower than program. Pressure is dictated by PZR program from the master pressure controller which is not affected by the dropped rod or any other things that happen due to this. Plausible if the student confuses the pressure controller with the level which is controlled from Tref and Tave.

Incorrect - Tref returns to previous value. Plausible if the student confuses what controls Tref. Tref is controlled by first stage impulse pressure which will remain lower due to lower SG pressure and valves not moving in this mode of operation Meets K/A because it asks for knowledge of rods and turbine load changes (reason)

RO knowledge because this is fundamental understanding of the reactor and the secondary plant Question 19 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98334 User-Defined ID: 98334

Reference:

LO1130641 Topic: 19 RO main turbine impacts from a dropped rod RO Importance Rating: 3.2 SRO Importance Rating: 3.7 K/A Number: 003 AK 1.01 Comments: NEW Lesson Plan Objective: LO1130641 R22, Explain reactor response to a control rod insertion.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 5 KA - Dropped control rod - Knowledge of the operational implications of the following concepts as they apply to dropped control rod - Reason for turbine following reactor on dropped rod event - Safety function 1 Modification History:

Rev 0

20 ID: 98335 Points: 1.00 Given the following:

• Plant is in MODE 5.

• Chemistry reports that RCS boron concentration is now 150 ppm lower than the previous sample.

• The crew recognizes that Shutdown Margin is NOT being met, and initiates an emergency boration using 'A' BA Transfer pump.

• The RO reports BG FI-183A, EMERG BORATE FLOW, indicates 25 gpm.

Based on this information, the crew will A. align the RWST to the CCP suction.

B. check when emergency boration can be stopped.

C. align a Safety Injection Pump for Emergency Boration.

D. check if normal letdown can be established with current plant conditions.

Answer: A Answer Explanation:

Correct - if flow is less than 30 gpm then the procedure directs using the RWST as a suction source Incorrect - check when emergency boration can be stopped. Plausible if the student thinks this is enough flow from the emergency boration flow path.

Incorrect - Use an SI pump. This alignment is used when Charging System is NOT aligned as operable flowpath (step 1) and since the stem indicates the crew has passed that step because they have indicated flow. Plausible if the student confuses the low flow with the need to use a different pump.

Incorrect - check if normal letdown can be established. Plausible if the student thinks the emergency boration flow rate is good and wants to continue to put the system back in a normal alignment.

Meets K/A because after the BA pump is started the ability to monitor its normal operation is needed to understand that 25 gpm is not enough flow from it so it is broke and the procedure has the crew use a different pump for this.

RO knowledge to understand that anything less than 30 gpm for emergency boration is to low and the next source needed.

Rev 0

Question 20 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98335 User-Defined ID: 98335

Reference:

OFN BG-009 20 RO emergency boration due to loss of SDM while in Topic:

mode 5 RO Importance Rating: 3.7 SRO Importance Rating: 3.5 K/A Number: 024 AA 1.02 Comments: BANK - 98095 Lesson Plan Objective: LO1732419 R3, Given a procedural flow path, EXAMINE the available options for procedure actions.

Tier # 1 Group # 2 Last Used - N/A Memory 55.41 part 10 KA - Emergency boration - Ability to operate and or monitor the following as they apply to emergency boration - Boric acid pump - Safety function 1 Modification History:

added to pedigree Rev 0

21 ID: 98336 Points: 1.00 A reactor startup is in progress. The RO has just completed pulling Control Rods and is checking nuclear instrumentation. The following indications are observed:

Based on the observed indications above which detector is malfunctioning AND can the startup continue?

A. SR NI N-31, NO B. SR NI N-32, YES C. SR NI N-31, YES D. SR NI N-32, NO Rev 0

Answer: A Answer Explanation:

Proper overlap is SR at ~ 2X104 and IR ~ 1X10-10.

OFN SB-008, attachment P step p12 if you can block the channel and power is above P-6 then continue the startup and fix it later.

Correct - N-31 is reading too low compared to the others.

Incorrect - N-32 and NO. It is showing proper overlap with both the IR detectors for this power. Plausible if the student doesn't know where the overlap between the SR and IR is. Second part is correct.

Incorrect - N-32 and YES. This overlap is correct based on both of the other IR detectors agreeing with it. Per the OFN the start up must be stopped. Plausible if the student misunderstands overlap and TS for the SR detectors.

Incorrect - N31 and YES. Correct detector but per the OFN the startup must be stopped. Plausible if the student misunderstands that the detector is still needed at this power level per TS.

Meets the K/A because it asks for ability to determine SR and IR overlap with the malfunction of a SR detector RO knowledge since this is system knowledge of the excore NIs Rev 0

Rev 0 Excore NIS Ranges "ABSOLUTE" POWER INTERMEDIATE POWER POST-ACCIDENT RANGE RANGE POWER RANGE 4

10

-3 2 10 10  % 100 100

-4 10 P 50 10

-5 W 0 10 10 SOURCE A R 0 1 RANGE -6 M 10 -1 P POST-ACCIDENT 10

-7  %

% -2 E 10 SOURCE RANGE -2 P 10 10 R -8 P O 6 10 -3 10 E O 10 W S -9

-4 5 10 5 W -4 E 10 10 10 10

-10 E R 4 -5 4 10 R 10 10 10 C -11

-6 3 10 3 -6 10 P 10 10 10 S C 2 2 -7 10 P 10 10

-8 1 S 1 -8 10 10 10 10 0

10 1

-10 -1 10 10 Westinghouse Gamma Metrics For Training Use Only K:\TRNG_CommonDrawings\SE\SE01.vsd

Question 21 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98336 User-Defined ID: 98336

Reference:

SY1301501 Topic: 21 RO SR overlap with IR RO Importance Rating: 3.1 SRO Importance Rating: 3.5 K/A Number: 032 AA 2.04 Comments: NEW Lesson Plan Objective: SY1301501 R3, Explain the operation of the Excore Nuclear Instrumentation System Source Range channels.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 7 KA - Loss of SR NI - Ability to determine and interpret the following as they apply to the loss of SR NI - Satisfactory source range intermediate range overlap - Safety function 7 Modification History:

0 - modified the stem and distractors based on Scotts feedback 4/25/15 Rev 0

22 ID: 98337 Points: 1.00 Given the following:

All Shutdown rods have been withdrawn An Intermediate Range NI detector fails high The Level Trip Bypass Switch is placed in Bypass After the switch is placed in bypass, annunciator 82D, IR HI FLUX ROD STOP, light will be ________ AND the intermediate range 'instrument power on' light will be ________.

A. ON, OFF B. OFF, OFF C. ON, ON D. OFF, ON Answer: D Answer Explanation:

Ran on desktop simulator to determine which alarms were received and not received.

Correct - After the bypass switch is placed in bypass the IR rod stop alarm will clear and the instrument will still have power so the power on light would still be lit.

Incorrect - OFF, OFF. Plausible if the student doesn't understand the function of the bypass switch which is to bypass the trips and rod stops from the IR detector Incorrect - ON, OFF. Plausible if the student reverses the logic for the use of this switch.

Incorrect - ON, ON. Plausible if the student doesn't understand the function of the bypass switch and thinks that the switch only is used to test the detector.

Meets K/A because asks ability to use the bypass switch RO knowledge because system knowledge Rev 0

Question 22 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98337 User-Defined ID: 98337

Reference:

SY1301501 Topic: 22 RO IR trip bypass switch understanding RO Importance Rating: 3.0 SRO Importance Rating: 3.1 K/A Number: 033 AA 1.02 Comments: NEW Lesson Plan Objective: SY1301501 R4, Explain the operation of the intermediate range nuclear instrumentation detector.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 2 KA - Loss of IR NI - Ability to operate and or monitor the following as they apply to the loss of IR NI - Level trip bypass - Safety function 7 Modification History:

0 - changed based on Scotts feedback 4/25/15 Rev 0

23 ID: 98338 Points: 1.00 Which of the following ALARMING radiation monitors is listed as an entry condition for OFN BB-006, HIGH REACTOR COOLANT ACTIVITY?

A. SJ RE-01, CVCS Letdown Monitor B. SD RE-39, Reactor Seal Table Area Monitor C. GT RE-59, CTMT High Range Area Monitor D. GT RE-31, Containment Atmosphere Monitor Answer: A Answer Explanation:

Correct - this monitor is listed as the only rad monitor that is a direct entry into the off normal procedure. It monitors the letdown line for activity in the RCS water which would be an indication of high RCS activity.

Incorrect - GT RE-59. This monitor is the containment high range rad monitor whose function is to give rad levels in the event of a LOCA inside containment.

This monitor would show higher than normal radiation levels but since it only shows a high range the letdown monitor would alarm well prior to this one showing any change since the bottom range is 1 R. Plausible if the student confuses monitor functions.

Incorrect - SD RE-39. This monitor checks for high radiation at the reactor seal table which would show higher than normal rad levels but is not intended to be an RCS activity monitor. Plausible if the student confuses monitor functions.

Incorrect - GT RE-31. This monitor checks for high containment airborne activity.

This is an atmosphere monitor so this would look for a leak of RCS water into containment. Plausible if the student confuses monitor functions.

Meets the K/A because it asks for the monitor that the procedure will look for to enter.

RO knowledge system understanding and procedure entry conditions.

Rev 0

Question 23 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98338 User-Defined ID: 98338

Reference:

OFN BB-006 23 RO which rad monitor confirms high reactor coolant Topic:

activity RO Importance Rating: 2.6 SRO Importance Rating: 3.0 K/A Number: 076 AK 2.01 Comments: BANK - 59116 Lesson Plan Objective: LO1732416 R1, IDENTIFY the procedure entry conditions.

Tier # 1 Group # 2 Last Used - 2007 Callaway Fundamental 55.41 part 11 KA - High reactor coolant activity - Knowledge of the interrelations between the high reactor coolant activity and the following - Process radiation monitors. Safety function 9

Modification History:

0 - feedback from Scott 4/25/15 Rev 0

24 ID: 98339 Points: 1.00 OFN RP-017, CONTROL ROOM EVACUATION, has been entered by the CRS due to a fire in the Control Room.

Which of the following is vital equipment that the RO will operate for this condition AND why?

A. Borate the RCS to ensure proper SDM exists.

B. Locally line up CCW to containment to ensure thermal barriers have cooling flow.

C. Ensure NK4421, BB PCV-456A PORV, breaker is CLOSED to allow for auto cycling to control RCS pressure.

D. OPEN NK4401, BUS NB02 BRKR CONTROL POWER, to prevent 'B' train equipment from loading on the bus due to possible hot shorts.

Answer: D Answer Explanation:

OFN RP-017 has the crew shutdown the plant and maintain it in a hot standby condition until subsequent actions can be taken to place the plant in the desired long term condition. OFN RP-017A is used to cooldown the plant if desired.

Correct - step C2 (immediate action step for the RO) has them open breakers to prevent control systems and buses from operating erratically due to hot shorts caused by the fire.

Incorrect - Borate the RCS to ensure proper SDM. This step will be performed in OFN RP-017A if the plant is to be cooled down and placed in cold shutdown.

Plausible if the student confuses which steps are performed in which procedure.

Incorrect - locally line up CCW to containment. This step will be performed in OFN RP-017A if the plant is to be cooled down and placed in cold shutdown.

Plausible if the student confuses which steps are performed in which procedure.

Incorrect - Close PORV breaker. This breaker is OPENED by the RO in the immediate actions steps to prevent them from cycling due to hot shorts.

Plausible if the student doesn't understand the intent of this procedure.

Meets KA asks what vital equipment will be operated during a plant fire.

RO knowledge these steps are in the immediate action section of the procedure.

Rev 0

Question 24 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98339 User-Defined ID: 98339

Reference:

OFN RP-017 REV 46 Topic: 24 RO vital equipment operated in case of fire RO Importance Rating: 3.3 SRO Importance Rating: 4.0 K/A Number: 067 AA 2.16 Comments: NEW Lesson Plan Objective: LO1732427, R2, RECOGNIZE the available situations which are addressed by procedure OFN RP-017.

Tier # 1 Group # 2 Last Used - N/A Memory 55.41 part KA - Plant fire on site - Ability to determine and interpret the following as they apply to the plant fire on site - Vital equipment and control systems to be maintained and operated during a fire - Safety function 8 Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

25 ID: 98340 Points: 1.00 During the performance of EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE, which of the following valves will be checked closed AND why?

A. KC HV-253, Fire Protection System Header Outside CTMT Isolation To ensure all non-essential containment penetrations are closed to prevent the release of radioactive materials.

B. EG HV-058, CCW To RCS Outer CTMT Isolation Valve To ensure all the containment penetrations are isolated since this could be the source of the high containment pressure.

C. EG HV-058, CCW To RCS Outer CTMT Isolation Valve To ensure all non-essential containment penetrations are closed to prevent the release of radioactive materials.

D. KC HV-253, Fire Protection System Header Outside CTMT Isolation To ensure all the containment penetrations are isolated since this could be the source of the high containment pressure.

Answer: A Answer Explanation:

Correct - per FR-Z1 basis for step 1 and 2 the reason for ensuring the valves are isolated is to prevent a radioactive release from a non-essential containment penetration. CISA closes the fire header valve (this valve is manually closed during normal ops but still receives a signal to close)

Incorrect - EG HV-058 and prevent the release of rad materials. Phase B will isolate the CCW system from containment not phase A. Plausible if the student confuses which signal closes which valves. The reason is correct.

Incorrect - EG HV-058 and pens closed for high pressure. Phase B will isolate the CCW system from containment not phase A. Plausible if the student confuses which signal closes which valves. The reason is plausible since water injected from any system could raise CTMT pressure.

Incorrect - KC HV-253 and pens closed for high pressure. Correct valve wrong reason. Plausible as any injection from any water source could cause CTMT pressure to rise.

Meets K/A asks for a loss of CTMT integrity what is checked for rad control, ability to control the release by ensuring valves are in correct position for event.

RO knowledge due to step basis of EMGs Rev 0

Question 25 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98340 User-Defined ID: 98340

Reference:

BD EMG FR-Z1 25 RO Basis for closing phase A valves on high Topic:

containment pressure RO Importance Rating: 3.8 SRO Importance Rating: 4.3 K/A Number: E14 2.3.11 Comments: NEW Lesson Plan Objective: LO1732350 R3, DISCUSS the major action steps for procedure EMG FR-Z1.

Tier # 1 Group # 2 Last Used - N/A Fundamental 55.41 part 9 KA - Loss of containment integrity (high pressure) - Rad control - Ability to control radiation release - Safety function 5

Modification History:

Rev 0

26 ID: 98341 Points: 1.00 The crew is performing EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION, with the following conditions:

RCS pressure is 585 psig down slow RCS temperature is 404°F down slow PZR level is 50% stable RCP A and D are running RCP B and C are stopped Both CCPs are running A SI pump is running B SI pump is tagged out for maintenance Both RHR pumps are stopped Containment pressure is 6.5 psig stable The crew is determining if CCPs and SI pumps should be stopped Using the attached reference, determine the CCP and SI pump configuration after the procedure steps are completed based on the given indications?

A. The B CCP will be running with the A SI pump running.

B. The A CCP will be running with NO SI pumps running.

C. The A CCP will be running with the A SI pump running.

D. The B CCP will be running with NO SI pumps running.

Answer: A Answer Explanation:

To get to this point in this procedure you will have had to go from E-0 to E-1 to ES-11. All prior procedures were checked to determine the appropriate indications given for this setup.

RCS subcooling - with pressure at 585 psig saturation temperature is 486°F so 486°F - 404°F = 82°F of subcooling In ES-11 steps 21 and 22 Correct - 'B' CCP and A SI pump. There are two CCPs running so the step will be performed. There are two RCP's running and one SI pump running and with adverse containment 70°F of subcooling is needed to stop the A CCP. There is a note stating that pumps should be stopped in alternate trains so you would not stop the B CCP with the B SI pump tagged out. The next step asks if any SI pumps are running which is yes, so with one CCP running and two RCP's running and one SI pump running 150°F of subcooling is needed. Subcooling is not met so step 22.c RNO directs the crew if RCS temperature is greater than 375°F then go to step 40 and not to stop the running SI pump.

The procedure step states pumps should be stopped in opposite trains. Without a reason for not stopping pumps in this manner they will be.

Rev 0

Incorrect - B CCP with NO SI pumps. There is a note that states pumps should be stopped in alternate trains so the B CCP will be the pump that is stopped. SI pump will not be stopped because subcooling is not met. Plausible if the student fails to use the adverse containment subcooling value and chooses to stop the pump.

Incorrect - A CCP with NO SI pumps. A CCP is correct. SI pump will not be stopped because subcooling is not met. Plausible if the student fails to use the adverse containment subcooling value and chooses to stop the pump.

Incorrect - B CCP with A SI pump running. There is a note that states pumps should be stopped in alternate trains so the B CCP will be the pump that is stopped. The SI pump is correct Meets K/A uses procedure steps to determine if operator can perform steps of LOCA cooldown RO knowledge procedure is given and steps are performed Rev 0

Question 26 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98341 User-Defined ID: 98341

Reference:

EMG ES-11 Topic: 26 RO Which ECCS pumps to stop in ES-11 RO Importance Rating: 3.9 SRO Importance Rating: 4.2 K/A Number: E03 2.1.25 Comments: Handout provided NEW Lesson Plan Objective: LO1732321 R3, DISCUSS the major action steps of procedure EMG ES-11.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.41 part 10 KA - LOCA cooldown and depressurization - Conduct of ops - Ability to interpret reference materials such as graphs curves and tables etc - Safety function 4

• Reference provided EMG ES-11 steps 21 and 22 Modification History:

Rev 0

27 ID: 98342 Points: 1.00 The unit has tripped due to a LOCA and ESF equipment has failed to start.

EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING, has been entered.

A depressurization of the Steam Generators (SGs) is being performed in accordance with EMG FR-C2, when the STA reports that there is a Red Path on the Integrity CSF Status Tree.

Which ONE (1) of the following describes the actions that will be taken?

A. Complete the S/G depressurization and then transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS, if the red path still exists B. Immediately transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS C. Complete FR-C2 and then transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS, if the red path still exists D. Stop the S/G depressurization and, if the red path does NOT clear, transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS Answer: C Answer Explanation:

Correct - Because of the SG depressurization, a Red condition on Integrity is expected. If this is not performed and a transition is made then the overall affect would end up being entry into C-1 inadequate core cooling and that is worse than staying in C-2. There is a NOTE in C2 telling the crew NOT to go to P1 if a red condition exists.

Incorrect - immediately transition to P-1. Plausible if the student misunderstands the direction in the note with respect to the direction in the procedure users guide about higher tear procedures.

Incorrect - stop the SG depressurization. Plausible since the action the operator did caused the red path so if they stop it and the red path clears then they might want to stay. Procedure use and adherence understanding.

Incorrect - complete the SG depressurization. Plausible since the FR procedure directed the action after the action then the student may think a transition now needs to be completed.

Meets KA asks about PTS with regards to cooldown and depressurization using SG at max rate RO knowledge EMG procedure note basis Rev 0

Question 27 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98342 User-Defined ID: 98342

Reference:

BD EMG FR-C2 27 RO FR-C2 note to not go to P-1 on red path until after Topic:

C-2 is complete RO Importance Rating: 3.6 SRO Importance Rating: 4.0 K/A Number: E08 EK 2.2 Comments: BANK - 59338 Lesson Plan Objective: LO1732341 R9, EXPLAIN the bases and knowledge requirements for selected procedure steps of EMG FR-C2.

Tier # 1 Group # 2 Last Used - 2007 Callaway exam Comprehension 55.41 part 10 KA - PTS - Knowledge of the interrelations between the PTS and the following - Facility heat removal systems including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility -

Safety function 4 Modification History:

Rev 0

28 ID: 98343 Points: 1.00 The plant is operating at 45% power, steady state when 'A' RCP experiences an Under-Frequency trip due to an electrical fault. The PA buses continue to operate as normal.

Which of the following describes the 'A' loop RCS temperatures and 'A' S/G response from time 0 to 1 minute after the RCP trip?

(assume NO operator actions)

A. Initially to several seconds later there is little change in Thot and Tcold, after that out to 1 minute the difference is lowering rapidly.

S/G pressure is higher than pre-trip and steam flow is higher.

B. Initially to several seconds later the difference between Thot and Tcold is changing rapidly, after that out to 1 minute the change is smaller.

S/G pressure is higher than pre-trip and steam flow is higher.

C. Initially to several seconds later there is little change in Thot and Tcold, after that out to 1 minute the difference is lowering rapidly.

S/G pressure is the same as pre-trip and steam flow is close to 0.

D. Initially to several seconds later the difference between Thot and Tcold is changing rapidly, after that out to 1 minute the change is smaller.

S/G pressure is the same as pre-trip and steam flow is close to 0.

Answer: C Answer Explanation:

Ran this on the desktop simulator to gather information from. From the time the RCP trip was inserted the RCS loop temperatures slowly converged toward each other. After the flywheel lost forward inertia the temperature difference rapidly converged and then Thot ended up lower than Tcold because of reverse flow in the loop and the SG still removing some heat (not much). The SG pressures were equal to the other SGs since they are all still connected together but the steam flow from the affected loop went down to near 0. Flow from the other pumps goes up due to the loss of the flow from the affected loop.

Correct - The response of Th and Tc is due to the design coast down of the RCPs which lasts approximately 1-1.5 minutes (30 seconds per TS). Although both temperatures will be lower, Th will lower faster than Tc due to the sudden, significant reduction in heat generated by the reactor going to that SG. SG pressure will be relatively stable Incorrect - time 0 little change then rapid change and SG pressure higher. The first part is correct. The affected SG is now not producing any power for the turbine so the other SG must make up the loss of steam flow from this loop so overall pressure is lower not higher. Plausible if the student thinks that the loss of steam flow is because the pressure is higher.

Incorrect - time 0 rapid change then Th and Tc equal and SG pressure same.

The SG pressure is correct. Plausible if the student forgets about the flywheel effect on the RCS indications.

Rev 0

Incorrect - time 0 rapid change then Th and Tc equal and SG pressure higher.

Plausible if the student forgets about the flywheels effect on the RCS indications.

Also the affected SG is now not producing any power for the turbine so the other SG must make up the loss of steam flow from this loop so overall pressure is lower not higher. Plausible if the student thinks that the loss of steam flow is because the pressure is higher.

Meets the KA because it asks for RCS parameters after a loss of an RCP (coastdown with the flywheel)

RO knowledge since this is system understanding of how the flywheel on the RCP will affect the other RCS indications after the loss of the pump Question 28 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98343 User-Defined ID: 98343

Reference:

SY1300300 28 RO RCS temperature response to loss of RCP's during Topic:

the first minute RO Importance Rating: 2.8 SRO Importance Rating: 3.2 K/A Number: 003 K 5.02 Comments: Modified - Millstone Lesson Plan Objective: SY1300300 R9, DETERMINE the operational implications that a loss of flow will have on the core operating parameters.

