ML15163A231

Draft Written Exam (Folder 2)

ML15163A231
Person / Time
Site:	Indian Point
Issue date:	06/04/2015
From:	D'Antonio J Operations Branch I
To:	Entergy Nuclear Indian Point 3