Tier # 2 Group # 1 Last Used - Millstone 2011 Comprehension 55.41 part 2, 3 KA - RCP - Knowledge of the operational implications of the following concepts as they apply to the RCPs - Effects of RCP coastdown on RCS parameters Modification History:

0 - changed based on feedback from Scott 4/25/15 Rev 0

29 ID: 98344 Points: 1.00 Given the following:

The plant is recovering from a forced outage Core is at middle of life The crew is performing a reactor startup per GEN 00-003 Reactor power is 1 X 10-8 amps and stable Auto makeup to the VCT occurs During the auto makeup, BG FK-110, BA FLOW CTRL, was inadvertently set at 0.0 turns With NO operator action which of the following describes the reactor power response to this event?

A. Reactor power will continue to rise until trip on IR high flux.

B. Reactor power will continue to rise too slightly above the POAH.

C. Reactor power will continue to lower until subcriticality is reached.

D. Reactor power will remain constant because charging is flowing to RCP seals ONLY at this point.

Answer: B Answer Explanation:

Correct - At this power level there is no feedback from temperature of the RCS.

As the dilution continues power will rise until the POAH is reached at then power will level off and stay there.

Incorrect - no plant response since charging to seals only. This is a possible lineup for the RCP seals and charging. Plausible if the student mistakes the current normal plant lineup.

Incorrect - lower power until subcritical. This would be true if the setting on the pot was higher than the current RCS boron concentration. Plausible if the student forgets which way the controller works.

Incorrect - trip on IR high flux. If power were to rise above the POAH and did not turn this would happen next. Plausible if the student mistakes the dilution and no operator action with the low power trips and forgets the POAH effects of temperature.

Meets the KA by asking the operational implications of a dilution from the CVCS system. This has a part about understanding that the pot has an effect on this (dilution) but if the stem states a dilution is in progress then the question will be a LOD1 RO knowledge based on system response for a dilution.

Rev 0

Question 29 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98344 User-Defined ID: 98344

Reference:

LO1130641 Topic: 29 RO dilution effect on the reactor without operator action RO Importance Rating: 3.6 SRO Importance Rating: 3.7 K/A Number: 004 K 5.20 Comments: MODIFIED - PI Lesson Plan Objective: LO1130641 R12, Discuss the concept of the point of adding heat (POAH) and its relationship to reactor power.

SY1300400 R17 Tier # 2 Group # 1 Last Used - 2010 Prairie Island Comprehension 55.41 part 1 KA - CVCS - Knowledge of the operational implications of the following concepts as they apply to the CVCS -

Reactivity effects of xenon, boration, and dilution Modification History:

0 - feedback from Scott 4/25/15 Rev 0

30 ID: 98345 Points: 1.00 The plant is operating at 50% power and stable. An electrical failure in controller BG PK-131, LETDOWN HEAT EXCHANGER OUTLET PRESSURE CONTROL, has caused the controller to fail to 0% output.

Which of the following lists the changes in the indications for this failure?

LTDN HX OUTLET LTDN HX OUTLET FLOW PRESS A. DOWN UP B. UP UP C. UP DOWN D. DOWN DOWN Answer: A Answer Explanation:

Correct- With a failure of the BG PK-131 controller to 0% demand this closes the valve fully. This will cause the letdown relief valve inside containment to lift, causing all letdown flow to be directed to the PRT. This will cause pressure to go up and flow back to the VCT (outlet flow) to go down.

Incorrect - Down, Down. Plausible if the student doesn't understand how the system works under failure modes.

Incorrect - Up, Up. Plausible if the student doesn't understand how the system works under failure modes.

Incorrect - Up, Down. This is exactly backwards if the student thinks that 0% is with the valve open. Plausible if the student confuses what 0% means for the valve position.

Meets the KA asks for ability to monitor auto ops of CVCS during letdown RO knowledge since this asks for system understanding Rev 0

Question 30 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98345 User-Defined ID: 98345

Reference:

SY1300400 Topic: 30 RO CVCS indications with BG PF-131 failure closed RO Importance Rating: 3.6 SRO Importance Rating: 3.4 K/A Number: 004 A 3.11 Comments: NEW Lesson Plan Objective: SY1300400 R24, Integrate system and plant response to transient and equipment failures, including interactions with related systems Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - CVCS - Ability to monitor auto operation of the CVCS including - Charging and letdown Modification History:

Rev 0

31 ID: 98346 Points: 1.00 Which of the following busses directly supply power to RHR Pumps 'A' and 'B'?

A RHR Pump B RHR Pump A. NN03 NN04 B. NG01 NG02 C. NB01 NB02 D. PB03 PB04 Answer: C Answer Explanation:

Obj 2 SY1300500 Correct - NB01 and NB02 is CORRECT. The electrical power supplies for the RHR pumps are as follows:

NB01 RHR pump 'A', (Breaker NB0101).

NB02 RHR pump 'B', (Breaker NB0204).

Incorrect - PB03 and PB04. Plausible because PB buses power the condensate, heater drain, and NCP pumps Incorrect - NG03 and NG04. Plausible because RHR valves are powered from NG01 and NG02 (e.g. EJ HV-8809A from NG01B and EJ HV-8809B from NG02B)

Incorrect - NN03 and NN04. Plausible because RHR instrumentation is powered from NN.

Meets the KA by asking the power supply to the RHR pump RO knowledge based on power supply understanding Rev 0

Question 31 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98346 User-Defined ID: 98346

Reference:

KD-7496 ONE LINE POWER DISTRIB Topic: 31 RO RHR power supplies RO Importance Rating: 3.0 SRO Importance Rating: 3.2 K/A Number: 005 K 2.01 Comments: BANK - 59457 Lesson Plan Objective: SY1300500 R2, Explain the function, location, operation, and electrical interlocks of the major components.

Tier # 2 Group # 1 Last Used - Wolf Creek 2006 Fundamental 55.41 part 8 KA - RHR - Knowledge of bus power supplies to the following - RHR pump Modification History:

Rev 0

32 ID: 98347 Points: 1.00 The unit is at 100% power with the following conditions given for the 'B' Accumulator:

Boron concentration sample 2305 ppm Pressure 485 psig Borated Water Volume 6294 gal Outlet isolation valve OPEN breaker OPEN If a LBLOCA were to now occur on the 'A' loop hot leg what would be the effect (if any) on the RCS?

A. None, the accumulator would inject as design.

B. The low boron concentration could allow for a reactor re-start.

C. N2 injection could impede natural circulation flow for the 'B' loop.

D. Fuel clad could overheat due to NOT enough water injection from the accumulator.

Answer: D Answer Explanation:

Correct - With the low N2 pressure of the accumulator the full volume of water would not inject and the purpose of the accumulator is to flood the core on a LBLOCA until the ECCS pumps have started and begin to inject water to continue to remove heat.

Incorrect - None. Plausible if the student confuses the TS requirements for the accumulator and believes all spec are made which would indicate that it would perform its design function and fully inject Incorrect - Low boron. Plausible if the student confuses the TS requirements for the accumulator and believes the boron is low and the reason for the boron is to prevent a restart.

Incorrect - N2 injection. Plausible if the student confuses the TS requirements for the accumulator and believes this will cause N2 to inject which would impede NC.

Meets the KA asks for effect on the ECCS injection (blowdown phase) of an accumulator that is not within TS (malfunctioned)

RO knowledge system understanding Rev 0

Question 32 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98347 User-Defined ID: 98347

Reference:

TS 3.5.1 Topic: 32 RO accumulator malfunction due to N2 out of TS low RO Importance Rating: 3.4 SRO Importance Rating: 3.9 K/A Number: 006 K 6.02 Comments: NEW Lesson Plan Objective:

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 8 KA - ECCS - Knowledge of the effect of a loss or malfunction on the following will have on the ECCS - Core flood tanks accumulators Modification History:

Replaced 8/25/15 based on Scotts feedback of missing the KA Rev 0

33 ID: 98348 Points: 1.00 The Pressurizer Relief Tank (PRT) can be drained, using no pumps, to the:

A. Waste Holdup Tank B. Recycle Hold Up Tank (RHUT)

C. Instrument Tunnel Sump D. Containment Normal Sump Answer: D Answer Explanation:

M-12BB02 SYS BB-202 Correct - The PRT has a drain line that will drain the tank directly to the containment normal sump Incorrect - RHUT. This is a normal place to put contaminated drains. Plausible if the student forgets the pumps are required to drain the PRT to this location.

Incorrect - Waste Holdup Tank. The path takes water from the PRT through the RCDT pumps then to the waste tank. Plausible if the student forgets the pumps are required to drain the PRT.

Incorrect - Instrument Tunnel sump. Since this is inside containment this is a possible solution but no drains off of the PRT will take water to this location.

Plausible if the student links the PRT location and the reactor cavity sump location together inside containment.

Meets KA by knowledge of physical connection between PRT and containment RO knowledge bases on understanding the system interconnections Rev 0

Question 33 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98348 User-Defined ID: 98348

Reference:

M-12HB01 Topic: 33 RO Where can the PRT be drained to RO Importance Rating: 2.9 SRO Importance Rating: 3.1 K/A Number: 007 K 1.01 Comments: BANK - Indian Point Lesson Plan Objective: SY1300200 R2, EXPLAIN the function and operation of the major RCS components.

Tier # 2 Group # 1 Last Used - Indian Point 2010 Fundamental 55.41 part 3 KA - PRT - Knowledge of the physical connections and or cause effect relationships between the PRT and the following systems - Containment system Modification History:

Rev 0

34 ID: 98349 Points: 1.00 The unit is operating at 100% power when the following takes place:

061B, PROCESS RAD HI, comes in alarm 061A, PROCESS RAD HIHI, comes in alarm shortly after 061B

'B' CCW surge tank vent valve indicates closed

'B' CCW surge tank level is up slow Based on the given indications:

1) Where is the leak coming from?

2) What actions will the crew perform per procedure?

A. 1. Letdown Heat Exchanger

2. Bypass the heat exchanger B. 1. Seal Water Heat Exchanger

2. Bypass the heat exchanger C. 1. Seal Water Heat Exchanger

2. Isolate seal water D. 1. Letdown Heat Exchanger

2. Isolate letdown Answer: D Answer Explanation:

The given conditions are indicative of a leak into the CCW system and since the vent valve has closed the water leaking in is radioactive. The surge tank level rising with no makeup also gives indication that water is leaking into the CCW system. Other indications do change but are not required to diagnose this leak location. Of the two locations given in the distractors for the leak only one is higher pressure than CCW. The seal water heat exchanger is lower pressure so the leak would be lowering surge tank level. Since the water in the seal water heat exchanger is radioactive and the procedure does have the crew bypass this component if the leak is suspected there the distractors are very plausible if the student confuses which pressure is higher.

Correct - Per procedure if the leak is not in the seal water heat exchanger the crew is directed to isolate the leak and return to procedure in effect.

Incorrect - Letdown, bypass. The location is correct but there is no procedure direction to bypass this heat exchanger only to isolate it. Plausible if the student confuses the letdown and seal water heat exchanger procedure directions.

Incorrect - Seal water, isolate. This is the wrong location since this heat exchanger is lower pressure than CCW. The procedure has this heat exchanger bypass and not isolated. Plausible if the student confuses which is at higher pressure and which gets bypassed vs isolated.

Rev 0

Incorrect - Seal water, bypass. This is the wrong location but the correct action for a leak in this heat exchanger. Plausible if the student confuses which heat exchanger is higher pressure since the procedure will direct bypassing the seal water one.

Meets the KA ability to predict how a rad alarm on the CCW system will affect operations and use procedures to address RO knowledge based on system understanding.

Question 34 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98349 User-Defined ID: 98349

Reference:

OFN EG-004 Topic: 34 RO Leak in letdown heat exchanger procedure actions RO Importance Rating: 3.3 SRO Importance Rating: 3.5 K/A Number: 008 A 2.04 Comments: NEW Lesson Plan Objective: SY1400800 R9, DETERMINE the protection afforded by the design of the CCW System for different portions of the system.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 5, 7 KA - Component Cooling Water - Ability to predict the impacts of the following malfunctions or operations on the CCWS and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - PRMS alarm Modification History:

Rev 0

35 ID: 98350 Points: 1.00 Reactor power is at 60% and stable awaiting repairs to the 'B' MFP. The RO notices the following:

• BB ZL-455B, PZR SPRAY LOOP 1 CTRL VLV, both green and red lights lit with controller demand output at 35%

• BB ZL-455C, PZR SPRAY LOOP 2 CTRL VLV, green light lit red light out with controller demand output at 0%

• BB PK-455A, PZR PRESS MASTER CTRL, demand indicates 40% and down slow

• ALR 033C, PZR PRESS LO HTRS ON, is lit Which of the following procedures AND actions will mitigate this event?

A. Enter ALR 033C, PZR PRESS LO HTRS ON, and energize PZR backup heaters.

B. Enter ALR 033C, PZR PRESS LO HTRS ON, trip the reactor and stop 'A' and

'D' RCP's.

C. Enter OFN SB-008, INSTRUMENT MALFUNCTIONS, take manual control of BB PK-455B and CLOSE the valve.

D. Enter OFN SB-008, INSTRUMENT MALFUNCTIONS, and select an alternate PZR pressure channel for control.

Answer: C Answer Explanation:

Correct - with 35% demand and both green and red lights lit on one spray valve RCS pressure will be lowering. OFN SB-008 Att V memory action steps have the RO take manual control of either the spray valve or the PZR pressure master controller and control spray flow, in this case close the spray valve is the only choice since the master controller is at 40% and down slow and the other spray valve is closed. The master pressure controller will not correct this failure.

Incorrect - ALR 033C and trip reactor and stop A and D RCP's. This ALR would be correct for this issue but not to trip the reactor and stop RCP's. Plausible if the student feels no actions taken will stop the pressure from lowering to less than the reactor trip setpoint and stopping RCP's will stop the pressure from dropping any lower.

Incorrect - ALR 033C and energize PZR heaters. This ALR would be correct for this issue but energizing heaters will only at most delay a reactor trip on low pressure if more action is not taken. Plausible if the student feels that the heaters alone will mitigate the pressure lowering.

Rev 0

Incorrect - OFN SB-008 and select alternate pressure channel. Correct procedure but with the indications given a pressure channel has not failed.

Plausible if the student feels that selecting out a failed pressure channel will stop the pressure from lowering any more.

Meets KA by asking how to respond to a PZR malfunction using the procedure RO knowledge since this is a memory action step of the procedure but high cog to interpret what the lights mean Question 35 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98350 User-Defined ID: 98350

Reference:

OFN SB-008 35 RO PZR spray open which procedure to enter and Topic:

action to take RO Importance Rating: 4.3 SRO Importance Rating: 4.4 K/A Number: 010 2.1.23 Comments: MODIFIED - 58685 Lesson Plan Objective: LO1732418 R4, EXPLAIN the plant response for each instrument failure identified in procedure OFN SB-008.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 5, 7, 10 KA - PZR pressure control - Conduct of ops - Ability to perform specific system and integrated plant procedures during all modes of plant operation Modification History:

Rev 0

Q36, Rev 1 ID: 98351 Points: 1.00 The unit has experienced a complete loss of bus NN02. 076A - SSPS B GENERAL WARNING alarm is received.

Which of the following describes why this alarm is received?

A. Due to a loss of a DC power supply on one train of RPS.

B. Due to a loss of an output relay cabinet in RPS.

C. Due to a loss of input relays on one train of RPS.

D. Due to the loss of the ability to trip the 'B' reactor trip breaker.

Answer: A Answer Explanation:

Correct - NN02 supplies power to RPS in three places. Two are the input relays for the white train of logic. The other one is 15/48 VDC logic cabinet power supply to B train only.

Incorrect - output relay cabinet. The output relay cabinets are powered from NN01 and NN04.

Incorrect - input relay on one train. NN02 powers up both trains white channel inputs not just one.

Incorrect - loss of ability to trip the B trip breaker. General warning alarm is associated with the rod control system and trip breakers are the source of power for the rod control system. Plausible if the student puts this together wrong and believes this will prevent the trip breaker opening from the control room.

Meets KA by asking what RPS power supply is lost RO knowledge of power supplies to RPS. High cog because of integrating the NN02 loss and the SSPS B general warning alarm.

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q36, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98351 User-Defined ID: 98351

Reference:

SY1301200 Topic: 36 RO RPS power supply SSPS GW from loss of NN02 RO Importance Rating: 3.3 SRO Importance Rating: 3.7 K/A Number: 012 K 2.01 Comments: NEW Lesson Plan Objective: SY1301200 R5, Explain the RPS operation.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - Reactor protection - knowledge of bus power supplies to the following - RPS channels components and interconnections Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

37 ID: 98352 Points: 1.00 The crew is responding to a reactor trip when it becomes necessary to manually actuate SI. 50 seconds after the crew actuates SI the CRS has the RO reset the SI signal.

Will the SI signal reset AND what effect (if any) will this action have on related plant equipment?

A. YES, H2 fans will need to be manually started.

B. NO, H2 fans will need to be manually started.

C. YES, there will be NO effect on plant equipment.

D. NO, there will be NO effect on plant equipment.

Answer: D Answer Explanation:

M-744-00025 Correct - SI signals once started require a 60 second timer to complete before any other actions with the related equipment can be completed. This also locks out any reset of the SI prior to this timer completing.

Incorrect - No, H2 fan need manual start. The SI signal part is correct you cannot reset it until the timer is complete. H2 fans will start at 60 seconds since there is no input from the SI signal except to start an independent timer that will restart these fans after 60 seconds regardless of any other action taken by the crew.

Incorrect - Yes, H2 fan need manual start. The SI signal will block all reset attempts until its 60 second timer completes. H2 fans will start at 60 seconds since there is no input from the SI signal except to start an independent timer that will restart these fans after 60 seconds regardless of any other action taken by the crew.

Incorrect - Yes, on effect to plant equipment. The SI signal will block all reset attempts until its 60 second timer is complete. Second part is correct plant equipment will continue to operate as design.

Meets the KA by understanding the physical connection of ESFAS and reset RO knowledge system understanding Rev 0

Question 37 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98352 User-Defined ID: 98352

Reference:

M-744-00025 Topic: 37 RO Reset SI before 60 seconds RO Importance Rating: 3.7 SRO Importance Rating: 4.1 K/A Number: 013 K 1.18 Comments: NEW Lesson Plan Objective: SY1301301 R4, Describe Operation Of The Engineered Safety Features Actuation System; Including Automatic Actuation, And Bypass And Reset Operation.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - ESFAS - Knowledge of the physical connections and or cause effect relationship between the ESFAS and the following systems - Premature reset of ESF actuation Modification History:

Rev 0

38 ID: 98353 Points: 1.00 The plant has experienced a LOCA. SI, CISB, and CSAS have all auto actuated as required.

Which ONE of the following Annunciators would direct the Control Room staff to transfer the Containment Spray Pump Suctions to the Recirc Sump?

A. ALR 047B, RWST EMPTY B. ALR 047C, RWST LEV LO-LO 2 C. ALR 047E, RWST LEV HI/LO D. ALR 047D, RWST LEV LO-LO 1 AUTO XFR Answer: B Answer Explanation:

ALR 00-047C, EMG ES-12 Correct - per ES-12 and the ALR 047C if the RWST level is 12% or lower then spray pumps are placed on recirc.

Incorrect - RWST EMPTY. This alarm has the crew stop all pumps taking a suction from the RWST but the tank is not empty it still has 6% in it of usable volume per the ALR. Plausible if the student forgets which alarm provides for which action.

Incorrect - RWST LEV HI/LO. This alarm is if the RWST is lower than the TS value to alert the crew to refill it and also if it is to hi overflowing it will come in.

Plausible if the student forgets which alarm provides for which action.

Incorrect - RWST LO-LO 1 AUTO XFR. This alarm has the crew perform ES-12 to transfer all the ECCS equipment over to the recirc sump. It doesn't have the crew swap spray at this time. Plausible if the student forgets which alarm provides for which action.

Meets the KA asks for ability to monitor changes in RWST level for ESFAS changes RO knowledge this is a setpoint of an alarm Rev 0

Question 38 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98353 User-Defined ID: 98353

Reference:

EMG ES-12 Topic: 38 RO RWST LO LO 2 realignment of spray pumps RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: 013 A 1.06 Comments: BANK - 46455 Lesson Plan Objective: LO1732322 R3, DISCUSS the major action steps of procedure EMG ES-12.

Tier # 2 Group # 1 Last Used - 2009 Callaway Fundamental 55.41 part 7, 10 KA - ESFAS - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the ESFAS controls including - RWST level Modification History:

Rev 0

39 ID: 98354 Points: 1.00 The unit is operating at 100%. The 'B' Cavity Cooling fan was running but its output breaker trips and will NOT reset.

Which of the following is the correct response to this event AND why?

The 'A' Cavity Cooling fan...

A. auto starts to provide cooling to the incore instrumentation vessel connections.

B. must be manually started to provide cooling to the incore instrumentation vessel connections.

C. must be manually started to provide cooling to the excore NIs to prevent possible damage due to high temperatures.

D. auto starts to provide cooling to the excore NIs to prevent possible damage due to high temperatures.

Answer: D Answer Explanation:

SY1302600 Correct - a loss of power to the running fan will auto start the other fan. One of the functions of these fans is to cool the excore NI detectors to a max of 135F.

This allows them to operate properly and give reliable indications. As stated in the NI LP if temperature gets too high for the detectors they will not provide accurate indication.

Incorrect - Auto start and cool the instrument vessel connections. The first part is correct it will auto start. The instrument vessel connections are on top of the head and not cooled by the cavity cooling fans. Plausible if the student confuses where the fans discharge to.

Incorrect - manual start and cools excore NIs. The fans will auto start on a loss of power, that is all that will auto start them. A failure of the running fan will not start its counterpart. The second part is correct. Plausible if the student remembers that the fans don't start on a malfunction of the fan but forgets that loss of power is not a malfunction of the fan.

Incorrect - manual start and cools instrument vessel connections. The fans will auto start on a loss of power, that is all that will auto start them. A failure of the running fan will not start its counterpart. The instrument vessel connections are on top of the head and not cooled by the cavity cooling fans. Plausible if the student forgets what starts the fans and where they discharge to.

Meets the KA by asking what the RO monitors from the control room, cooling fan, and what is the expected response to the plant, the other should auto start.

RO knowledge based on system understanding Rev 0

Question 39 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98354 User-Defined ID: 98354

Reference:

SY 1302600 Topic: 39 RO cavity cooling fan loss of power and function RO Importance Rating: 3.2 SRO Importance Rating: 3.1 K/A Number: 022 A 4.02 Comments: NEW Lesson Plan Objective: SY1302600 R11, DESCRIBE the Containment Cooling System design feature(s) and system interlocks.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 7 KA - Containment cooling system - Ability to manually operate and or monitor in the control room: - CCS pumps Modification History:

Rev 0

40 ID: 98355 Points: 1.00 Review the attached test data sheet from STS EN-100A, CONTAINMENT SPRAY PUMP A INSERVICE PUMP TEST.

Which of the following statements is correct with regard to the information contained on the test data sheet?

The RO should notify the SM/CRS that the 'A' Containment Spray Pump...

A. has failed the surveillance due to pump Dp being too low.

B. static suction pressure is UNSAT but NO actions are required at this time.

C. valve ENV0099 leak rate is UNSAT and a CR should be initiated ONLY.

D. has failed the surveillance due to dynamic suction pressure being too high.

Answer: C Answer Explanation:

AP 29B-003, SURVEILLANCE TESTING, is the admin procedure that governs performing surveillances.

Per the surveillance test data sheet attached there are only one item that are not within the normal range of acceptance. ENV0099 leak rate being too high. Per required action 2 this is only an UNSAT item and the pump is considered OPERABLE and a CR should be initiated ONLY.

Correct - ENV0099 leak rate to high and the note states that only a CR needs to be initiated.

Incorrect - pre suction pressure unsat. Plausible if the student mis reads or gets confused on the table.

Incorrect - failed due to pump Dp. Plausible if the student mis reads or gets confused on the table.

Incorrect - failed due to post suction pressure to high. Plausible if the student mis reads or gets confused on the table.

Meets KA because the surveillance is over the 026 Containment Spray system and is determining that the surveillance for this equipment is or is not being met (knowledge of surveillance procedures).

RO knowledge simple procedure usage.

Rev 0

Question 40 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 98355 User-Defined ID: 98355

Reference:

AP 29B-003 Topic: 40 RO containment spray surveillance test data pass/fail RO Importance Rating: 3.7 SRO Importance Rating: 4.1 K/A Number: 026 2.2.12 Comments: Handout provided NEW Lesson Plan Objective: LO1733214 R8, DEMONSTRATE proper application of the Definitions, Responsibilities and Procedural Requirements associated with AP29B-003, Surveillance Testing.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - Containment spray - Equipment control - Knowledge of surveillance procedures Modification History:

Rev 0

41 ID: 98356 Points: 1.00 An event has occurred in the plant and 'B' S/G pressure has risen to 1217 psig.

The total number of safety valves that will be open at this pressure is...

(ignore the effects of blowdown/accumulation)

A. 2 B. 3 C. 4 D. 5 Answer: B Answer Explanation:

SG safety valves open at 1185, 1197, 1210, 1222, 1234 psig respectively.

Correct - with current pressure at 1217 psig there should only be 3 safety valves open Incorrect - 2. Plausible if the student confuses setpoints or only recalls one and assumes the others.

Incorrect - 4. Plausible if the student confuses setpoints or only recalls one and assumes the others.

Incorrect - 5. Plausible if the student confuses setpoints or only recalls one and assumes the others.

Meets KA asks to monitor pressure to ensure main steam system design limits are not exceeded RO knowledge system design and setpoints Rev 0

Question 41 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 100318 User-Defined ID: 98356

Reference:

SY1503900 Topic: 41 RO main steam safety valve setpoints RO Importance Rating: 3.0 SRO Importance Rating: 3.1 K/A Number: 039 A 1.06 Comments: NEW Lesson Plan Objective: SY1503900 R4, Discuss the operation of the Steam Generator Atmospheric Relief Valves.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 7 KA - Main and reheat steam system - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the main and reheat steam system controls including - Main steam pressure Modification History:

Rev 0

42 ID: 98357 Points: 1.00 What design features protect the containment spray nozzles from plugging during recirculation phase?

Containment Sump Strainers Debris barriers at bio shield wall Self cleaning strainer between the RWST discharge and spray pump suction A. 1, 2, & 3 B. 1&3 C. 1&2 D. 2&3 Answer: C Answer Explanation:

Correct - strainers are part of the filtering process but the installed barriers from the loops also stop this debris.

Incorrect - 1 & 3. First part is correct, second part is a device that is used on systems at Wolf Creek to filter the water and then self clean. Plausible if the student confuses where the installed self cleaning strainers are located.

Incorrect - 2 & 3. First part is correct, second part is a device that is used on systems at Wolf Creek to filter the water and then self clean. Plausible if the student confuses where the installed self cleaning strainers are located.

Incorrect - 1, 2, & 3. The first two are correct, last part is a device that is used on systems at Wolf Creek to filter the water and then self clean. Plausible if the student confuses where the installed self cleaning strainers are located.

Meets KA asks how the nozzles are protected from debris during recirc RO knowledge system design and components Rev 0

Question 42 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98357 User-Defined ID: 98357

Reference:

SY1302600 Topic: 42 RO CTMT sump strainer function RO Importance Rating: 2.8 SRO Importance Rating: 3.2 K/A Number: 026 K 4.05 Comments: NEW Lesson Plan Objective: SY1302600 R3, Describe the function of major Containment Spray System components and controls.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 8 KA - Containment Spray System - Knowledge of CSS design feature and or interlocks which provide for the following: - Prevention of material from clogging nozzles during recirculation Modification History:

Rev 0

43 ID: 98358 Points: 1.00 With the plant operating at full power which of the following failures of AE PT-508, Main Feed Header Pressure Channel, would cause an INITIAL RISE in feedwater flow to all S/G's AND what procedural action will the crew take to mitigate the failure?

A. Fails LOW. Take manual control of Main Feedwater Regulating valves.

B. Fails LOW. Take manual control of Main Feedwater pump speed.

C. Fails HIGH. Take manual control of Main Feedwater pump speed.

D. Fails HIGH. Take manual control of Main Feedwater Regulating valves.

Answer: B Answer Explanation:

Correct - this is an input to the feed pump speed control circuit so when this fails low the MFP will speed up causing an initial SG level rise. To fix this per OFN SB-008 take manual control of the MFP speed controller.

Incorrect - AE PT-508 fails high take manual control of main feed reg valves.

This failure will cause a lowering of main feed pump to restore program differential pressure for the main feed pump. The OFN will have the BOP take manual control of the pump not each feed reg valve. Plausible if the student confuses what inputs to the main feed pump speed and what affect it would have on the system as well as this action is required for other failures within the feed system.

Incorrect - AE PT-508 fails high take manual control of main feed pump speed.

Correct procedure action to take but the failure will cause a lowering of main feed pump to restore program differential pressure for the main feed pump. Plausible if the student confuses what inputs to the main feed pump speed and what affect it would have on the system.

Incorrect - AE PT-508 fails low take manual control of main feed reg valves.

Correct failure to cause the initial rise in main feed pump speed to raise flow to all SG but the OFN will have the BOP take manual control of the pump not each feed reg valve. Plausible since this is an action for different failures within the main feed pump and feed reg valves.

Meets KA because asks what is the malfunction that would cause an effect to SG from MFW RO knowledge since it is system knowledge Rev 0

Question 43 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98358 User-Defined ID: 98358

Reference:

SY1505900 Topic: 43 RO SG level effects of AE PT 508 failure RO Importance Rating: 3.5 SRO Importance Rating: 3.7 K/A Number: 059 K 3.03 Comments: MODIFIED - 58915 Lesson Plan Objective: SY1505900 R11, Discuss the instrumentation and controls of the Feedwater System, including trips and automatic actions of a Feedwater Isolation Signal (FWIS).

Tier # 2 Group # 1 Last Used - 2007 Callaway Comprehension 55.41 part 5 KA - Main feedwater - Knowledge of the effect that a loss or malfunction of the MFW will have on the following - S/G Modification History:

Rev 0

44 ID: 98359 Points: 1.00 Which ONE of the following is the correct power supply to AL HV-34, CST to MD AFP B?

A. NG01 B. NG02 C. NG03 D. NG04 Answer: D Answer Explanation:

E-13AL02B MDAFWP's don't have MOVs for discharge valves Correct - per the electrical print NG04 Incorrect - NG01. There are only four safety related power sources for this valve and this is not it. Plausible if the student doesn't know the power supply.

Incorrect - NG02. There are only four safety related power sources for this valve and this is not it. Plausible if the student doesn't know the power supply.

Incorrect - NG03. There are only four safety related power sources for this valve and this is not it. Plausible if the student doesn't know the power supply.

Meets KA because it asks for the power supply to the AFW MOVs RO knowledge system understanding Rev 0

Question 44 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98359 User-Defined ID: 98359

Reference:

E-13AL02C Topic: 44 RO Power supply to AL HV-34 RO Importance Rating: 3.2 SRO Importance Rating: 3.3 K/A Number: 061 K 2.01 Comments: MODIFIED - 59879 Lesson Plan Objective: SY1406100 R3, Explain the operation of the MDAFWP Discharge Valves.

Tier # 2 Group # 1 Last Used - 2009 Callaway Audit Memory 55.41 part 7 KA - AFW - Knowledge of bus power supplies to the following - AFW system MOV's Rev 0

45 ID: 98360 Points: 1.00 Given the following with the unit at 100% power:

Reactor trip

'B' MDAFWP fails to start Based on the information given what is the status of the following MDAFW valves 1 minute later?

X=CLOSED O=OPEN A S/G MD B S/G MD C S/G MD D S/G MD A. X O O X B. X X X X C. O X X X D. O X X O Answer: C Answer Explanation:

Validated on the desktop simulator With the reactor trip from 100% AFW will start on SG low level. The TDAFWP will start and flow to all SGs and discharge valves will not throttle they will stay full open. The B MDAFWP starting allows the signal to be sent to the smart valves to arm and throttle flow. The D SG MD valve works different. When the B MD pump trips the D SG valve will throttle closed base on flow from the TD pump but the A SG valve will stay open since the B MD pump is not running. For the smart valves to work the MD pump must be running but D SG is different.

Correct - With a start or not start of the B MD pump the D SG smart valve will still throttle closed due to flow from the TDAFW pump. But the A SG valve will not close because the B MD pump is not running which is required to make this valve work as design Incorrect - all closed. This is the normal response if no failure happens. All smart valve close based on flow. Plausible if the confuses the fact that since flow is going to all SG that all smart valves should function as normal.

Incorrect - A and D open, B and C closed. This is what would happen if the A MD pump failed to start. Plausible if the student confuses what pump feeds what SG. Correct is A - BC and B - AD Incorrect - A and D closed, B and C open. Plausible if the student understands how the smart valves work but forgets the D SG valve is wired different.

Meets KA asks for effect on AFW if pumps malfunction Rev 0

RO knowledge system understanding Question 45 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98360 User-Defined ID: 98360

Reference:

SY1406100 Topic: 45 RO AFW discharge valve positions with B pump failure RO Importance Rating: 2.6 SRO Importance Rating: 2.7 K/A Number: 061 K 6.02 Comments: NEW Lesson Plan Objective: SY1406100 R3, Explain the operation of the MDAFWP Discharge Valves.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7, 8 KA - AFW - Knowledge of the effect of a loss or malfunction of the following will have on the AFW components - Pumps Modification History:

Rev 0

46 ID: 98361 Points: 1.00 Which of the following is an acceptable power lineup per T.S. to an NN bus with the plant in MODE 4?

A. NG04 NK26 NK04 NN16 NN04 B. PG20 NK26 NK02 NN12 NN02 C. NG03 NK21 NK01 NN11 NN01 D. NG01 NK21 NK01 NN15 NN03 Answer: A Answer Explanation:

Correct - This line up is from a safeguards source through the correct path to the NN bus.

Incorrect - PG20. This power supply is a backup for the safeguards supply and is not allowed without entering the TS for loss of power to the battery charger.

Plausible as this is a correct line up to the NN bus.

Incorrect - NG03. This line up connects the two safeguards buses not allowing train separation. Plausible if the student confuses power supplies for this system.

Incorrect - NG01. This line up connects the two safeguards buses not allowing train separation. Plausible if the student confuses power supplies for this system.

Meets KA asks for AC to DC to AC physical connection for the class 1E instrument bus RO knowledge system design and above the double line TS Rev 0

Question 46 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98361 User-Defined ID: 98361

Reference:

SY1506300 Topic: 46 RO AC power to NN bus per TS RO Importance Rating: 3.5 SRO Importance Rating: 4.0 K/A Number: 062 K 1.03 Comments: NEW Lesson Plan Objective: SY1506300 R5, Explain the relationship between Technical Specifications and the Class 1E 125V DC and Class 1E 120V AC power systems at the level of detail expected for the job position.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 7 KA - AC electrical distribution - Knowledge of the physical connections and or cause effect relationships between the AC distribution system and the following systems - DC distribution Modification History:

Rev 0

47 ID: 98362 Points: 1.00 A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run of NE01 is in progress using STS KJ-011A, EDG NE01 24 HOUR RUN.

The EDG has been in parallel with offsite power for 90 minutes when the following indications are noted:

AC Megawatts 6801 kW and stable NB01 bus voltage 4.25 kV and stable Which of the following is correct with regard to this surveillance?

A. Lower bus voltage to prevent damage to the bus.

B. Lower bus voltage to prevent damage to the EDG.

C. Limit total run time at this kW level to prevent damage to the EDG.

D. Limit total run time at this kW level to prevent damage to the bus.

Answer: C Answer Explanation:

Correct - continuous loading of the EDG is limited to 6.2 Mw, any loading over that is limited by time. Loading over 6.2 to 6.8 Mw is limited to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by procedure. Operation with this high of loading can cause overheating and insulation breakdown over time.

Incorrect - limit total time to prevent damage to the bus. The bus can handle more load than the EDG so this is a correct limit for the EDG but not the bus.

Plausible if the student confuses the voltage and the MW limit.

Incorrect - lower voltage to prevent damage to the bus. This voltage is a little high but within normal limits. Plausible if the student confuses the voltage limit and the MW limit and believes that overall bus damage would occur at this voltage level.

Incorrect - lower voltage to prevent damage to the EDG. This voltage is a little high but within normal limits. Plausible if the student confuses the voltage limit and the MW limit.

Meets KA asks for ability to predict and or monitor parameters to prevent exceeding design limits with regards to overall EDG loading RO knowledge based on system precautions and limitations.

Rev 0

Question 47 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98362 User-Defined ID: 98362

Reference:

SYS KJ-011A Topic: 47 RO EDG overloaded rating for how long RO Importance Rating: 3.4 SRO Importance Rating: 3.8 K/A Number: 062 A 1.01 Comments: NEW Lesson Plan Objective: SY1406401 R3, Assess the functional interrelationship with the AC Distribution System.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 10 KA - AC electrical distribution - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the AC distribution system controls including - Significance of DG load limits Modification History:

0 - modified based on feedback from Scott 4/25/15 Rev 0

48 ID: 98363 Points: 1.00 Given the following:

The unit is at full power DC Breaker NK0111, INVERTER NN11 input from NK01, fails and opens Which of the following describes the Main Control Board annunciator response AND the operation of the NN11 transfer switch?

A. 025A, NN01 INST BUS UV, is received The transfer switch automatically transfers to the bypass source B. 025A, NN01 INST BUS UV, is received The transfer switch must be manually transferred to the bypass source C. 025B, NN11 INV TRBL/XFR, is received The transfer switch automatically transfers to the bypass source D. 025B, NN11 INV TRBL/XFR, is received The transfer switch must be manually transferred to the bypass source Answer: C Answer Explanation:

Correct - With the loss of the input DC from the battery to the NN inverter the inverter will auto swap to the bypass source (static transfer switch). This will cause only the trouble alarm 025B to come into the control room. The NN bus is still energized and working just with a different power supply.

Incorrect - 025B and manually transferred. The alarm is correct but the operation of the transfer switch is auto not manual. Plausible since this piece of equipment was replaced with a new one in spring of 2015.

Incorrect - 025A and auto transferred. The NN bus will not lose voltage and will operate normally just on the bypass source. The second part is correct.

Plausible if the student thinks this DC input loss will cause the NN bus to become de-energized.

Incorrect - 025A and manually transferred. The NN bus will not lose voltage and will operate normally just on the bypass source. The operation of the transfer switch is auto not manual. Plausible if the student thinks this DC input loss will cause the NN bus to become de-energized and since this equipment was replace in spring of 2015.

Meets KA because student must know the expected response of the DC electrical system in order to monitor that response. As the DC system feeds the NN inverters and when there is a problem with the DC input the inverter will swap to an alternate source. This indicates that there is a problem with the DC source.

RO knowledge system understanding Rev 0

Question 48 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98363 User-Defined ID: 98363

Reference:

ALR 025B REV 9 48 RO MCB alarms with loss of DC supply to NN01 and Topic:

transfer switch RO Importance Rating: 2.7 SRO Importance Rating: 3.1 K/A Number: 063 A 3.01 Comments: BANK - Ginna Lesson Plan Objective: SY1506300 R7, Integrate system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 2 Group # 1 Last Used - 2012 Ginna #47 Comprehension 55.41 part 7 KA - DC electrical distribution - Ability to monitor auto operation of the DC electrical system including - Meters annunciators dials recorders and indicating lights Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

49 ID: 98364 Points: 1.00 The unit was operating at full power when a LOCA occurred.

20 seconds AFTER the LOCA sequencer activated a CSAS signal was generated.

At what time will the Containment Spray pumps start AND why?

A. Time 60 seconds, to prevent a trip of the EDG output breaker.

B. Time 40 seconds, to prevent a trip of the EDG output breaker.

C. Time 40 seconds, to ensure all Containment fans are running prior to spray starting.

D. Time 60 seconds, to ensure all Containment fans are running prior to spray starting.

Answer: B Answer Explanation:

E-12NF01; SY 1406401 and Figure 7.

If a CSAS is not present at the 15 second step on the LOCA sequencer circuit, the Containment Spray Pumps are prevented from automatically starting until the 25 second timer is completed. This timer starts at time 15 seconds on the LOCA sequencer if the spray signal does not exist at that time. The earliest it could start in this case would be 40 seconds (15 + 25 = 40). The reason is to allow the starting current to decay down prior to starting the next required load to prevent the EDG output breaker from tripping on over current.

Correct - based on the block of the CSAS until the LOCA sequencer is at 40 seconds.

Incorrect - time 40, ensure all containment fans are running. Time is correct.

The H2 fans shift to slow speed at time 60 seconds (separate timer than the LOCA sequencer). Plausible if the student knows the time but not the reason since fans do start all the way out to the 60 second mark.

Incorrect - time 60, prevent a trip of the EDG output breaker. The H2 fans shift to slow speed at time 60 seconds (separate timer than the LOCA sequencer).

Plausible if the student confuses which loads sequenced on the safeguards bus since the H2 fans do shift at this point.

Incorrect - time 60, ensure all containment fans are running. The H2 fans shift to slow speed at time 60 seconds (separate timer than the LOCA sequencer).

Plausible if the student confuses the loads sequenced on the safeguards bus and the reason for the delay in the start for the spray pump.

Meets KA asks system design to provide for auto load sequencer actuation with respect to the EDG RO knowledge system understanding Rev 0

Question 49 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98364 User-Defined ID: 98364

Reference:

E-12NF01 Topic: 49 RO CSAS and LOCA sequencer activation RO Importance Rating: 3.5 SRO Importance Rating: 4.0 K/A Number: 064 K 4.11 Comments: NEW Lesson Plan Objective: SY1406401 R5, Discuss the Sequencer System purpose.

Tier # 2 Group # 1 Last Used - N/A Memory 55.41 part 7, 8 KA - EDG - Knowledge of EDG system design features and or interlocks which provide for the following - Auto load sequencer, safeguards Modification History:

Rev 0

50 ID: 98365 Points: 1.00 At 10:00 Wolf Creek experienced a loss of off-site power.

At 10:30 Off-site power has been restored and is ready to be re-aligned to the safety related busses.

The RO is performing SYS NB-201, TRANSFERRING NB01 POWER SOURCES, to restore normal off-site power to NB01.

The synchroscope is placed in the Main Feeder Position and is rotating slowly in the SLOW direction (counter clockwise).

Incoming voltage is 4185 VAC NB01 bus voltage is 4130 VAC What actions must the RO take to parallel the EDG with the off-site source?

A. Raise EDG speed, Raise EDG voltage B. Lower EDG speed, Raise EDG voltage C. Raise EDG speed, Lower EDG voltage D. Lower EDG speed, Lower EDG voltage Answer: B Answer Explanation:

This starts in OFN NB-035, LOSS OF OFFSITE POWER RESTORATION, and then that sends you to the SYS procedure to complete the swap.

Correct - by lowering the speed of the diesel then the offsite source will be faster (sync scope is rotating with the incoming source not the diesel) and the sync permissives will be met. The voltage must be raised for the diesel so the breaker can be closed and because the procedure has the operator match voltage. This will allow all the permissives to be met and the normal feeder breaker to be closed.

Incorrect - Lower and lower. Lower speed is correct. Lowering voltage will cause the diesel to be greater than 50 volts lower than offsite power. The procedure says to match voltage with offsite. Plausible if the student confuses which source is on the bus and which source is incoming.

Incorrect - Raise and lower. If speed is raised then the sync scope will rotate fast in the slow direction since it is looking and the incoming source compared to the running source. Lowering voltage will cause the diesel to be greater than 50 volts lower than offsite power. The procedure says to match voltage with offsite.

Plausible if the student confuses which source is on the bus and which source is incoming.

Incorrect - Raise and raise. If speed is raised then the sync scope will rotate fast in the slow direction since it is looking and the incoming source compared to the running source. Raise voltage is correct.

Rev 0

Meets KA asks for ability to operate from the control room parallel operation of the diesel to the grid with load on the diesel.

RO knowledge system understanding of parallel operations for matching voltage and speed for sync Question 50 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98365 User-Defined ID: 98365

Reference:

SYS NB-201 50 RO parameters checked as the EDG is synced with Topic:

offsite power RO Importance Rating: 3.4 SRO Importance Rating: 3.4 K/A Number: 064 A 4.07 Comments: BANK - Callaway Lesson Plan Objective: SY1406401 R3, Assess the functional interrelationship with the AC Distribution System.

Tier # 2 Group # 1 Last Used - 2011 Callaway #47 Comprehension 55.41 part 7 KA - EDG - Ability to manually operate and or monitor in the control room - Transfer EDG with load to the grid Modification History:

0 - replaced based on Scotts comments 4/25/15 Rev 0

51 ID: 98366 Points: 1.00 The RO is performing a source check of a detector on RM-11R (SP056A).

What is the reason for performing this function prior to an effluent release?

A. Cause the ALERT alarm to come in to check its setpoint.

B. Calibrate the detector prior to performing a radioactive release.

C. Prove the monitor is functional to ensure monitoring of the release.

D. Check that the display colors change due to increased rad levels from the detector.

Answer: C Answer Explanation:

SYS SP-121, ALR 062C, SY1407200. All discuss the way to perform a check source and the proper outcome of the test to ensure the monitor is operable. It also discusses what is seen if the check source fails. Precaution of the SYS states that if the monitor check source is energized and the detector reaches the source limit that it is a SAT check. The source check is only about a third decade per minute so no alarms will come in due to the check. The check is to determine if the monitor can detect radiation only, not a calibration check of alarm setpoints or accuracy.

Correct - per the SYS the monitor check is to determine if the monitor is operable or not.

Incorrect - Calibrate the detector. The check source is only a response test not a calibration of the detector. Plausible as the detector does need to be calibrated prior to use but this is not the correct way to perform that.

Incorrect - Cause the alert alarm. This will only indicate that the detector will show a response not raise it to the level of an alarm. Plausible since all the detectors do have alarm setpoint that should be checked prior to a release but this check will not set them off.

Incorrect - Display color change. While this will cause the display color to change from green to half intensity cyan for the duration of the test it will go back to green when completed sat. The color change is not why the test is ran just an indication that the test is running. Plausible since this does take place but not to just see the color change its for response check.

Meets KA asks ability to operate in the control room a check source for a detector RO knowledge system understanding of the rad monitor Rev 0

Question 51 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98366 User-Defined ID: 98366

Reference:

SYS SP-121 Topic: 51 RO RM-11 check source reason RO Importance Rating: 3.1 SRO Importance Rating: 3.2 K/A Number: 073 A 4.03 Comments: NEW Lesson Plan Objective: SY1407200 R3, EXPLAIN the basic operation of the Area Radiation Monitoring System components.

Tier # 2 Group # 1 Last Used - N/A Fundamental 55.41 part 11 KA - Process Rad Monitor - Ability to manually operate and or monitor in the control room - Check source for operability demonstration Modification History:

Rev 0

52 ID: 98367 Points: 1.00 Given the following plant conditions:

The unit is at 100% power.

Annunciator 009B, SERV WTR PMP TRIP, alarms.

Investigation reveals that ALL Service Water pumps have tripped and CANNOT be started.

Which ONE of the following describes actions required by the crew?

A. Start both ESW Trains, with one train supplying Service Water. Trip the Turbine if any Turbine trip setpoint is reached.

B. Start both ESW Trains, with one train supplying Service Water. Trip the Reactor if any Turbine trip setpoint is reached.

C. Place both ESW Trains in service. Trip the Turbine if any Turbine trip setpoint is reached.

D. Place both ESW Trains in service. Trip the Reactor if any Turbine trip setpoint is reached.

Answer: D Answer Explanation:

Technical

References:

ALR 00-009B AP 15C-003 SYS EF-200 CORRECT - Based on system response the reactor is tripped to ensure fuel integrity.

Incorrect - Place both ESW Trains in service. Trip the Turbine if any trip setpoint is reached. INCORRECT, plausible since a Turbine trip setpoint may be challenged, however to ensure adequate core cooling the reactor is tripped first and the P-4 signal will trip the turbine. Tripping the turbine will not necessarily trip the reactor if a load runback occurred due to stator water temperature increasing.

Incorrect - Start both ESW Trains, with one train supplying Service Water. Trip the Reactor if any Turbine trip setpoint is reached. INCORRECT, plausible since the loss of heat sink to the secondary components is the concern.

Incorrect - Start both ESW Trains, with one train supplying Service Water. Trip the Turbine if any trip setpoint is reached. INCORRECT, plausible since the loss of heat sink to the secondary components is the concern.

Meets KA asks for ability to predict impact of loss of service water has on plant and use procedures to correct RO knowledge overall understanding of system interrelations Rev 0

Question 52 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98367 User-Defined ID: 98367

Reference:

ALR 009B Topic: 52 RO loss of service water with the reactor at 100%

RO Importance Rating: 3.5 SRO Importance Rating: 3.7 K/A Number: 076 A 2.01 Comments: BANK - 46137 Lesson Plan Objective: LO1733203 R14, DISCUSS procedure implementation IAW AP 15C-003.

Tier # 2 Group # 1 Last Used - 2012 Palo Verde #52, 2009 Callaway #51 Comprehension 55.41 part 10 KA - Service water - Ability to predict the impacts of the following malfunction or operations on the SWS and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Loss of SWS Modification History:

0-Rev 0

53 ID: 98368 Points: 1.00 Given the following:

The plant is in MODE 4 RHR Train "A" is in service RHR Heat Exchanger Bypass Valve EJ FCV-618 is set to maintain 3400 GPM RHR Heat Exchanger Outlet Valve EJ HCV-606 demand position set at 30%

The Instrument Air supply line to RHR Heat Exchanger Bypass Valve EJ FCV-618 becomes severed and is completely detached No other air operated valves are impacted by the failure Which ONE of the following describes the plant parameter changes from the initial steady state conditions?

RCS Temperature Total RHR flow A. Higher Higher B. Lower Higher C. Lower Lower D. Higher Lower Answer: C Answer Explanation:

Correct - FCV-618 fails closed, so there is less bypass flow mixing with more HX flow, resulting in a lower temperature on the HX outlet.

Incorrect - higher/higher. Total RHR flow is controlled by FCV-618 and would lower, forcing more water through the RHR HX for cooling.

Incorrect - higher/lower. FCV 618 failing closed will result in full cooling through the RHR HX and the HX outlet temperature will lower along with the total RHR flow lowering. Plausible because the applicant may confuse valves for total flow versus HX flow.

Incorrect - lower/higher. Temperature effect is correct and plausible because the applicant may confuse valves for total flow versus HX flow Meets KA asks system response to a loss of air RO knowledge system design Rev 0

Question 53 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98368 User-Defined ID: 98368

Reference:

M-12EJ01 Topic: 53 RO EJ RHR loss of EJ FCV-618 RO Importance Rating: 3.4 SRO Importance Rating: 3.6 K/A Number: 078 K 3.02 Comments: BANK - 58852 Lesson Plan Objective: SY1300500 R7, INTEGRATE system and plant response to transient and equipment failures, including interactions with related systems.

Tier # 2 Group # 1 Last Used - 2007 Callaway Comprehension 55.41 part 7 KA - Instrument air - Knowledge of the effect that a loss or malfunction of the air system will have on the following -

Systems having pneumatic valves and controls Modification History:

Rev 0

54 ID: 98369 Points: 1.00 The following plant conditions exist:

Compressor Sequencer Selector Switch, KA HSS-310, is selected to the C-A-B position All three air compressors are selected to AUTOMATIC

'A' Air Compressor (CKA01A) is running unloaded

'B' Air Compressor (CKA01B) is NOT running

'C' Air Compressor (CKA01C) is running loaded KA PV-11, Service Air Isolation Valve, is open What is the automatic system response to an air leak that results in air system pressure decreasing to 105 psig?

A. All three air compressors are running, but only two are loaded. KA PV-11 remains OPEN.

B. Only CKA01A and CKA01C are running and both are loaded. KA PV-11 remains OPEN.

C. All three air compressors are running and all three are loaded, KA PV-11 CLOSES.

D. Only CKA01A and CKA01C are running and both are loaded, KA PV-11 CLOSES.

Answer: C Answer Explanation:

Lead cycles between 116 and 125 psig Lag cycles between 114 and 123 psig Standby cycles between 112 and 121 psig KA PV-11 closes at 110 psig.

With pressure at 105 psig, all three compressors are running loaded and KA PV-11 valve is closed.

Correct - at 105 psig all the compressors should be loaded and KA PV-11 will have closed at 110 psig.

Incorrect - only A and C running with KA PV-11 open. With pressure this low all are running and 11 is closed. Plausible if the student confuses values of start stop and or sequencer switch position.

Incorrect - all three running two loaded with KA PV-11 open. With pressure this low all are running and 11 is closed. Plausible if the student confuses values of start stop and or sequencer switch position.

Incorrect - A and C running with KA PV-11 open. With pressure this low all are running and 11 is closed. Plausible if the student confuses values of start stop and or sequencer switch position.

Meets KA asks auto operation of the instrument air system based on pressure Rev 0

RO knowledge system level of understanding of system Question 54 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98369 User-Defined ID: 98369

Reference:

SY1407800 Topic: 54 RO KA Air Compressor loading RO Importance Rating: 3.1 SRO Importance Rating: 3.2 K/A Number: 078 A 3.01 Comments: BANK -18505 Lesson Plan Objective: SY1407800 R5, Explain the alarms, controls, indications, and interlocks associated with the system.

Tier # 2 Group # 1 Last Used - 2015 ILO systems Fundamental 55.41 part 7 KA - Instrument Air - Ability to monitor automatic operation of the IAS including: Air pressure Modification History:

Rev 0

Q55, Rev 1 ID: 98370 Points: 1.00 Given the following:

• Core off load is in progress

• Equipment Hatch is open

• ALL MSIVs are closed

• ALL S/G secondary manways are removed

• Both doors in the Personnel Air Lock are OPEN but OPERABLE

• ONE door is closed in the Emergency Air Lock

• Control Room was just now informed that last night maintenance removed ONE 'B' S/G safety valve and shipped it offsite for calibration

• NO compensatory actions are in place Which of the following actions (if any) will be required at this time?

A. NO action is required at this time B. IMMEDIATELY suspend fuel movement in Fuel Building C. IMMEDIATELY suspend fuel movement in Containment D. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> install a blank flange over the removed S/G safety opening Answer: C Answer Explanation:

T.S. 3.9.4 Correct - This is an immediate action in TS 3.9.4 with fuel movement in containment.

Incorrect - No action required. Plausible if the student misunderstands containment integrity with respect to the SG manways and the safety being removed.

Incorrect - Immediately stop fuel movement in fuel building. TS only discusses in containment. Plausible if the student misunderstands the relationship of the fuel building and containment with regards to integrity.

Incorrect - Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> install blank flange. This will restore containment integrity but the TS action with fuel movement in progress is immediately not 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible if the student forgets the TS requirement.

Meets KA by asking what to do if CTMT integrity is lost during fuel movement RO knowledge this is TS above the line and less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q55, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98370 User-Defined ID: 98370

Reference:

T.S. 3.9.4 Topic: 55 RO containment closure during fuel movement RO Importance Rating: 3.7 SRO Importance Rating: 4.1 K/A Number: 103 K 3.03 Comments: NEW Lesson Plan Objective: LO1732701 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.41 part 9, 10 KA - Containment - Knowledge of the effect that a loss or malfunction of the containment system will have on the following - Loss of containment integrity under refueling operations Modification History:

0 - modified based on Scotts feedback 4/25/15 1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

56 ID: 98371 Points: 1.00 Given the following with the unit at 100% power:

OFN BB-007, RCS LEAKAGE HIGH, was entered based on PZR level lowering

'B' CCP is running BG FK-121, CCP DISCH FLOW CTRL, is in manual with 100% output demand ALL letdown orifice isolation valves are CLOSED PZR level is 48% and down at 1% per minute RCS pressure is stable at 2231 psig As the crew performs the appropriate actions which of the following describes the injection path for the CCPs FIVE minutes later?

A. Both CCPs are injecting into ALL four cold legs in ECCS mode B. 'A' CCP is injecting into loop 1 & 2 cold legs in ECCS mode ONLY

'B' CCP is injecting into loop 3 & 4 cold legs in ECCS mode ONLY C. 'A' CCP is secured

'B' CCP is injecting into the loop 1 cold leg in charging mode D. Both CCPs are injecting into the loop 1 cold leg in charging mode Answer: A Answer Explanation:

Correct - OFN BB-007 foldout page states that if charging is maximized from one pump with letdown isolated and PZR level lowering then trip the Rx and actuate SI (ECCS mode injection). Five minutes later that crew will have performed this and still be in EMG E-0 working through it. The CCPs will be both running, due to the SI, and injecting through the BIT header which is both pumps discharge coming together and flowing to all four cold legs in the ECCS mode.

Incorrect - A CCP secured and B injecting. OFN BB-007 will have the crew actuate SI and this will realign the pumps to ECCS injection mode. Plausible if the student forgets the OFN foldout page criteria for SI and thinks that the crew will continue to work through the OFN since PZR level is lowering very slow.

Incorrect - Both pumps injecting in charging mode. Both pumps will be running due to the SI but will not be using the normal charging flowpath. Plausible if the student remembers the SI off of foldout page but confuses where the injection for the CCPs is for the ECCS mode vs normal charging mode.

Incorrect - A CCP injecting into 1 and 2 B CCP injecting into 3 and 4. The SI realign both pumps to be running but not to the loops given. They will both be injecting to all loops. Plausible if the student confuses the RCS tap points for the CCPs between the normal charging and ECCS mode.

Meets KA asks for physical connection from the RCS to ECCS pumps (CCPs not used in charging mode)

RO knowledge system connection understanding Rev 0

Question 56 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98371 User-Defined ID: 98371

Reference:

M-12BB01 REV 32 56 RO RCS leak from OFN BB-007 to ECCS injection path Topic:

for CCPs RO Importance Rating: 4.5 SRO Importance Rating: 4.6 K/A Number: 002 K 1.08 Comments: NEW Lesson Plan Objective: SY1300600 R7, Determine the flow path(s), including major valve positions and pump alignments, during each phase of ECCS operation.

Tier # 2 Group # 2 Last Used - N/A Memory 55.41 part 7 KA - RCS - Knowledge of the physical connections and or cause effect relationships between the RCS and the following system - ECCS Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

57 ID: 98372 Points: 1.00 The plant is operating at 100% when the upper selected PZR level channel fails to 0%.

Which of the following is an expected plant response to this failure AND why?

(assume NO operator action)

A. Actual PZR level lowers due to charging flow lowering.

B. PZR spray valves will open due to actual PZR pressure rising.

C. ALR 033E, PZR HTR GROUP LOCKOUT, will alarm due to heaters being tripped off.

D. PZR backup heaters will energize due to actual PZR level being greater than 5% above program.

Answer: B Answer Explanation:

M-744-0028 Correct - With this failure letdown isolates and charging goes to max so no water out and more water in will raise pressure in the PZR. Heaters are off due to letdown isolation at 17% (failed channel low) so this is not adding to the pressure rise. Sprays open to lower pressure but will not lower pressure back to 2235 psig.

Incorrect - ALR 033E. This alarm will come in if any PZR heater breaker is tripped open. Since the failed channel failed low the breakers are not tripped but locked out from coming on. Plausible if student confuses how the PZR heaters operate under these conditions.

Incorrect - PZR backup heaters energize. These heaters are controlled by the upper selected channel so when it fails low they backup heaters are lock out from coming on. Plausible since actual level will rise and be greater than 5% above program which should have turned them on baring the failure.

Incorrect - PZR level lowers. This would happen if the level channel had failed HI. Plausible if the student confuses the different failure modes, when letdown isolates then charging should lower to keep from over filling the PZR.

Meets KA by asking how PZR level failure affects PZR pressure control system RO knowledge system understanding Rev 0

Question 57 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98372 User-Defined ID: 98372

Reference:

SY1301000 Topic: 57 RO PZR level channel failure effect on pressure RO Importance Rating: 3.2 SRO Importance Rating: 3.7 K/A Number: 011 K 3.03 Comments: NEW Lesson Plan Objective: SY1301000 R10, Predict the impact of a given instrument failure, heater failure, or spray valve failure on the pressurizer pressure and/or level control system.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.41 part 5, 7 KA - PZR level control - Knowledge of the effect that a loss or malfunction of the PZR level control system will have on the following - PZR PCS Modification History:

0 - modified based on Scotts feedback 4/25/15 Rev 0

Q58, Rev 1 ID: 98373 Points: 1.00 The plant is operating at 100% power when the following takes place:

• Reactor trip occurs

• 'A' train reactor trip breaker is closed

• 'B' train reactor trip breaker is open With NO operator action what affect (if any) will this failure have on the operation of the steam dump system?

A. Steam dumps will ARM and RCS temperature will go to 557°F.

B. Steam dumps will ARM and RCS temperature will go to 559°F.

C. Steam dumps will NOT ARM and RCS temperature will go to 557°F.

D. Steam dumps will NOT ARM and RCS temperature will go to 561°F.

Answer: A Answer Explanation:

Correct - The A train reactor trip breaker controls steam dump arming but arming is also controlled by loss of load from C-7 so steam dumps are armed. B train reactor trip breaker controls which controller the steam dumps work off of. With this breaker open the steam dumps will be controlled by the plant trip controller.

Incorrect - armed and 559F. Plausible if the student confuses which reactor trip breaker does what. This statement is if they reversed the correct breaker inputs.

Incorrect - not armed and 557F. Plausible since the A train reactor trip breaker still being closed means no arming from it but C-7 does arm the steam dumps. Since the B train breaker opened the plant trip controller is controlling the temperature at 557F.

Incorrect - not armed and 561F. Plausible since the A train reactor trip breaker still being closed means no arming from it but C-7 does arm the steam dumps. But if the student doesn't think the steam dumps are armed then he would think the temperature is being controlled by the ARVs which are set at 561F Meets KA asks if student can monitor the auto selection of NNIS inputs with regards to steam dump controller used during a reactor trip with a failure RO knowledge basic system design and interlocks OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q58, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98373 User-Defined ID: 98373

Reference:

SY1504100 58 RO steam dump controlling temp with reactor trip breaker Topic:

closed RO Importance Rating: 2.9 SRO Importance Rating: 2.9 K/A Number: 016 A 3.01 Comments: NEW Lesson Plan Objective: SY1504100 R3, Explain the various modes of system operation.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.41 part 7 KA - Non nuclear instrumentation - Ability to monitor auto operation of the NNIS including - Auto selection of NNIS inputs to control systems.

Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q59, Rev 1 ID: 98374 Points: 1.00 The unit has experienced an earthquake, loss of Off Site power, and a LOCA. The crew has transitioned to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT. The following indications are noted:

• RCS pressure 1790 psig

• CETC temperature 1210 °F

• RVLIS 55% down slow

• 'A' CCP OOS due to planned maintenance

• 'B' CCP Breaker tripped and will NOT reset

• Total AFW flow 150,000 lbm/hr stable

• All S/G NR levels 2% thru 8%

Based on the indications given the crew will...

A. transition to EMG FR-C1, RESPONSE TO INADEQUATE CORE COOLING B. transition to EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING C. transition to EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK D. continue with EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT Answer: A Answer Explanation:

With a loss of offsite power RCP's are off. Subcooling for the 1790 psig is around 621°F and with CETC at 1150°F subcooling is not met. Also with the power loss the NCP is unavailable.

FR procedures are monitored after the transition out of E-0 is made.

Correct - Based on CETCs over 1200F the entry condition has been made for red path to C1.

Incorrect - FR-C2 orange. Plausible if the student fails to use the 1200F as a entry condition and moves on down the F-0 to the orange path for C2 based on RCPs stopped, RVLIS, and CETC.

Incorrect - FR-H1 red path. AFW flow is lower than the 270,000 lbm/hr requirement but since SG levels are above 6% NR this can be lowered. Plausible if the student mistakes the low flow with a requirement from F-0 to go here.

Incorrect - E-1. RCS pressure is used in multiple locations throughout this procedure to determine if ECCS flow should be reduced and with pressure staying high this is a concern with RHR pumps. Plausible if the student mistakes high RCS pressure with the need to secure RHR pumps and reduce ECCS flow over the red path entry.

Meets KA asks ability to monitor core exit temperature to prevent exceeding a design limit so a procedure change is needed RO knowledge this is knowing the entry conditions for FR red or orange paths OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question 59, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98374 User-Defined ID: 98374

Reference:

EMG F-0 Topic: 59 RO C1 red path entry RO Importance Rating: 3.7 SRO Importance Rating: 3.9 K/A Number: 017 A 1.01 Comments: NEW Lesson Plan Objective: LO1732341 R6, DISCUSS the entry conditions for procedure EMG FR-C2.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.41 part 5, 7, 10 KA - In core temp monitor - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the in core temp monitor system controls including - Core exit temperature Modification History:

Rev 1: ??

Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q60, Rev 1 ID: 98375 Points: 1.00 The unit is operating at 100% power at end of life when the following takes place:

• RO reports SFP level is -25" and lowering rapidly

• OFN EC-046, FUEL POOL COOLING AND CLEANUP MALFUNCTIONS, has been entered

• Aux Building operator confirms RO report using the local level indicator

• The leak location is currently UNKNOWN Which of the following areas is required to be evacuated AND what makeup source is preferred?

A. Fuel Building Area ONLY, RWST B. Fuel Building Area ONLY, RMWST (water only)

C. Fuel Building Area AND the Aux Building, RWST D. Fuel Building Area AND the Aux Building, RMWST (water only)

Answer: A Answer Explanation:

Correct - Foldout page of the OFN directs to Attachment A which evacuates the fuel building and makeup is from a borated water source first.

Incorrect - fuel building and makeup from RMWST. The first part is correct but makeup will be from a borated water source before an unborated sources is used and since the stem says nothing about an issue with the RWST it will be used. Plausible if the student forgets what type of water to use to fill the pool.

Incorrect - fuel and aux building makeup from RWST. Makeup source is correct but the procedure only has the fuel building evacuated. Foldout page items are required knowledge for RO's. Plausible if the student thinks the two buildings have the same ventilation system.

Incorrect - fuel and aux building makeup from RMWST. Procedure only has the fuel building evacuated. Makeup will be from a borated source first. Plausible if the student confuses the ventilation systems or where makeup will come from first.

Meets KA asks ability to use procedures to correct low SFP level issues.

RO knowledge this is foldout page item and the makeup is system understanding of which source is the correct one for the SFP.

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98375 User-Defined ID: 98375

Reference:

OFN EC-046 Topic: 60 RO loss of SFP level actions per OFN RO Importance Rating: 3.1 SRO Importance Rating: 3.5 K/A Number: 033 A 2.03 Comments: NEW Lesson Plan Objective: LO1732454 R3, Given a procedure flow path, EXAMINE the available options for procedure actions.

Tier # 2 Group # 2 Last Used - N/A Fundamental 55.41 part 10 KA - SFP cooling - Ability to predict the impacts of the following malfunctions or operations on the SFP cooling system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Abnormal spent fuel pool water level or loss of water level Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

RO 75 OPS INITIAL NRC Page: 2 of 2 14 October 2015

61 ID: 98376 Points: 1.00 Spent Fuel assemblies are being moved in preparation for an upcoming refueling.

Fuel Building ventilation is in normal lineup.

Fuel Building Rad Monitor GG RE-28 loses power.

Which of the following automatic actuations occur, if any?

A. Fuel Building Supply Fan stops, both trains of Emergency Exhaust start.

B. Fuel Building Supply Fan stops, ONLY 'B' train of Emergency Exhaust starts.

C. Fuel Building Supply Fan remains running, both trains of Emergency Exhaust start.

D. Fuel Building Supply Fan remains running, ONLY 'B' train of Emergency Exhaust starts.

Answer: A Answer Explanation:

The loss or malfunction of a rad monitor will have NO effect on fuel handling system only a procedure action if this happens.

Correct - FBIS isolates the fuel building and sends a signal to isolate the control room as well.

Incorrect - fuel building supply fan stops, one train emergency exhaust starts.

Plausible if the student misunderstands that the rad monitors in the fuel building cross trip to both trains.

Incorrect - fuel building supply fan remains running, both trains of emergency exhaust starts. Plausible if the student remembers that the emergency system will start but forgets that the normal system stops.

Incorrect - fuel building supply fan remains running, one train emergency exhaust starts. Plausible if the student misunderstands that the rad monitors in the fuel building cross trip to both trains and that the normal system will stop when this occurs Meets KA by showing how a loss of rad monitors in the fuel building will affect the ventilation system which is part of the fuel handling system isolation.

RO knowledge system knowledge of ESFAS actuation signals Rev 0

Question 61 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98376 User-Defined ID: 98376

Reference:

SY1408803 Topic: 61 RO what gives a FBIS RO Importance Rating: 2.6 SRO Importance Rating: 3.3 K/A Number: 034 K 6.02 Comments: BANK - 47936 Lesson Plan Objective: SY1408803 R9, DISCUSS the function of major Fuel Building HVAC System components and controls.

Tier # 2 Group # 2 Last Used - Commanche Peak 2007 Comprehension 55.41 part 11 KA - Fuel handling equipment - Knowledge of the effect of a loss or malfunction of the following will have on the fuel handling system - Radiation monitoring systems Modification History:

Rev 0

62 ID: 98377 Points: 1.00 What is ONE purpose of the S/G Flow Restrictor?

A. Restricts flow of steam to the Main Turbine.

B. Provides a measuring point for steam flow rate.

C. Provides for the upper tap of the S/G WR level detector.

D. Provides a tap for the Main Steam Header pressure detector.

Answer: B Answer Explanation:

Correct - One of the purposes of the flow restrictor.

Incorrect - provides tap for steam pressure. Plausible since this does supply a point for the steam flow which uses pressure to density compensation.

Incorrect - restricts flow to main turbine. While it does provide a flow restriction on a steam line break (SG pressure to atmosphere) it doesn't restrict flow to the main turbine. The flow rate to the main turbine is lower than the flow would be on a steam line break. Plausible if the student confuses what flow is being restricted.

Incorrect - provides tap for SG WR level. Since there is a tap for steam flow here it is plausible that the upper tap for SG level could be at this same location.

Meets the KA asks purpose of SG component RO knowledge system understanding Rev 0

Question 62 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98377 User-Defined ID: 98377

Reference:

SY1503900 Topic: 62 RO purpose of the SG flow restrictor RO Importance Rating: 3.9 SRO Importance Rating: 4.0 K/A Number: 035 2.1.27 Comments: NEW Lesson Plan Objective: SY1503900 R1, Organize into a flow path the major components of the main and reheat steam system.

Tier # 2 Group # 2 Last Used - N/A Fundamental 55.41 part 8 KA - S/G - Conduct of ops - Knowledge of system purpose and or function Rev 0

63 ID: 98378 Points: 1.00 During a plant heat up RCS temperature is to be maintained automatically at 500°F using Condenser Steam Dumps.

What Steam Pressure Controller (AB PK-507) setting is required?

A. 3.33 B. 4.44 C. 4.54 D. 4.64 Answer: B Answer Explanation:

This controller is a 10 turn pot with 0-1500 psig Steam Tables:

500°F = 680.86 psia 680.86 psia - 14.7 psia = 666.16 psig Correct - 666.16 psig/150psig/turn = 4.44 turns Incorrect - 4.54, plausible since 680.86/150 Incorrect - 4.64, plausible since 680.86 + 14.7/150 Incorrect - 3.33, plausible since 500/150 Meets KA asks for ability to operate the steams dumps in steam pressure mode RO knowledge system understanding of the steam dumps in steam pressure mode Rev 0

Question 63 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98378 User-Defined ID: 98378

Reference:

STEAM TABLES 63 RO steam dump setpoint for maintaining RCS temp in Topic:

pressure mode RO Importance Rating: 2.7 SRO Importance Rating: 2.7 K/A Number: 041 A 4.04 Comments: BANK - 17644 Lesson Plan Objective: SY1504100 R3, Explain the various modes of system operation.

Tier # 2 Group # 2 Last Used - 2015 Systems exam #15 Comprehension 55.41 part 5, 7 KA - Steam dump - Ability to manually operate and or monitor in the control room - Pressure mode Rev 0

64 ID: 98379 Points: 1.00 The operators are raising power from 35% to full power. Current RCS boron concentration is 630 ppm.

Which of the following will be used to control reactivity during the power increase?

A. Use control rods to maintain I on the 100% power target.

B. Allow rods to move automatically to follow turbine load without diluting.

C. Use rods and normal dilution to maintain I and Tavg / Tref within band.

D. Maintain the control rods at the 'parked position' and use normal dilution to change power.

Answer: C Answer Explanation:

Correct - since a combination of rods and dilution will raise power, delta flux has to be within limits above 50% power per TS 3.2.3 and QPTR is valid at or above 50% TS 3.2.4.

Incorrect - Use control rods to maintain delta-I on 100% power target. Axial Offset (AO) is maintained as close to 100% value as possible, however, delta-I can be far from the 100% target since rods start out at a very low position causing delta-I to be much more negative than AO. Plausible if the student misunderstands the relationship between rods and boron as power changes.

Incorrect - Maintain the control rods at the 'parked position' and use normal makeup to change power. If power were higher to start with this could be true but with power crossing the 50% value rods must be moved to maintain delta-I.

Plausible if the student doesn't recognize the power change encompasses TS items.

Incorrect - Allow rods to move automatically to follow turbine load without diluting.

If power were to be lowered this would be correct rods are used in auto to help follow the turbine. Also rods will reach their full out position prior to the unit reaching full power so dilution will be required. Plausible if the student mistakes raising and lowering power rod control or misunderstands that rods will be full out prior to full power.

Meets KA asks knowledge of operations impacts of rods and boration/dilution for turbine load changes RO knowledge system interrelations of rod control and boron (CVCS) for turbine load changes Rev 0

Question 64 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98379 User-Defined ID: 98379

Reference:

GEN 004 64 RO raise power from 35% to 100% rods and dilution Topic:

relation RO Importance Rating: 2.7 SRO Importance Rating: 2.8 K/A Number: 045 K 5.23 Comments: BANK - 19533 Lesson Plan Objective: LO1732104 R5, EXPLAIN the major steps of procedure GEN 00-004.

Tier # 2 Group # 2 Last Used - Normal ops #2, WC 2001 Memory 55.41 part 1 KA - Main turbine generator - Knowledge of the operational implications of the following concepts as they apply to the main turbine generator - Relationship between rod control and RCS boron concentration during TG load increases Rev 0

65 ID: 98380 Points: 1.00 What type of fire detection device detects the presence of abnormal heat AND where are they most widely used at Wolf Creek?

A. Infrared Flame detector, in the EDG rooms.

B. Ionization detectors, in the Turbine Building.

C. Protectowire Linear Heat Detectors, in Containment.

D. Photoelectric Smoke detector, in areas of higher radiation levels.

Answer: C Answer Explanation:

Correct - Protectowire. Detects temperature in a given location based on melting of the plastic coating separating the wires. This is used only in containment.

Incorrect - Photoelectric. Detects smoke (something to block light). Plausible as this is a device used in plant to detect fire but it doesn't look for hi temperature Incorrect - Infrared flame. Detects light emitted by a flame. Plausible as this is a device used in plant to detect fire but it doesn't look for hi temperature Incorrect - Ionization. This detector that detects combustion products in the air (particles from material that have burned). This detects fire before smoke and flame is present. Plausible as this is a device used in the plant to detect fire but it doesn't look for hi temperature.

Meets KA asks about detection devices RO knowledge system design understanding Rev 0

Question 65 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98380 User-Defined ID: 98380

Reference:

SY1408600 Topic: 65 RO Protecowire detector operation and where used RO Importance Rating: 3.1 SRO Importance Rating: 3.7 K/A Number: 086 K 4.03 Comments: NEW Lesson Plan Objective: SY1408600 R3, Explain the characteristics of the system major components.

Tier # 2 Group # 2 Last Used - N/A Fundamental 55.41 part 7 KA - Fire protection - Knowledge of design features and or interlocks which provide for the following - Detection and location of fires Rev 0

66 ID: 98381 Points: 1.00 The plant has experienced an event and entered the EMG procedure network. The CRS has assigned you as the RO to monitor two Continuous Action steps from EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT. The crew now transitions to EMG FR-C1, RESPONSE TO INADEQUATE CORE COOLING The Continuous Actions from EMG E-1...

A. become optional upon entry into EMG FR-C1.

B. are NOT applicable upon entering EMG FR-C1.

C. remain applicable throughout the performance of EMG FR-C1.

D. remain applicable until superseded by directed actions of EMG FR-C1.

Answer: B Answer Explanation:

AP 15C-003 6.6.6 and 7 After transitioning to another procedure, continuous action steps are applicable unless superseded by alternate guidance in the new procedure or stated to be inapplicable.

If a Red or Orange path ERG is entered, any continuous action steps from the previous procedures should not be performed. Entry into a Red or Orange path ERG indicates that plant conditions have severely degraded and the strategy of the suspended procedures is not effective.

Correct - since a C-1 is a red path procedure any continuous actions prior to entry are no longer performed.

Incorrect - remain applicable until superseded by directed actions of EMG FR-C1. Plausible if the student misunderstands the requirement for continuous action steps in FRs.

Incorrect - are NOT applicable upon transitioning from EMG E-1. Plausible if the student confuses the transition to FRs and other EMGs since normally they do apply.

Incorrect - become optional upon entry into EMG FR-C1. Plausible since the continuous actions are applicable if transitions are made to other EMG procedures that are not red or orange path until superseded.

Meets KA asks for a conduct of ops knowledge item which procedure use is one of RO knowledge procedure rules of usage is an admin requirement for RO's to understand and be able to adhere to without the procedure in hand Rev 0

Question 66 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98381 User-Defined ID: 98381

Reference:

AP 21C-003 Topic: 66 RO admin AP continuous action steps RO Importance Rating: 3.8 SRO Importance Rating: 4.2 K/A Number: 2.1.1 Comments: BANK - 17424 Lesson Plan Objective: LO1732312 R1, DISCUSS the EMG rules of usage in accordance with procedure AP 15C-003.

Tier # 3 Group #

Last Used - WC 2001 Memory 55.41 part 10 KA - Conduct of ops - Knowledge of conduct of operations requirements Rev 0

67 ID: 98382 Points: 1.00 Which of the following conditions will have the crew:

Trip the reactor Enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION A. PZR level is 91%

B. 'A' S/G NR level is 77%

C. RCS pressure is 2185 psig D. 'B' RCP Frame vibration is 8 mils Answer: D Answer Explanation:

AP 21-001, CONDUCT OF OPERATIONS, if any trip setpoint has been reached and no auto reactor trip has worked after you verify the validity of the alarm then the reactor must be placed in a safe condition.

Correct - The RCP has exceeded the rapid shutdown criteria for the pump which will require a manual reactor trip and a stop of the affected RCP.

Incorrect - PZR level. PZR level has not exceeded the auto trip setpoint.

Plausible if the student confuses the trip setpoint with action contained in OFNs for this condition.

Incorrect - RCS pressure is low. RCS pressure has not lowered enough to cause an auto reactor trip and one is not warranted at this point. Plausible if the student thinks they should try and 'beat' the reactor trip before any setpoint is reached.

Incorrect - SG MFRV. At 77% SG level a FWIS has not been generated.

Plausible if the student confuses the trip setpoint with action contained in OFNs for this condition.

Meets KA asks what should the operator do if system limits or precautions are exceeded (conduct of ops)

RO knowledge as they are required to know when an auto trip should have occurred or if they need to take action to protect the reactor.

Rev 0

Question 67 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98382 User-Defined ID: 98382

Reference:

AP 21-001 Topic: 67 RO conduct of ops when to trip if no auto trip works RO Importance Rating: 3.8 SRO Importance Rating: 4.0 K/A Number: 2.1.32 Comments: NEW Lesson Plan Objective: LO1733211, R7, DISCUSS the purpose / scope and selected knowledge requirements of procedure AP 21-001, Conduct Of Operations.

Tier # 3 Group #

Last Used - N/A Memory 55.41 part 10 KA - Conduct of ops - Ability to explain and apply system limits and precautions Rev 0

68 ID: 98383 Points: 1.00 Which of the following actions REQUIRES a peer check prior to performance?

A. Changing rod position for rod parking.

B. Tripping of the main turbine during an ATWS event.

C. Hanging a clearance order tag in a High Radiation area.

D. Placing a MFRV in manual after a S/G controlling level fails.

Answer: A Answer Explanation:

Correct - This is required since rod parking is a planned event that will change reactivity.

Incorrect - tripping the main turbine during an ATWS. This is one of the first steps in EMG E-0 which peer checks are suspended for so it will not slow down the mitigation strategy. Plausible is the student thinks of this as a reactivity change and peer checks are required for it except during EMGs Incorrect - placing the main feed reg valve in manual. During OFNs peer checks are suspended until the plant is stable. Plausible if the student confuses this with EMG guidance.

Incorrect - hanging a clearance order tag in a high rad area. Plausible since the procedure directs this as a shall except in a high rad area.

Meets KA because a peer check is required for all planned reactivity changes which is a guideline for reactivity management at WC RO knowledge conduct of ops peer checks when required Rev 0

Question 68 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98383 User-Defined ID: 98383

Reference:

AP 21-001 Topic: 68 RO peer check required for planned reactivity changes RO Importance Rating: 4.3 SRO Importance Rating: 4.6 K/A Number: 2.1.37 Comments: NEW Lesson Plan Objective:

Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 10 KA - Conduct of ops - Knowledge of procedures, guidelines, or limitations associated with reactivity management Modification History:

0 - replaced based on Scotts feedback 4/25/15 1 - replaced based on validation comments 8/19/15 Rev 0

69 ID: 98384 Points: 1.00 Given the following RCS leak rate data at 100% power:

Total RCS leak rate is 6.9 gpm Leakage into the PRT is 3.2 gpm Leakage into the RCDT is 0.2 gpm

'A' SG tube leakage is 0.08 gpm

'B' SG tube leakage is 0.03 gpm Which one of the following Technical Specification RCS leakage limits has been exceeded?

A. Unidentified Leakage B. Total S/G Leakage C. Primary to Secondary Leakage in 'A' S/G D. Identified Leakage Answer: A Answer Explanation:

TS 3.4.13 Correct - 6.9 - (3.2 + .2 + .08 + .03) = 3.39 gpm unidentified leakage. Leakage to the PRT and RCDT is identified leakage since it can only come from a given source.

Incorrect - identified. All the identified leakage adds up to 3.51 gpm so this is less than the 10 gpm allowed by TS. Plausible if the student confuses TS values.

Incorrect - total SG. There is no limit for total SG leakage just 150 gpd for each SG but the two SG are leaking 158.4 gpd total. Plausible if the student confuses TS values.

Incorrect - primary to sec. .08 gpm is 115.2 gpd less than the limit. .03 is 43.2 gpd less than the limit. Plausible if the student confuses TS values.

Meets KA asks knowledge of TS (conditions of license) for equipment control RO knowledge TS above the double line Rev 0

Question 69 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98384 User-Defined ID: 98384

Reference:

T.S. 3.4.13 Topic: 69 RO TS operational leakage RO Importance Rating: 3.6 SRO Importance Rating: 4.5 K/A Number: 2.2.38 Comments: BANK - ANO Lesson Plan Objective: LO1732701 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 3 Group #

Last Used - 2011 ANO Comprehension 55.41 part 10 KA - Equipment control - Knowledge of conditions and limitations in the facility license Rev 0

70 ID: 98385 Points: 1.00 A NPIS computer point is coming in alarm. The RO has investigated and determined that the point is NOT in alarm and the NPIS point has malfunctioned.

Which of the following actions will the crew perform for removing / tracking this malfunctioning NPIS computer point?

A. Pull the annunciator card, place an OOS sticker on the NPIS terminal B. Pull the annunciator card, delete the analog point from alarm processing C. Make an Equipment Out-Of-Service log entry, place an OOS sticker on the NPIS terminal D. Make an Equipment Out-Of-Service log entry, delete the analog point from alarm processing Answer: D Answer Explanation:

Correct - Per AP 21F-001, EQUIPMENT OUT OF SERVICE CONTROL, since this is a NPIS point only a log entry and the deletion of the alarm from processing is all that is required.

Incorrect - Make an out of service entry and place an OOS sticker on the NPIS terminal. The procedure specifically states NOT to put a sticker on alarm points.

Plausible if the student confuses the type of alarm that is being deleted.

Incorrect - Pull the card and delete the point. This is a NPIS point and cards are associated with annunciators. Plausible if the student confuses the type of alarm that is being deleted.

Incorrect - Pull the card and place an OOS sticker on the NPIS terminal. This is a NPIS point and cards are associated with annunciators. The procedure states NOT to place an OOS sticker on alarm points. Plausible if the student confuses the type of alarm that is being deleted.

Meets KA asks for knowledge of the process of tracking inop alarms per procedure RO knowledge admin for how to document an inoperable alarm Rev 0

Question 70 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98385 User-Defined ID: 98385

Reference:

AP 21F-001 Topic: 70 RO tracking inoperable alarms RO Importance Rating: 3.0 SRO Importance Rating: 3.3 K/A Number: 2.2.43 Comments: NEW Lesson Plan Objective: LO1733213 R6, DISCUSS the operator responsibilities assigned by procedure AP 21F-001.

Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 10 KA - Equipment control - Knowledge of the process used to track inoperable alarms Rev 0

71 ID: 98386 Points: 1.00 A Wolf Creek employee with a current Form NRC-4 record needs to perform work in an area with general radiation levels of 75 mR/hr. The worker has NOT received any exposure today.

The worker's exposure history is:

Lifetime: 24.5 Rem Year to date: 1400 mR Current quarter 225 mR Which one of the following is the MAXIMUM time that the worker can work in the area without exceeding any Administrative exposure limits? (Assume NO special authorization.)

A. 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> B. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C. 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> and 20 minutes D. 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> and 40 minutes Answer: B Answer Explanation:

Correct - 2000 mR - 1400 mR = 600 mR / 75 mR = 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Incorrect - 3000 mR - 1400 mR = 1600 mR / 75 mR = 21.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. AP 25B-100 allows for up to 3000 mR if some is from another site. Plausible if the student confuses limits.

Incorrect - 2000 mR - 1400 mR - 225 mR = 375 mR / 75 mR = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The AP allows for up to 4000 mR with approval. Plausible if the student confuses limits.

Incorrect - 5000 mR -1400 mR = 2600 mR / 75 mR = 34.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. The 10 CFR 20 limits are higher. Plausible if the student confuses limits.

Meets KA asks knowledge of exposure limits under normal conditions.

RO knowledge admin procedure exposure limits is every rad workers responsibility Rev 0

Question 71 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98386 User-Defined ID: 98386

Reference:

AP 25B-100 Topic: 71 RO exposure limits under normal conditions RO Importance Rating: 3.2 SRO Importance Rating: 3.7 K/A Number: 2.3.4 Comments: BANK - 22235 Lesson Plan Objective: LO1733204 R1, DISCUSS the requirements of procedure AP 25B-100, Radiation Worker Guidelines as pertaining to the responsibilities of rad.

workers, exposure limits, and contamination controls.

Tier # 3 Group #

Last Used - N/A Comprehension 55.41 part 12 KA - Rad control - Knowledge of radiation exposure limits under normal or emergency conditions Rev 0

72 ID: 98387 Points: 1.00 Which of the following evolutions will raise radiation levels in the local area?

A. Starting up the S/G Blowdown system B. Batching a new batch of boric acid C. Adding hydrogen peroxide to the RCS D. Makeup to the BL tank (TBL01, RMWST)

Answer: C Answer Explanation:

Correct - SYS BG-207 discusses that HP must be aware that this is being done so room can be high rad.

Incorrect - SG blowdown. This water is not contaminated so this will not change rad levels. Plausible if the student confuses what water is contaminated and which is not.

Incorrect - makeup to BL tank. This will only add pure water to a tank that makes up to the RCS water. Plausible if the student confuses what water is contaminated and which is not.

Incorrect - batching boric acid. This is adding water from the RMWST to the batch add tank and adding boric acid bags. There is no caution in SYS BG-206 for changing rad levels for this evolution. Plausible if the student doesn't understand how to batch acid.

Meets KA asks for knowledge of changing rad levels while performing normal activities RO knowledge system understanding Rev 0

Question 72 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98387 User-Defined ID: 98387

Reference:

SYS EJ-110A Topic: 72 RO Changing rad levels for various evolutions RO Importance Rating: 3.4 SRO Importance Rating: 3.8 K/A Number: 2.3.14 Comments: NEW Lesson Plan Objective: LO1733204 R1, DISCUSS the requirements of procedure AP 25B-100, Radiation Worker Guidelines as pertaining to the responsibilities of rad.

workers, exposure limits, and contamination controls Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 12 KA - Rad control - Knowledge of rad or contamination hazards that may arise during normal abnormal or emergency conditions or activities Modification History:

Rev 0

73 ID: 98388 Points: 1.00 An earthquake has impacted the Wolf Creek plant. Extensive damage is observed. The operating crew notes the following conditions / indications:

Offsite power Unavailable Time since trip 40 minutes RCS pressure 2345 psig CETCs 715°F PZR level 0%

ECCS Flow indicated Rod bottom lights All lit IR SUR + 0.1 DPM RVLIS NC range 42%

S/G NR levels A 7%

B 3%

C 3%

D 5%

AFW flow 272,000 lbm/hr Containment normal sump 2004' 2" Which of the following CSFSTs is the highest priority?

A. Subcriticality B. Core Cooling C. Heat Sink D. Containment Answer: B Answer Explanation:

Current conditions show:

subcriticality - orange due to + SUR on IR core cooling - red due to low RVLIS and CETC heat sink - yellow due to not all levels greater than 6%

integrity - green due to a less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cooldown but all temps greater than 270°F containment - orange due to sump level being higher than 2003' 11" inventory - yellow due to PZR level lower than 17%

Correct - This is the only red path based on EMG F-0 Incorrect - subcriticality. This path is orange. Plausible if the student misdiagnoses the event.

Incorrect - containment. This path is orange. Plausible if the student misdiagnoses the event.

Incorrect - heat sink. This path is yellow. Plausible if the student misdiagnoses the event.

Rev 0

Meets KA asks if they can use indications, determine if entry conditions are met, and then prioritize which procedure takes priority over others (knowledge of procedure organization for emergency)

RO knowledge entry into red and orange path is RO and priority of them Question 73 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98388 User-Defined ID: 98388

Reference:

EMG F-0 REV 17 Topic: 73 RO status tree determination RO Importance Rating: 3.7 SRO Importance Rating: 4.3 K/A Number: 2.4.5 Comments: NEW Lesson Plan Objective: LO1732338 R3, DISCUSS the major action steps of procedure EMG F-0.

Tier # 3 Group #

Last Used - N/A Comprehension 55.41 part 10 KA - Emergency procedures - Knowledge of the organization of the operating procedures network for normal abnormal and emergency evolutions Modification History:

Rev 0

74 ID: 98389 Points: 1.00 Given the following plant conditions:

Plant is in MODE 4 RCS pressure is 340 psig RHR pump flow begins oscillating between 2200 and 3500 gpm The crew has entered OFN EJ-015, LOSS OF RHR COOLING Which of the following describes the required action for the crew if this condition continues?

A. Open BN HV-8812A, RWST outlet to 'A' RHR pump B. Stop the affected RHR pump C. Open FCV-618, RHR HX A BYPASS CTRL valve D. Throttle open EJ-V033, 'A' train CCW Heat Exchanger outlet valve Answer: B Answer Explanation:

Correct - per OFN EJ-015 foldout page if RHR pump flow is cycling over 1000 gpm then stop the RHR pump Incorrect - open EJ-V033. This action is discussed in this procedure but not for this condition. Plausible if the student thinks lowering RHR outlet temperature will correct the cavitation.

Incorrect - open BN HV-8812A RWST suction. Plausible as the student thinks this would raise suction pressure but since RCS pressure is 340 psig this would cause water to go from the RCS to the RWST Incorrect - open FCV-618. Plausible if the student confuses how this valve works and thinks this will lower flow, but this will raise flow.

Meets KA knowledge of RHR shutdown implications with loss of RHR and mitigation strategies. Even though this is a Tier 3 KA question, the KA is specific to the conditions presented in the stem.

RO knowledge foldout page item Rev 0

Question 74 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98389 User-Defined ID: 98389

Reference:

OFN EJ-015 Topic: 74 RO actions for RHR pumps on cavitation RO Importance Rating: 3.8 SRO Importance Rating: 4.2 K/A Number: 2.4.9 Comments: NEW Lesson Plan Objective: LO1732425 R4, Given a procedure flow path, EXAMINE the available options for procedure OFN EJ-015 actions.

Tier # 3 Group #

Last Used - N/A Fundamental 55.41 part 10 KA - Emergency procedures - Knowledge of low power shutdown implications in accident, LOCA or loss of RHR, mitigation strategies Rev 0

75 ID: 98390 Points: 1.00 Wolf Creek has declared a Site Area Emergency. You have been assigned the duties of the ENS communicator.

Which of the following items is required to be reported to the NRC over the ENS phone line per EPP 06-001, CONTROL ROOM OPERATIONS?

A. Terminating the event.

B. Dispatching a repair team into Containment C. Stopping ECCS pumps D. S/G ARV lifting at setpoint Answer: A Answer Explanation:

7.4.3 Provide the following additional information to the NRC:

1. Any further degradation in the level of safety of the plant or other worsening plant conditions

2. Any change from one emergency class to another

3. Termination of an emergency class

4. The results of ensuing evaluations or assessments of plant conditions

5. The effectiveness of response or protective measures taken

6. Any information related to plant behavior that is not understood by the NRC Correct - This is stated in the EPP as an item to inform the NRC of.

Incorrect - stopping of ECCS pumps. This is per procedure and not a degradation of the plant. Plausible if the student don't understand the required items from the EPP to report.

Incorrect - SG ARV lifting. This is not a degradation in the level of safety since it is lifting at setpoint. Plausible if the student don't understand the required items from the EPP to report.

Incorrect - dispatching of a repair team. Plausible if the student don't understand the required items from the EPP to report.

Rev 0

Question 75 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98390 User-Defined ID: 98390

Reference:

EPP 06-001 Topic: 75 RO ENS communicator reporting requirements RO Importance Rating: 3.9 SRO Importance Rating: 3.8 K/A Number: 2.4.39 Comments: NEW Lesson Plan Objective: LO1733201 R4, EXPLAIN the duties and responsibilities of the Reactor Operator IAW AP 17C-007 Tier # 3 Group #

Last Used - N/A Comprehension 55.41 part 10 KA - Emergency procedures - Knowledge of RO responsibilities in emergency plan implementation Rev 0

76 ID: 98391 Points: 1.00 The unit was at 100% power when the following indications are observed:

Power Stable RCS pressure Down fast SI occurs Tave Stable PZR level Down fast Containment pressure Up slow Containment humidity Up slow S/G pressure Stable Using the attached reference, based on the given information on what tree will the classification be made on AND why?

A. EAL-4, MAIN STEAM LINE BREAK based on SI actuated ONLY.

B. EAL-3, LOSS OF REACTOR COOLANT BOUNDARY based on SI actuated ONLY.

C. EAL-4, MAIN STEAM LINE BREAK based on Containment pressure and humidity AND SI actuated.

D. EAL-3, LOSS OF REACTOR COOLANT BOUNDARY based on apparent RCS leakage AND SI actuated.

Answer: D Answer Explanation:

Correct - EAL-3, 1,2,3,5,6,7 Alert. A LOCA pressure and level will lower fast. A steam break removes more heat from the coolant and steals some from the main turbine. The extra steam flow also raises reactor power. In this case power is constant and pressure is lowering so this is a LOCA Incorrect - EAL-4, SI actuated ONLY. A steam break will cause an SI but with power and Tave stable this is not the case here. Plausible if the student misdiagnoses this event.

Incorrect - EAL-3, SI actuated ONLY. A steam break will cause an SI based on the cooldown alone. Plausible if the student misdiagnoses this event.

Incorrect - EAL-4, Containment pressure and humidity AND SI actuated. Steam break will lower RCS pressure but it also lowers Tave. Plausible if the student misdiagnoses this event.

Meets KA asks for ability to interpret differences in overcooling (steam break) and LOCA (loss of coolant only)

SRO knowledge classify an event Rev 0

Question 76 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98391 User-Defined ID: 98391

Reference:

APF 06-002-01 Topic: 76 SRO EAL classification tree usage and why RO Importance Rating: 3.7 SRO Importance Rating: 3.7 K/A Number: 011 EA 2.13 Comments: Handout provided NEW Lesson Plan Objective: LO1733215, R1, DISCUSS how to classify an event IAW EPP 06-005.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Large break LOCA - Ability to determine or interpret the following as they apply to a large break LOCA -

Difference between overcooling and LOCA indications -

Safety function Modification History:

Rev 0

77 ID: 98392 Points: 1.00 Given the following:

Unit is shutdown for a refueling outage RCS temperature 195°F down slow

'B' RHR pump is providing shutdown cooling Containment equipment hatch is open for equipment load in

'A' RHR pump breaker has been removed from service for clean and inspect Fire in the 'D' CCW pump causes significant damage to that pump AND the 'B' CCW pump Fire brigade has the fire OUT Which of the following procedures will the SRO direct based on priority?

A. OFN BB-031, SHUTDOWN LOCA B. OFN EJ-015, LOSS OF RHR COOLING C. OFN EG-004, CCW SYSTEM MALFUNCTIONS D. OFN EC-046, FUEL POOL COOLING AND CLEANUP MALFUNCTIONS Answer: B Answer Explanation:

Correct - OFN EJ-015. This procedure contains action that will protect the RHR pumps and establish containment closure. This procedure is the highest priority based on conditions given even though other entry conditions are met for other procedures they will not correct the issue.

Incorrect - OFN BB-031. Plausible if the student confuses the loss of CCW (thermal barrier cooling) with loss of seal injection which would lead to a LOCA through the seals. Also this procedure can only be entered if the plant is in mode 3 after accumulators are isolated or mode 4.

Incorrect - OFN EG-004. Plausible if the student thinks the loss of CCW is higher priority than the loss of RHR. This procedure will not correct the loss of RHR.

Incorrect - OFN EC-046. Plausible if the student understands that the loss of CCW will lead to a loss of fuel pool cooling and thinks this is higher priority than loss of RHR.

Meets KA asks procedure usage in mode 5 with a loss of RHR cooling SRO knowledge asks for specific procedure mitigation and has the SRO prioritize which procedure will need to be used to correct the larger issue not just symptoms of the issue (loss of RHR)

Rev 0

Question 77 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98392 User-Defined ID: 98392

Reference:

OFN EJ-015 Topic: 77 SRO loss CCW leads to a loss of RHR RO Importance Rating: 4.3 SRO Importance Rating: 4.4 K/A Number: 025 2.1.23 Comments: NEW Lesson Plan Objective: LO1732425 R3, Given a procedure flow path, EXAMINE the available options for procedure OFN EJ-015 actions.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Loss of RHR - Conduct of ops - Ability to perform specific system and integrated plant procedures during all modes of plant operation - Safety function 4 Modification History:

0 - replaced based on Scotts feedback 4/25/15 1- replace based on Scotts feedback 6/18/15 Rev 0

78 ID: 98393 Points: 1.00 What is the T.S. limit for RCS DOSE EQUIVALENT I-131 AND what design base accident is this limit based on?

A. 60 uCi/gm AND LOCA B. 60 uCi/gm AND SGTR C. 500 uCi/gm AND SGTR D. 500 uCi/gm AND LOCA Answer: B Answer Explanation:

Correct - This limit is based on a SGTR event and release of steam through a SG ARV.

Incorrect - less than 500 uCi/gm and SGTR. Correct event wrong limit. Plausible as this limit is in the same TS as the iodine limit but this is for XE-133.

Incorrect - less than 60 uCi/gm and LOCA. Correct limit but wrong accident.

Plausible as a LOCA will release activity into the containment and you would want this to be as small as possible.

Incorrect - less than 500 and LOCA. Plausible based on LOCA will release activity into containment and the value is that of XE-133 from same TS.

Meets KA because asks for LCO and limits with regard to SGTR event as the design bases accident SRO because asks for TS bases knowledge and limits below the double line in TS Rev 0

Question 78 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98393 User-Defined ID: 98393

Reference:

TS 3.4.16 78 SRO dose equivalent I-131 limit and what this is based Topic:

on RO Importance Rating: 4.0 SRO Importance Rating: 4.7 K/A Number: 038 2.2.22 Comments: NEW Lesson Plan Objective: LO1732708 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 1 Group # 1 Last Used - N/A Fundamental 55.43 part 2 KA - SGTR - Equipment control - Knowledge of limiting conditions for operations and safety limits - Safety function 3

Rev 0

Q79, Rev 1 ID: 98394 Points: 1.00 Given the following with the plant at 100% power:

• Fuse disconnect NK0311, Feeder to NN13 from NK03 opened

• A loss of bus NN03 occured

• The crew stabilized the plant

• NN03 is now energized from NN15 on the bypass transformer via the static switch Based on the information given:

1. What procedure will the crew perform to correct this issue?

2. What is the OPERABILITY status of NN03?

A. 1. OFN NN-021, LOSS OF VITAL 120 VAC INSTRUMENT BUS

2. INOPERABLE B. 1. OFN NK-020, LOSS OF VITAL 125 VDC BUS NK01, NK02, NK03 AND NK04

2. INOPERABLE C. 1. OFN NN-021, LOSS OF VITAL 120 VAC INSTRUMENT BUS

2. OPERABLE D. 1. OFN NK-020, LOSS OF VITAL 125 VDC BUS NK01, NK02, NK03 AND NK04

2. OPERABLE Answer: A Answer Explanation:

We have new instrument inverters. The swing inverters can be powered from an NK (DC) source or NG (AC) source and per TS any inverter suppling an NN bus must be powered from an NK source.

Correct - Since the statement in the stem for the power source to NN15 (bypass transformer) which is NG01, this is not considered an operable line up for the instrument buses in this mode Incorrect - OFN NK-020, inoperable. This procedure is plausible since it does cover issues dealing with the NK, safeguards batteries. The stem states that a loss of power from the NK source happened so the NN was lost due to the loss of the NK. The OFN NK only deals with actual battery issues not the fuse disconnects on the busses. The inoperable part is correct.

Incorrect - OFN NN-021, operable. Correct procedure. NN15 on the bypass transformer is not operable as the power supply currently is NG01 and not an NK source per TS.

Plausible if the student is not familiar with the power supplies to the swing NN inverters and thinks this lineup is OK per TS.

Incorrect - OFN NK-020, operable.This procedure is plausible since it does cover issues dealing with the NK, safeguards batteries. The stem states that a loss of power from the NK source happened so the NN was lost due to the loss of the NK. The OFN NK only deals with actual battery issues not the fuse disconnects on the busses. Also the stem states NN15 is powered from the bypass transformer. Plausible if the student is not familiar with the power supplies to the swing NN inverters and thinks this lineup is OK per TS. Since there is two power supplies to the swing inverters the student must know which one is correct to satisfy TS OPS INITIAL NRC Page: 1 of 2 14 October 2015

Meets the K/A asks for entry conditions for abnormal operating procedures for the instrument buses SRO knowledge operability calls are SRO job function Question Q79, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98394 User-Defined ID: 98394

Reference:

OFN NN-021 79 SRO operabliity of NN03 with NN15 suppling on the Topic:

bypass RO Importance Rating: 4.5 SRO Importance Rating: 4.7 K/A Number: 057 2.4.4 Comments: NEW Lesson Plan Objective: SY1506300 R5, Explain the relationship between Technical Specifications and the Class 1E 125V DC and Class 1E 120V AC power systems at the level of detail expected for the job position.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 2, 5 KA - Loss of vital AC instrument bus - Emergency procedures - Ability to recognize abnormal indications for system operating parameters that are entry level conditions for emergency and abnormal operating procedures - Safety function 6 Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 1: ??

Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q80, Rev 1 ID: 98395 Points: 1.00 The plant is operating at 62% power while maintenance repairs the 'B' MFP. The following indications are observed by the crew: (items listed in a time line format)

• Letdown flow Lowers to 0 gpm

• VCT level 35% and lowering

• Charging flow 75 gpm and lowering

• PZR level 51.3 % and rising

• PZR pressure 2285 psig and rising

• PZR sprays Closed

• Alarm 032D, PZR LEV DEV HTRS ON, is in

• Alarm 064C, RCS WR PRESS HI, is in

• Alarm 033B, PZR HI PRESS DEV is in

• Alarm 035B, PORV OPEN, is in What event is in progress AND what procedure actions will the crew take?

A. Select away from all RED train channels B. Place excess letdown in service C. Select alternate PZR level channel for control D. Place PZR master pressure controller in manual and control pressure Answer: B Answer Explanation:

If the line downstream of KA HIS-29 which is the instrument air supply line to all of containment fails then a loss of air to containment will happen. Every air operated valve in containment will go to is failure position over time. RCS pressure will rise because all letdown is isolated and PZR sprays fail closed so PZR level will be rising and when level is 5% over program the backup heaters will turn on due to a possible insurge into the PZR.

This will cause pressure to rise to the PORV setpoint and they will open to control pressure.

Correct - Loss of air to containment so place excess letdown in service. This procedure has steps to place excess letdown in service and control seal injection control PZR level.

Incorrect - PZR level channel failure select alternate level channel for control. Plausible as the student could interpret the indications given as a level channel failure which the procedure would direct the crew to select away from the failed channel.

Incorrect - PZR pressure channel failure place master controller in manual and control pressure. Plausible as the student could interpret the indications given as a pressure channel failure and the procedure would direct the crew to take manual control of the master PZR pressure controller and control pressure.

Incorrect - Loss of NN bus select away from all red train channels. Plausible as the student could interpret the indications given as a loss of the red train instrument bus and the procedure would direct the crew to select away from all red train equipment.

Meets KA because asks for loss of air and failure understanding of components SRO only because asks to identify the correct event in progress (RO) and then what steps in the procedure to mitigate the event (SRO)

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q80, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98395 User-Defined ID: 98395

Reference:

OFN KA-019 80 SRO loss of air to containment affects and procedure Topic:

actions RO Importance Rating: 2.9 SRO Importance Rating: 3.3 K/A Number: 065 AA 2.08 Comments: NEW Lesson Plan Objective: LO1732429 R2, RECOGNIZE the available situations which are addressed by OFN KA-019.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Loss of IA - Ability to determine and interpret the following as they apply to the loss of IA - Failure modes of air operated equipment - Safety function 8 Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

81 ID: 98396 Points: 1.00 The switchyard is split East and West due to grid instability which has been reported to Wolf Creek by the TSO. As the grid instability continues to get worse indicated voltage on NB01 is now 3740 and down slow.

If the above indications continue for the next 2 minutes which of the following statements is correct with regards to NE01 AND offsite circuit operability?

A. NE01 is carrying NB01 All offsite circuits are OPERABLE B. NE01 is carrying NB01 One offsite circuit is INOPERABLE C. NE01 is in STBY All offsite circuits are OPERABLE D. NE01 is in STBY One offsite circuit is INOPERABLE Answer: B Answer Explanation:

With the east and west buses split NB01 is being powered from the east bus and NB02 is being supplied by the west (startup transformer). NB01 is a 4160 V bus and 90% of that is 3744 V so anything less than that will start the degraded voltage timers to start counting down and after a total of 119 seconds will trip both the normal and alternate feeder breakers to NB01 in this case. This starts NE01 and it closes onto the bus. This doesn't make NE01 inoperable since it is performing its safety function. The offsite circuit is inoperable since the voltage supplied to the site is now too low per TS 3.8.1 and bases.

Correct - When less than 3744 V is reached the 119 second timer starts and NE01 will then start on NB01 undervoltage and take the bus. The offsite circuit is inoperable because it can't carry the NB bus. The diesel is operable because it is performing its safety function.

Incorrect - NE01 running and offsite operable. The first part is correct. The offsite circuit is not operable since it indicates low voltage by NE01 starting on degraded voltage. Plausible if the student doesn't understand the switchyard and NB01 relationship with operability.

Incorrect - NE01 in STBY offsite inoperable. The diesel is running at this point on undervoltage on NB01 due to the degraded voltage signal opening the normal and alternate feeder breakers. Plausible if the student forgets the 90% degraded voltage setpoint or time relay.

Incorrect - NE01 in STBY and offsite operable. The diesel is running at this point on undervoltage on NB01 due to the degraded voltage signal opening the normal and alternate feeder breakers. The offsite circuit is not operable since it indicates low voltage by NE01 starting on degraded voltage. Plausible if the student doesn't understand the switchyard to NB01 relationship with operability and forgets the 90% degraded voltage setpoint or time relay.

Rev 0

Meets KA because interrelates the offsite sources with the ESF switchgear SRO because asks for an operability determination for the offsite source Question 81 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98396 User-Defined ID: 98396

Reference:

OFN AF-025 81 SRO degraded grid voltage and NE01 and offsite Topic:

operability RO Importance Rating: 3.6 SRO Importance Rating: 4.0 K/A Number: 077 AA 2.07 Comments: NEW Lesson Plan Objective: SY1406401 R3, Assess the functional interrelationship with the AC Distribution System.

Tier # 1 Group # 1 Last Used - N/A Comprehension 55.43 part 2 KA - Generator voltage and electric grid disturbances -

Ability to determine and interpret the following as they apply to generator voltage and electric grid disturbances -

Operational status of engineered safety features - Safety function 6 Rev 0

82 ID: 98397 Points: 1.00 The unit is in MODE 3 heating up after a refueling outage. SR NIs are indicating:

SE NI-31B 5 X 101 CPS SE NI-32B 5 X 101 CPS

'A' bank shutdown rods are pulled in preparation for a reactor startup. SR NIs are now indicating:

SE NI-31B 6 X 101 CPS SE NI-32B 6 X 101 CPS Five minutes pass with NO other operator action. SR NIs are now indicating:

SE NI-31B 7 X 101 CPS SE NI-32B 8 X 102 CPS Which of the following statements is correct with regard to SR NIs AND what actions are required?

A. SE NI-31B is INOPERABLE Fully insert all rods and then open reactor trip breakers B. SE NI-32B is INOPERABLE Fully insert all rods and then open reactor trip breakers C. SE NI-31B is INOPERABLE Open reactor trip breakers IMMEDIATELY D. SE NI-32B is INOPERABLE Open reactor trip breakers IMMEDIATELY Answer: B Answer Explanation:

Correct - Per TS 3.3.1 table 2 SR NI are required in mode 3, 4, or 5. With one inoperable then action must be taken to restore both to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OR rods must be inserted and reactor trip breakers opened within 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br />. Channel checks that they both agree within 1 decade. Count rate should not have changed by a full decade with just a single bank pull.

Incorrect - SE NI-32B inoperable open trip breakers immediately. First part is correct. The procedure has the rods inserted and then trip breakers opened.

Plausible if the student confuses TS and the procedure for required actions.

Incorrect - SE NI-31B inoperable open trip breakers immediately. Indication of SR NIs should be close to the same given rods were withdrawn all over in an equal pattern. Plausible if the student thinks that this detector should have raised higher (after any startup rate has decayed off) and it did not. Also the TS do not require the breakers open immediately but after rods have been manually inserted.

Rev 0

Incorrect - SE NI-31B inoperable fully insert all rods. Indication of SR NIs should be close to the same given rods were withdrawn all over in an equal pattern.

Plausible if the student thinks that this detector should have raised higher (after any startup rate has decayed off) and it did not. The second part is correct.

Meets KA asks for ability to determine expected SR count rate change when rods are moved SRO knowledge of procedure actions for the loss of a single SR NI and operability of the detector Question 82 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98397 User-Defined ID: 98397

Reference:

T.S. 3.3.1 Topic: 82 SRO operability of SR NI RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: 032 AA 2.02 Comments: NEW Lesson Plan Objective: LO1732701 R1, Given a set of conditions determine Technical Specification Operability.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.43 part 2 KA - Loss of SR NI - Ability to determine and interpret the following as they apply to the loss of SR NI - Expected change in source range count rate when rods are moved -

Safety function 7 Modification History:

Rev 0

Q83, Rev 1 ID: 98398 Points: 1.00 Given the following:

• The unit was operating at 100% power

• SGTR in 'A' S/G occurs and the unit is tripped

• SI is actuated

• S/G 'C' is faulted into Containment

• AFW flow CAN NOT be established to any S/G

• Both Containment Mini Purge Exhaust isolation outlet dampers, remain open A _____(1)________ is required to be declared. If an operator is later successful in manually closing at least one Mini Purge damper Containment integrity will be verified by ____(2)______.

(Reference attached)

A. 1. Site Area Emergency

2. CTMT PURGE ISO SYS status light illuminating B. 1. Site Area Emergency

2. Header for CTMT ISO SYS PHASE A, lit solid white C. 1. General Emergency

2. Header for CTMT ISO SYS PHASE A, lit solid white D. 1. General Emergency

2. CTMT PURGE ISO SYS status light illuminating Answer: A Answer Explanation:

Correct - The flow path for the classification is 1, 2, 3, 4, 6, 7, 5, 8 which is a SAE. The evidence that the manual action was successful is the status lights on SA066-X and Y illuminating when the damper is closed.

Incorrect - SAE and Header for CTMT ISO SYS PHASE A, lit solid white. The classification is correct. Plausible if the student forgets that these status lights on the ESFAS panels will come on to show isolation was successful. Also closing the mini purge valve doesn't bring in this white header it brings in the CPIS header Incorrect - GE and Header for CTMT ISO SYS PHASE A, lit solid white. The classification is wrong but if the student doesn't follow the correct path from box 7 to 5 then 8 and just goes to 5 this will get a GE. Plausible if the student forgets that these status lights on the ESFAS panels will come on to show isolation was successful or gets on the wrong path in the trees. Also closing the mini purge valve doesn't bring in this white header it brings in the CPIS header Incorrect - GE and containment status lights on. The classification is wrong but if the student doesn't follow the correct path from box 7 to 5 then 8 and just goes to 5 this will get a GE. The second part is correct. Plausible if the student gets on the wrong path on the trees.

Meets KA because asks to classify based on containment integrity and then after it is achieve what indication will prove you have it SRO because classification job is SRO OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q83, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98398 User-Defined ID: 98398

Reference:

APF 06-002-01 83 SRO site area emergency and containment mini purge Topic:

valves open RO Importance Rating: 3.9 SRO Importance Rating: 4.4 K/A Number: 069 AA 2.02 Comments: Provided reference - EAL sheets MODIFIED - Prairie Island Lesson Plan Objective: LO1733215 R1, DISCUSS how to classify an event IAW EPP 06-005.

Tier # 1 Group # 2 Last Used - 2010 Prairie Island #88 Comprehension 55.43 part 1, 5 KA - Loss of containment integrity - Ability to determine and interpret the following as they apply to the loss of containment integrity - Verification of auto and manual means of restoring integrity - Safety function 5 Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

84 ID: 98399 Points: 1.00 The crew has entered EMG ES-01, REDIAGNOSIS.

Given the following indications prior to and after a reactor trip and SI:

Parameter Just prior to Rx trip After Rx trip and SI RCS pressure Down fast Down slow Tave Stable Down slow PZR level Down fast Up fast SI Standby/Armed Actuated Rx power Down slow Tripped Main generator power Down slow Tripped CTMT pressure Stable Up slow S/G levels ALL stable ALL up slow S/G pressures ALL stable ALL down slow NB01 Energized Locked out NB02 Energized Energized MSIVs ALL open ALL closed What procedure will the SRO transition to from EMG ES-01?

A. EMG C-0, LOSS OF ALL AC POWER.

B. EMG E-3, STEAM GENERATOR TUBE RUPTURE.

C. EMG E-2, FAULTED STEAM GENERATOR ISOLATION.

D. EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

Answer: D Answer Explanation:

The parameters given were taken from the desktop simulator for a large PZR steam space leak. The general nature of what is given is all that is needed in ES-01 to make decisions as to where to transition to so no specific values are given.

Correct - Since E-0 must have been exited prior to coming to ES-01 the first transition out is not used. E-1 is correct since PZR level is rising and no faulted SG was found. This is the last transition out of ES-01 in the RNO column Incorrect - E-3. With levels rising in all SGs this could be thought of as a rupture.

This is from the normal AFW flow that will actuate on the reactor trip and subsequent shrink of the SGs. Plausible if the student doesn't understand all the auto actuation that will take place on the trip.

Incorrect - C-0. With a loss of one emergency bus and having and OFN and an EMG for loss of power this transition could be made. Plausible if the student fails to see that only one bus is locked out and it only takes one bus to complete mitigation actions in EMG procedures.

Rev 0

Incorrect - E-2. With SG pressure slowly lowering the student could mistake the cooldown from the ECCS pumps injecting and SG pressure lowering with a fault.

Plausible if the student fails to think about the cooldown from ECCS.

Meets KA because asks for understanding of steps in rediagnosis SRO because procedure selection based on indications and involves a procedure transition in the EOPs Question 84 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98399 User-Defined ID: 98399

Reference:

EMG ES-01 84 SRO rediagnosis and transition to E-1 from PZR steam Topic:

space leak RO Importance Rating: 4.6 SRO Importance Rating: 4.6 K/A Number: E01 2.1.20 Comments: NEW Lesson Plan Objective: LO1732314 R4, EXPLAIN the bases and knowledge requirements for selected procedure steps.

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.43 part 5 KA - Rediagnosis / SI termination - Conduct of ops - Ability to interpret and execute procedure steps - Safety function 3 Rev 0

85 ID: 98400 Points: 1.00 Given a LOCA inside Containment, which of the following is the correct procedure to enter AND why?

(use the attached indications for decision basis)

A. EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE, Red path.

Based on containment pressure.

B. EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE, Orange path.

Based on only one containment spray pump running.

C. EMG FR-Z2, RESPONSE TO CONTAINMENT FLOODING, Orange path.

Based on containment sump level.

D. EMG FR-Z3, RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL, Yellow path.

Based on containment rad levels.

Answer: C Answer Explanation:

Correct - With containment adverse and the sump level high this is the correct procedure to enter. There is only ONE procedure at Wolf Creek that deals with containment flooding and it is an orange path FR-Z2.

This procedure contains 4 steps and one is return to procedure step in effect. Indications given for this question make the student determine the status of the containment spray pumps and the fact that the CS pumps were started by the LOCA sequencer and not by the crew. This interpretation is needed to justify distractors. Then the student must determine that containment is adverse to get to the correct procedure entry.

Incorrect - FR-Z3 yellow. Conditions are met to enter this procedure on a yellow path but since the orange FR-Z2 is met procedure use and adherence has the higher level procedure enter first. Plausible since this is a correct entry but not with the orange path also met.

Incorrect - FR-Z1 red. Plausible since containment pressure is well above normal but the red path entry would be over 60 psig.

Incorrect - FR-Z1 orange. Plausible since there is one spray pump not running but this entry would be based on two spray pumps stopped or 'as least one running' per procedure.

Meets KA The student has to evaluate/determine if indications are correct for the current plant condition from the attached graphic depicting MCB switch positions, equipment status, and meter values and goes on to ask what procedure needs to be entered based on those indications SRO knowledge procedure selection based on conditions given Rev 0

Question 85 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98400 User-Defined ID: 98400

Reference:

EMG F-0 Topic: 85 SRO FR-Z entry for flooding RO Importance Rating: 4.6 SRO Importance Rating: 4.3 K/A Number: E15 2.1.31 Comments: NEW Lesson Plan Objective:

Tier # 1 Group # 2 Last Used - N/A Comprehension 55.43 part 5 KA - Containment Flooding - Ability to locate control room switches controls and indications and to determine that they correctly reflect the desired plant lineup - Safety function 5 Rev 0

Rev 0 86 ID: 98401 Points: 1.00 Given the following:

Unit is recovering from a refueling outage The RCP seal injection throttle valves were replaced ALL valves have been throttled (set)

RCS temperature is 410°F Using the attached reference and the given pressures, which of the following is an acceptable seal injection flow to each RCPs AND the basis for that flow restriction?

A. Charging header pressure 2350 psig RCS pressure 2235 psig 9.5 gpm Limit the MAXIMUM flow to the seal to prevent premature seal failure.

B. Charging header pressure 2550 psig RCS pressure 2235 psig 13.8 gpm Limit the MAXIMUM flow to the seal to prevent premature seal failure.

C. Charging header pressure 2550 psig RCS pressure 2235 psig 13.8 gpm Limit the amount of flow that is diverted from the normal injection path in accident conditions.

D. Charging header pressure 2350 psig RCS pressure 2235 psig 9.5 gpm Limit the amount of flow that is diverted from the normal injection path in accident conditions.

Answer: C Answer Explanation:

Correct - The DP here is 315 and the seal injection flow given is within the acceptable region of the provided graph. The reason is out of the basis which is to limit flow diverted away from the injection.

Incorrect - 13.8 gpm, limit seal water flow. The flow number is correct. The reason is not per the TS basis flow diverted away from the normal injection path is the concern. The spec does discuss the flow to the seals to prevent damage but this is not a limit on max flow to the seal but min flow to the seal to prevent failure. Plausible if the student can use the graph but mistakes the reason.

Incorrect - 9.5 gpm, limit amount NOT going to injection. The DP here is 115 so the max would be close to 8.5 gpm. The reason is correct. Plausible if the student mis-uses the graph and understands the basis.

Rev 0

Incorrect - 9.5 gpm, limit seal water flow. The DP here is 115 so the max would be close to 8.5 gpm. The spec does discuss the flow to the seals to prevent damage but this is not a limit on max flow to the seal but min flow to the seal to prevent failure. Plausible if the student mis-uses the graph and mistakes the reason.

Meets KA asks ECCS TS basis SRO knowledge TS basis Question 86 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98401 User-Defined ID: 98401

Reference:

BD TS 3.5.5 Topic: 86 SRO seal flow DP and reason in TS basis RO Importance Rating: 3.2 SRO Importance Rating: 4.2 K/A Number: 006 2.2.25 Comments: Handout provided NEW Lesson Plan Objective: SY1300300 R8, EXPLAIN Technical Specifications associated with the RCPs at the level of detail expected for the job position.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 2 KA - ECCS - Equipment control - Knowledge of the bases in TS for limiting condition for operations and safety limits Modification History:

0 - replaced based on Scotts feedback 4/25/15 Rev 0

Q87, Rev 1 ID: 98402 Points: 1.00 Of the following reportable events, which one has a time limit maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />?

A. Wolf Creek deviated from T.S.

B. Wolf Creek declared a Site Area Emergency.

C. Wolf Creek experienced a reactor trip from 10% power.

D. Wolf Creek sent a contaminated / injured person to the Coffey County Hospital for medical treatment.

Answer: C Answer Explanation:

Correct - Per 10 CFR 50.72 and 73 and AP 26A-001 any event that results in a reactor trip when the reactor is critical is reportable within four hours.

Incorrect - SAE. Plausible as this is reportable but within one hour not four.

Incorrect - contaminated / injured person offsite. Plausible as this is reportable within eight hours not four.

Incorrect - Deviation of TS. Plausible since this is reportable within one hour not four.

Meets KA asks by when an event is reportable to the NRC and is associated with an ESFAS signal SRO knowledge reportability is SRO function OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q87, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 98402 User-Defined ID: 98402

Reference:

10 CFR 50.72 Topic: 87 SRO reportablity within four hours RO Importance Rating: 2.7 SRO Importance Rating: 4.1 K/A Number: 013 2.4.30 Comments: NEW Lesson Plan Objective: LO1734021 R5, Given initial conditions, determine the reportability requirements to the NRC.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 1 KA - ESFAS - Emergency procedures - Knowledge of events related to system operation status that must be reported to internal organizations or external agencies such as the state the NRC or the transmission system operator Modification History:

0 - modified based on Scotts feedback 4/25/15 1 - revised based on NRC comment 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

Q88, Rev 1 ID: 98403 Points: 1.00 Given the following:

• Reactor power was 100%

• LOCA occurred

• SI auto actuated

• CTMT pressure peaked at 32 psig

• Crew entered the appropriate procedures

• Annunciator 047D, RWST LEV LOLO 1 AUTO XFR, is received

• Crew transitioned to EMG ES-12, TRANSFER TO COLD LEG RECIRCULATION

• Crew has completed EMG ES-12

• Annunciator 047C, RWST LEVEL LOLO2, is received

• Both RHR pumps are now indicating fluctuating flow and amps

1. What is the current condition of the Containment Spray pumps?

2. What procedure will the crew transition to?

A. 1. Both spray pumps are running

2. EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE B. 1. Both spray pumps are stopped

2. EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE C. 1. Both spray pumps are stopped

2. EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION D. 1. Both spray pumps are running

2. EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION Answer: A Answer Explanation:

Correct - ES-12 has no step to stop containment spray pumps with RWST level greater than 6% (foldout page). It does have the spray pumps aligned to the sump if RWST level is less than 12%. C-13 will be entered based on RHR pump flow.

Incorrect - ES-13 and stopped. ES-12 has no step to stop containment spray pumps with RWST level greater than 6%. It does have the spray pumps aligned to the sump if RWST level is less than 12%. Plausible if the student uses foldout page of ES-12 to stop pumps.

Incorrect - C-11 and running. The running is correct while in ES-12 but a transition to C-11 would only be made if the valves to RHR could not be placed in the correct alignment.

Plausible if the student sees this as a valve problem and not a loss of suction due to blockage.

Incorrect - C-11 and stopped. ES-12 has no step to stop spray pumps with RWST level greater than 6%. Plausible if the student sees this as a valve problem and not a loss of suction due to blockage.

Meets KA asks for procedure usage to correct a loss of containment spray while in recirc mode due to a loss of suction SRO knowledge since this question sets up a time line of events and after the given procedure is complete it asks the status of equipment (SRO knowledge of procedure steps) and then asks what procedure the crew will transition to next (SRO knowledge based on assessment of conditions given).

OPS INITIAL NRC Page: 1 of 2 14 October 2015

Question Q88, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 4.00 System ID: 98403 User-Defined ID: 98403

Reference:

EMG C-13 Topic: 88 SRO transition to C-13 and status of spray pumps RO Importance Rating: 3.6 SRO Importance Rating: 3.9 K/A Number: 026 A 2.07 Comments: NEW Lesson Plan Objective: LO1732332, R3, SUMMARIZE the major action categories and the bases for the steps that accomplish each category.

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 5 KA - Containment spray - Ability to predict the impacts of the following malfunction or operations on the CSS and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Loss of containment spray pump suction when in recirc mode possibly caused by clogged sump screen pump inlet high temperature exceeded cavitation voiding or sump level below cutoff interlock limit Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

89 ID: 98404 Points: 1.00 The unit is operating at 100% power when a reactor trip occurs on S/G low level. 30 seconds later auto SI occurs on containment pressure. The RO then throttles AFW flow per procedure. The following indications are now observed:

Parameter Trip plus 2 minutes Trip plus 5 minutes RCS pressure 1965 psig down fast 1915 psig down slow RCS temperature 544°F down fast 520°F down slow PZR level 23.5% down fast 15.2% down slow CTMT pressure 12 psig up slow 18 psig stable CTMT Rad Normal Normal S/G pressure A 940 psig down slow 800 psig down slow B 936 psig down slow 783 psig down slow C 938 psig down slow 617 psig down fast D 941 psig down slow 798 psig down slow S/G WR level A 42.5% down slow 35.9% down slow B 42.3% down slow 35.3% down slow C 23.4% down slow 6.9% down slow D 41.8% down slow 33.7% down slow S/G steam flow 0.5 X 106 MPPH 0 MPPH A

B 0.5 X 106 MPPH 0 MPPH C 0 MPPH 0 MPPH D 0.5 X 106 MPPH 0 MPPH As the crew works through EMG E-0, REACTOR TRIP OR SAFETY INJECTION:

1) What event is in progress?

2) What procedure will mitigate the event?

3) What actions will be taken during the performance of this procedure?

A. 1) Main Feed Line break in Containment

2) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT

3) 'C' S/G steam flow will be isolated AND ECCS flow WILL be reduced B. 1) Main Steam Line break in Containment

2) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT

3) 'C' S/G AFW flow will be isolated to AND ECCS flow WILL be reduced C. 1) Main Steam Line break in Containment

2) EMG E-2, FAULTED STEAM GENERATOR ISOLATION

3) 'C' S/G AFW flow will be isolated to AND ECCS flow WILL NOT be reduced D. 1) Main Feed Line break in Containment

2) EMG E-2, FAULTED STEAM GENERATOR ISOLATION

3) 'C' S/G steam flow will be isolated AND ECCS flow WILL NOT be reduced Answer: D Rev 0

Answer Explanation:

All data was gathered from running this on the desk top simulator with AFW flow throttled at event initiation, a full feedline break inside containment, and no other operator actions. This would be consistent with working through E-0 since all actions up to diagnostics have no effect on the parameters used here.

Correct - Per E-0 C SG is faulted at step 16 a transition is made to E-2. The diagnosis key is early steam flow from all but C SG and C SG WR level being lower than all the rest with no indicated steam flow which points at the feed line break. At first a feed line break will look like a steam break until the feed ring is uncovered and then the SG pressure will drop rapidly and then the diagnosis can be made. E-2 will isolate all the flow from and feed to the SG. At the end of E-2 the operator is asked if ECCS flow should be reduced based on indications.

Since the RCS pressure is still lowering a transition is made to E-1 and when the SG blows dry and the pressure stabilizes out then ECCS flow can be reduced.

This is the last step of E-2.

Incorrect - steam line break, E-1, AFW iso. The steam line break in containment would drop SG pressure immediately. E-1 would not be the correct procedure entry if there was a main steam line break in containment. AFW will be isolated in E-0. Plausible if the student misdiagnosis this event.

Incorrect - feed line break, E-1, SG C iso. Correct diagnosis but E-1 is not the correct transition. Also ECCS flow will not be reduced since RCS pressure is still lowering. Plausible if the student misdiagnosis the event.

Incorrect - steam line break, E-2, AFW iso. The steam line break in containment would drop SG pressure immediately. E-2 is the correct procedure. AFW will be isolated in E-0 not E-2, only the steam flow from the C SG to the TDAFWP.

Plausible if the student misdiagnosis the event.

Meets KA asks for ability to predict the impact of a main feed line break inside containment and then select the correct procedure to mitigate the event and control the consequences.

SRO knowledge since it asks for specific procedure step actions i.e. not reducing ECCS flow.

Rev 0

Question 89 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 5 Difficulty: 4.00 System ID: 98404 User-Defined ID: 98404

Reference:

EMG E-2 Topic: 89 SRO diagnose feedline break in containment RO Importance Rating: 3.1 SRO Importance Rating: 3.4 K/A Number: 059 A 2.05 Comments: MODIFIED - 16504 Lesson Plan Objective: LO1732324 R4, EXPLAIN the bases and any knowledge requirements for selected procedure steps.

Tier # 2 Group # 1 Last Used - STP 2001 Comprehension 55.43 part 5 KA - Main feedwater - Ability to predict the impacts of the following malfunctions or operations on the MFW system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Rupture in MFW suction or discharge line Modification History:

Rev 0

90 ID: 98405 Points: 1.00 Given the following:

The crew is performing STS KJ-005A, MANUAL/AUTO START, SYNC AND LOADING OF EDG NE01 The RO is ready to close the output breaker for NE01 in order to parallel it with NB01 (off site)

NB01 voltage is indicating 4160 VAC and frequency is indication 60.1 Hz NE01 voltage is indicating 4080 VAC and frequency is indicating 60.3 Hz What action is required to complete the paralleling operation of NE01 to NB01?

Under these conditions, a copy of which procedure is required to be kept with the RO and NSO to keep NE01 OPERABLE during this surveillance run?

A. 1. Lower NE01 frequency.

2. OFN NB-042, LOSS OF OFFSITE POWER TO NB01(NB02) WITH EDG PARALLELED.

B. 1. Lower NE01 frequency.

2. SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR 'A'.

C. 1. Raise NE01 voltage.

2. SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR 'A'.

D. 1. Raise NE01 voltage.

2. OFN NB-042, LOSS OF OFFSITE POWER TO NB01(NB02) WITH EDG PARALLELED.

Answer: D Answer Explanation:

Correct - Per SYS KJ-123 the incoming voltage needs to be within +/- 50 volts to parallel the diesel. STS KJ-005A 4.11.1 states that if individuals and a copy of OFN NB-042 are present during the test the EDG remains operable.

Incorrect - raise voltage copy of SYS. The SYS is only the test procedure after maintenance and the OFN is required at the location and in the control room for the EDG to remain operable. Plausible if the student mistakes the contents of the SYS with the requirement to have the OFN.

Incorrect - lower frequency copy of OFN. Plausible if the student confuses the starting parameter requirements for the EDG. The OFN is correct to have at the locations.

Incorrect - lower frequency copy of SYS. Plausible if the student confuses the starting parameter requirements for the EDG. Also the SYS is only the test procedure after maintenance and the OFN is required at the location and in the control room for the EDG to remain operable. Plausible if the student mistakes the contents of the SYS with the requirement to have the OFN.

Rev 0

Meets KA asks about using procedures to operate the EDG when sync'ing it to other sources SRO only because asks about what it takes to maintain operability of the EDG during the test run Question 90 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98405 User-Defined ID: 98405

Reference:

SYS KJ-123 90 SRO EDG operability when paralleled and initial Topic:

conditions to sync RO Importance Rating: 3.1 SRO Importance Rating: 3.3 K/A Number: 064 A 2.09 Comments: NEW Lesson Plan Objective:

Tier # 2 Group # 1 Last Used - N/A Comprehension 55.43 part 2 KA - EDG - Ability to predict the impacts of the following malfunctions or operations on the EDG system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Sync of EDG with other electric power supplies Rev 0

91 ID: 98406 Points: 1.00 Given the following:

Reactor power is 85%

80B, RPI NON URG ALARM, is received DRPI indicates Data A failure 1,2,3 AND GW for ALL rods What is the status of the DRPI system and what is the current accuracy of the system?

A. OPERABLE, +10/-4 B. OPERABLE, -10/+4 C. INOPERABLE, +10/-4 D. INOPERABLE, -10/+4 Answer: A Answer Explanation:

Correct - With the loss of only one power supply per TS the DRPI system goes to half accuracy which is still within the +/- 12 steps the TS asks for. The DRPI system remains operable.

Incorrect - OPERABLE -10/+4. Correct TS call. The accuracy is +10/-4 (opposite common mistake). Plausible if the student confuses the accuracy but knows the TS.

Incorrect - INOPERABLE +10/-4. Accuracy is correct and with a power failure of one DRPI then the panel will show alarms and lights. Plausible if the student knows the accuracy for the power failure but applies TS wrong (common misconception) for only having one power supply available.

Incorrect - INOPERABLE -10/+4. Common misconception for accuracy.

Plausible if the student applies the TS wrong with the loss of power.

Meets KA asks ability to use plant procedures (TS) to control the loss of RPIS power SRO knowledge operability determinations are SRO only job function Rev 0

Question 91 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98406 User-Defined ID: 98406

Reference:

SY1301400, TS 3.1.7 Topic: 91 SRO loss of DRPI A TS operability and accuracy RO Importance Rating: 3.1 SRO Importance Rating: 3.6 K/A Number: 014 A 2.02 Comments: NEW Lesson Plan Objective: SY1301400 R4, Describe how the Digital Rod Position Indication System control board display unit indicates a half accuracy condition and State the rod position accuracy while in this condition.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.43 part 2 KA - 014 A 2.02 - Rod position indication system - Ability to predict the impacts of the following malfunctions or operations on the RPIS and based on those on those predications use procedures to correct control or mitigate the consequences of those malfunctions or operations -

loss of power to the RPIS Rev 0

92 ID: 98407 Points: 1.00 Given the following information with the plant in MODE 1:

1200 A fire has occurred in the Control Room 1203 Spurious equipment has actuated 1205 Control Room evacuation has taken place 1205 Halon has NOT activated for the cable trenches 1225 CRS places all the Isolate/Transfer switches in isolate What procedure will mitigate this event and what will be the MINIMUM classification that will be made?

A. OFN RP-013, CONTROL ROOM NOT HABITABLE Alert B. OFN RP-017, CONTROL ROOM EVACUATION Site Area Emergency C. OFN RP-017, CONTROL ROOM EVACUATION Alert D. OFN RP-013, CONTROL ROOM NOT HABITABLE Site Area Emergency Answer: B Answer Explanation:

Correct - OFN RP-017 is the only procedure that will mitigate this event since this is a fire in the control room. Since the notes for the procedure and the emergency plant state that the isolate switches must be in iso within 15 minutes and the time line shows more than that the SAE is called.

Incorrect - OFN RP-017 and Alert. Correct procedure but since the switches were not placed correctly within 15 minutes the alert is to low. Plausible if the student fails to see the time the switches were moved and just classifies off the notes and the control room evacuation Incorrect - OFN RP-013 and SAE. This procedure will be entered if the control room is evacuated for reasons OTHER than fire so this procedure will not mitigate this event. SAE is correct. Plausible if the student knows the correct classification but misses that a fire has occurred.

Incorrect - OFN RP-013 and Alert. This procedure will be entered if the control room is evacuated for reasons OTHER than fire so this procedure will not mitigate this event. Since the switches were not placed correctly within 15 minutes the alert is too low. Plausible if the student misses the fire and the switches.

Meets KA assumes failure of auto fire protection system and then asks for a procedure to mitigate consequences of this failure.

SRO knowledge asks for classification if control room switches are not isolated within the 15 minute clock.

Rev 0

Question 92 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98407 User-Defined ID: 98407

Reference:

OFN RP-017 92 SRO correct procedure to use with control room evac Topic:

and a classification RO Importance Rating: 3.3 SRO Importance Rating: 3.9 K/A Number: 086 A 2.04 Comments: NEW Lesson Plan Objective: LO1732426 R4, EXPLAIN the basis and any knowledge requirement for selected procedure steps.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.43 part 5 KA - Fire protection - Ability to predict the impacts of the following malfunctions or operations on the fire protection system and based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations - Failure to actuate the FPS when required, resulting in fire damage Modification History:

Rev 0

93 ID: 98408 Points: 1.00 The plant is in a refueling outage with the following:

Core off load is in progress with just over half of the assemblies remaining in the vessel The Upender is moving to the vertical position on the Refuel Pool side A fuel assembly is on the hoist and has just been withdrawn from the core Refuel Pool level suddenly begins to lower rapidly The Control Room reports that the 'A' S/G Nozzle Dam has failed Which of the following actions will the Fuel Handling SRO will direct?

A. Lower the fuel assembly into the vessel Send the Fuel Transfer Cart to the SFP B. Place the fuel assembly in the Upender Send the Fuel Transfer Cart to the SFP C. Lower the fuel assembly into the vessel Leave the Fuel Transfer Cart in the Refueling Pool D. Place the fuel assembly in the Upender Leave the Fuel Transfer Cart in the Refueling Pool Answer: A Answer Explanation:

Fuel handling is an SRO only function at Wolf Creek.

Correct - Per the OFN KE-018, FUEL HANDLING ACCIDENT, there are multiple actions the fuel handling SRO will perform. This list is all correct for what is required to completed. The fuel transfer cart must be on the SFP side or the fuel transfer gate valve cannot be closed. The fuel assembly is placed back in the only safe location on the refuel pool side. The transfer tube gate is closed to prevent losing SFP level. The equipment hatch is closed to regain containment integrity for this event.

Incorrect - Place the fuel assembly in the upender, send the cart to the SFP.

Plausible if the student does not understand the safe locations to place the fuel assembly, the rest is correct.

Incorrect - Leave the cart in the refuel pool, ensure the gate is closed. Plausible if the student does not understand that the cart has a cable connected to it that goes back to the SFP side stopping the gate valve from being closed.

Incorrect - leave the cart in the refuel pool, place the fuel assembly in the upender. Plausible if the student does not understand the safe locations to place the fuel assembly or where to put the transfer cart.

Rev 0

Meets KA asks if the student understands how the refueling process goes as far as sequence of events (if the fuel element is on the hoist and just been raised it is still over the vessel, upender is just coming to the upright position must understand where it was and were it is going to know what is next). Knowing what to do in this case is based on the prediction that level will drop far enough to force the action specified so that design limits for fuel cooling and radiation shielding are not exceeded.

SRO knowledge asks for procedure step recall. The actions taken are performed by the refuel SRO inside containment (as directed by the CRS).

Question 93 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98408 User-Defined ID: 98408

Reference:

OFN KE-018 Topic: 93 SRO refuel pool inventory loss fuel handling accident RO Importance Rating: 2.9 SRO Importance Rating: 3.7 K/A Number: 034 A 1.02 Comments: NEW Lesson Plan Objective: LO1732428 R3, Given a procedural flow path, EXAMINE the available options for procedure actions.

Tier # 2 Group # 2 Last Used - N/A Comprehension 55.43 part 7 KA - Fuel handling equip - Ability to predict and or monitor changes in parameters to prevent exceeding design limits associated with operating the fuel handling system controls including - water level in the refueling canal Rev 0

94 ID: 98409 Points: 1.00 Which of the following states the EDG and support systems design mission time?

Which system design limits are being protected by the EDGs per T.S. basis?

A. 1. 7 days

2. Fuel and Containment B. 1. 7 days

2. S/G (secondary side) and SFP C. 1. 10 days

2. Fuel and Containment D. 1. 10 days

2. S/G (secondary side) and SFP Answer: A Answer Explanation:

TS BD 3.8.3 Correct - Total support system time is design 7 days per TS. The systems is protected are the fission product boundary (fuel, RCS, containment)

Incorrect - 7 days S/G and SFP. First part is correct as well as RCS but SG are not part of the systems the EDG is protecting. Yes they are half in the RCS but the secondary side is not used in the design bases accident. Plausible if the student knows the time but not what the EDG function is during a DBA.

Incorrect - 10 days Fuel and containment. The systems are correct but the time is not. TS do discuss a 10 supply but not as the design of the system. Plausible if the student knows the systems but not the design run time.

Incorrect - 10 days SG and SFP. Time is wrong and the SG are wrong. The RCS is correct. Plausible if the student is not familiar with TS bases for the EDG or the mission time.

Meets KA asks ability to explain system design limits. This question asks for the EDG limits as it applies to the fission product boundaries.

SRO knowledge TS bases for EDG support systems Rev 0

Question 94 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 98409 User-Defined ID: 98409

Reference:

BD TS 3.8.3 Topic: 94 SRO EDG mission time and systems its protecting RO Importance Rating: 3.8 SRO Importance Rating: 4.0 K/A Number: 2.1.32 Comments: NEW Lesson Plan Objective: LO1732700R7, DESCRIBE the bases of Technical Specifications and how the bases relate to the various sections.

Tier # 3 Group #

Last Used - N/A Fundamental 55.43 part 2 KA - Conduct of ops - Ability to explain and apply system limits and precautions.

Rev 0

95 ID: 98410 Points: 1.00 Given the following with the plant shutdown and in a refueling outage:

Core offload is in progress A single head stud was left in the vessel flange (stuck)

The area was programed into the PLC of the Refueling Machine A need has arose which will require operating the Refueling Machine in Bypass mode Based on the conditions given can the Bypass operation be allowed?

Why and or how?

A. 1. NO

2. Bypass operation is NOT allowed with fuel in the vessel.

B. 1. YES

2. The Manipulator Crane will still prevent movement in the affected area.

C. 1. NO

2. The Manipulator Crane will NOT prevent movement inside of the affected area.

D. 1. YES

2. The SRO and crane operator can maintain heightened awareness around the stud area.

Answer: D Answer Explanation:

Correct - per FHP 03-001 the refuel machine may be used in bypass mode as long as the SRO and the crane operator maintain heightened awareness around the area.

Incorrect - NO, the Manipulator Crane will not prevent hitting the stud. Plausible as the reason is correct the PLC will not prevent the crane from hitting the stud in bypass mode but this operation is allowed by FHP procedure.

Incorrect - NO, bypass operation is not allowed with fuel in the vessel. Plausible as bypass operation will allow the crane to move in areas it normally would not and with fuel in the vessel this is a conservative choice.

Incorrect - YES, the Manipulator Crane will still maintain the area of the stud off limits. Plausible if the student does not understand the bypass operation of the refuel machine.

Meets KA asks about knowledge of the refuel process with respect to the refueling machine SRO knowledge the refueling process with regards to the refuel machine Rev 0

Question 95 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 98410 User-Defined ID: 98410

Reference:

FHP 001 Topic: 95 SRO refuel in bypass mode with a stuck stud RO Importance Rating: 2.8 SRO Importance Rating: 3.7 K/A Number: 2.1.41 Comments: NEW Lesson Plan Objective:

Tier # 3 Group #

Last Used - N/A Fundamental 55.43 part 7 KA - Conduct of ops - Knowledge of the refueling process Rev 0

96 ID: 98411 Points: 1.00 The unit has been shut down in preparation for a refueling outage. The Operations Manager is conducting a brief for GEN 00-008, RCS LEVEL LESS THAN REACTOR VESSEL FLANGE OPERATIONS.

During the performance of this 'Infrequently Performed and Potentially Degrading Evolution' brief which of the following is required to be discussed?

A. Contact information of field participants.

B. The need for managing breaks during the test.

C. Responsibilities of management for oversight of the test.

D. The importance of completing the test as quickly as possible.

Answer: C Answer Explanation:

SRO level question based on applicant having to have the knowledge of administrative requirements concerning the conduct of Infrequently Performed Test or Evolution (IPTE) which could affect the Margin of Safety for the plant.

This brief is required to be provided by management ONLY so ROs would not perform this.

Correct - as stated in the AP Incorrect - contact info. While important this would be discussed in the pre job brief not the IPTE brief by management. Plausible if the student is unfamiliar with the IPTE briefing requirements.

Incorrect - completing as quickly as possible. This is always part of every outage time pressure. For this case it would be important to have the test move in an efficient manor to minimize time in reduced inventory but not at the expense of safety or correctness. Plausible if the student is unfamiliar with the IPTE briefing requirements.

Incorrect - managing breaks. This could be discussed in the pre job brief if the test would take a long time but not applicable for the management oversight brief. Plausible if the student is unfamiliar with the IPTE briefing requirements.

Meets KA asks knowledge of process for conducting a special test brief SRO knowledge SRO job function Rev 0

Question 96 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98411 User-Defined ID: 98411

Reference:

AI 15C-006 96 SRO requirements during an IPTE brief for Topic:

management RO Importance Rating: 2.9 SRO Importance Rating: 3.6 K/A Number: 2.2.7 Comments: MODIFIED - Callaway Lesson Plan Objective: LO1733201 R3, EXPLAIN the duties and responsibilities of the Control Room Supervisor IAW AP 17C-006.

Tier # 3 Group #

Last Used - 2013 Callaway Fundamental 55.43 part 5, 6 KA - Equipment control - Knowledge of the process for conducting special or infrequent test Rev 0

97 ID: 98412 Points: 1.00 Maintenance activities are being performed during full power operations that have resulted in additional work needing to be added to the original job Work Scope and it now requires contingency measures be put into place to perform the activity.

What procedure would be used to screen, assess, and manage the addition to the original Work Scope?

A. AP 22C-002, WORK CONTROLS B. AP 22A-001, SCREENING, PRIORITIZATION, AND PRE-APPROVAL C. AP 16C-006, MPAC WORK REQUEST / WORK ORDER PROCESS CONTROLS D. AP 22C-003, ON-LINE NUCLEAR SAFETY AND GENERATION RISK ASSESSMENT Answer: D Answer Explanation:

Correct - This procedure assesses the added risk for the scope growth.

Incorrect - AP 22C-002. This procedure will not cover the work scope change.

Plausible since this procedure deals with work in general whether online or not.

Incorrect - AP 22A-001. This procedure will not cover the increase in risk.

Plausible since this procedure discusses the screening of work which for this case would seem to be correct.

Incorrect - AP 16C-006. This procedure will not cover the increase in risk.

Plausible since this procedure discusses the generation of the new work order or instructions.

Meets KA asks which procedure will assess the correct risk due to added work scope while online.

SRO knowledge due to job function Rev 0

Question 97 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98412 User-Defined ID: 98412

Reference:

AP 22C-003 Topic: 97 SRO AP 22C-003 scope and risk RO Importance Rating: 2.6 SRO Importance Rating: 3.8 K/A Number: 2.2.17 Comments: BANK - 16910 Lesson Plan Objective: LO1734018 R17, DESCRIBE the purpose, scope and operator responsibilities of procedure AP 22C-003, ON-LINE NUCLEAR SAFETY AND GENERATION RISK ASSESSMENT.

Tier # 3 Group #

Last Used - N/A Fundamental 55.43 part 5 KA - Equipment control - Knowledge of the process for managing maintenance activities during power operations such as risk assessments work prioritization and coordination with the transmission system operator Rev 0

98 ID: 98413 Points: 1.00 Given the following plant conditions:

A LOCA has occurred A General Emergency has been declared and the TSC has been activated A two man repair team was dispatched from the TSC into the Aux Building for emergency repairs on a valve to stop the release Due to an accident, both repair team members have been injured, one with a life threatening head injury

1. What is the maximum allowed dose (TEDE) that can be authorized for a rescue team in accordance with EPP 06-013, EXPOSURE CONTROL AND PERSONNEL PROTECTION?

AND

2. Under the conditions given, who approves this dose?

1 2 A. 25 rem Shift Manager B. 10 rem Shift Manager C. 10 rem Site Emergency Manager D. 25 rem Site Emergency Manager Answer: D Answer Explanation:

Correct - This is a non delegatable duty of the site emergency manager and since the repair team was sent from the TSC the shift manager is no longer the emergency manager. The 25 rem TEDE is per procedure.

Incorrect - 25 rem shift manager. Dose is correct but since the TSC is active the shift manager cannot make this determination, only the site emergency manager.

Plausible if the student doesn't know the non delegatable duties.

Incorrect - 10 rem site emergency manager. Approval is correct but the dose is less than allowed by procedure. Plausible if the student doesn't know the dose limits for life saving.

Incorrect - 10 rem shift manager. Dose is less than allowed by procedure and approval is non delegatable to shift manager since the TSC is active. Plausible if the student doesn't know the live saving dose limits or the non delegatable approvals.

Meets KA asks what rad dose is the limit for emergencies. Knowing this limit and who can authorize it is knowledge of the rad hazards that can arise during emergencies.

SRO knowledge job function Rev 0

Question 98 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 98413 User-Defined ID: 98413

Reference:

EPP 06-013 Topic: 98 SRO life saving exposure limits and who can approve RO Importance Rating: 3.4 SRO Importance Rating: 3.8 K/A Number: 2.3.14 Comments: BANK - Callaway Lesson Plan Objective: LO1734020 R1, Determine exposure limits and posting requirements in the ALARA program.

Tier # 3 Group #

Last Used - 2013 Callaway #98 Fundamental 55.43 part 4 KA - Rad control - Knowledge of rad or contamination hazards that may arise during normal abnormal or emergency conditions or activities

(#98 2011 ANO2 - least exposure)

Rev 0

Q99, Rev 1 ID: 98414 Points: 1.00 The crew is recovering from a reactor trip that occurred 5 minutes ago due to an I&C technician inadvertently manipulating incorrect switches in the back of the Control Room. The following indications are noted:

• Crew is performing EMG ES-02, REACTOR TRIP RESPONSE

• RCS pressure 2235 psig stable

• RCS temperature 555°F lowering at 2°F per minute (reason unknown)

• RCS T ~1 to 2 °F

• PZR level 26% stable

• AFW flow 100,000 lbm/hr - throttled per fold out page

• MSIVs Closed Assuming the above conditions do NOT change over the next 35 minutes which of the following actions will be taken?

A. Direct an SI and transition to EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

B. Transition to EMG FR H-1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.

C. Transition to OFN BG-009, EMERGENCY BORATION, then return to EMG ES-02 step in effect.

D. Direct the RO to perform OFN BG-009, EMERGENCY BORATION, while the CRS and the BOP continue with EMG ES-02.

Answer: D Answer Explanation:

SRO because to answer you have to know the procedure content (steps) to understand that the emergency boration will take place concurrently with the EMG in effect. Comes from AP 15C-003 Correct - Per step 2, continuous action step, if temperature lowers to less than 530°F and no SI (that is why we are in ES-01) then borate so you don't lose SDM. This is how OFN's and EMG's are performed together.

Incorrect - transition to OFN BG-009. This procedure is needed due to the cooldown but ES-02 directs in step 2 RNO that both procedures are to be completed concurrently.

Plausible if the student doesn't know about the temperature requirement AND the use of concurrent OFN and EMG.

Incorrect - direct an SI and transition to E-0. With temperature lowering a jump in logic could be made that will also lower pressure and PZR level which would require an SI. No other items are broke so PZR heaters will maintain pressure and charging will maintain level. Plausible if the student assumes other things are broke outside of what the stem states.

Incorrect - transition to FR H-1. With AFW flow less than 270,000 lbm/hr an assumption could be made that more flow is needed but since flow was throttled per fold out page then SG levels, at least one, must be over 6% NR. Plausible if the student doesn't understand why AFW flow was throttled.

Meets KA asks how to use OFNs with EOPs OPS INITIAL NRC Page: 1 of 2 14 October 2015

SRO knowledge procedure selection based on indications Question Q99, Rev 1 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 98414 User-Defined ID: 98414

Reference:

EMG ES-02 Topic: 99 SRO OFN procedure use with EMG's RO Importance Rating: 3.8 SRO Importance Rating: 4.5 K/A Number: 2.4.8 Comments: BANK - DC Cook Lesson Plan Objective: LO1733203 R6, DISCUSS procedure implementation IAW AP 15C-002.

Tier # 3 Group #

Last Used - 2011 DC Cook #99 Comprehension 55.43 part 5 KA - Emergency procedures - Knowledge of how abnormal operating procedures are used in conjunction with EOP's Modification History:

1 - revised based on NRC comments 10/5 Associated objective(s):

SRO 25 OPS INITIAL NRC Page: 2 of 2 14 October 2015

100 ID: 98415 Points: 1.00 The unit is operating at 100% power when a RED first out for OTT comes in.

The alarm is verified correct but the reactor does NOT trip.

The CRS directs entry into EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION / ATWS.

The crew performs their immediate actions.

The crew is now verifying the reactor and turbine trip when an SI occurs.

Which of the following statements is correct with regards to what action(s) the CRS will direct next AND what is the basis for this action?

A. Direct the RO to return to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, and perform steps 1-7, while the BOP and the CRS continue with EMG FR-S1.

This is to verify the emergency borate flow remains above 30 gpm until SDM is determined.

B. Direct the RO to return to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, and perform steps 1-7, while the BOP and the CRS continue with EMG FR-S1.

This action is to verify that all of the SI actuated equipment has actuated.

C. Continue with the EMG FR-S1 until a transition is made back to EMG E-0 step

1. This action is to verify that all of the SI actuated equipment has actuated.

D. Continue with EMG FR-S1 until a transition is made back to EMG E-0 step 1.

This is to verify the emergency borate flow remains above 30 gpm until SDM is determined.

Answer: B Answer Explanation:

Correct - At the step the crew is in they have passed the continuous action step to check for SI. This step sends the crew to perform the first 7 steps of E-0 and then when the CRS completes FR-S1 by moving out to step 25 and 27 a transition will be made to E-0 step 8 since the RO has completed the auto SI verification steps. The bases is there for the SI verification, all the equipment functioned properly and the SI is valid.

Incorrect - Direct the RO back to E-0 and verify emergency boration flow is above 30 gpm. The first part is correct the crew will send an operator back to E-0 to verify the SI. The bases is wrong because even if the emergency boration was started and flow verified prior to the SI when the SI happens the BA transfer pumps are load shed, the emergency borate valve is load shed, and the emergency borate flow goes to 0. This is because the SI is now injecting not the emergency flow path. Plausible if the student doesn't understand what actions take place on the SI.

Incorrect - Continue with FR-S1 and to verify SI. If the crew does not send an operator back to E-0 and then makes it to step 27 they could transition to E-0 at step 1 or step 8, the distinction is whether SI has occurred or not. The verification of the SI must be completed by either the operator sent back to perform it or the crew. The second part is correct to verify the SI. Plausible if the student understands the verification part but doesn't catch the fact that the first part is sending them back to step 1 so the verification will not be made.

Incorrect - Continue with FR-S1 and verify emergency borate flow above 30 gpm.

If the crew does not send an operator back to E-0 and then makes it to step 27 they will transition to E-0 at step 1 or step 8. The verification of the SI must be completed by either the operator sent back to perform it or the crew. The bases is wrong because even if the emergency boration was started and flow verified prior to the SI when the SI happens the BA transfer pumps are load shed, the emergency borate valve is load shed, and the emergency borate flow goes to 0.

This is because the SI is now injecting not the emergency flow path. Plausible if the student doesn't understand what actions take place on the SI and forgets that the SI verification steps must be completed.

Meets KA asks SRO knowledge because it asks about knowledge of the content of the procedure, the procedure steps and sequence must be known to understand what has been completed and what has not.

Question 100 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 4.00 System ID: 98415 User-Defined ID: 98415

Reference:

BD EMG FR S-1 100 SRO knowledge of bases for FR-S1 procedure steps Topic:

for E-0 return RO Importance Rating: 3.4 SRO Importance Rating: 4.4 K/A Number: 2.4.23 Comments: NEW Lesson Plan Objective: LO1732339 R3, Given an EMG FR-S1 procedure flow path, EXAMINE the available options for procedure actions.

Tier # 3 Group #

Last Used - N/A Comprehension 55.43 part 5 KA - Emergency procedures - Knowledge of the basis for prioritizing emergency procedure implementation during emergency operations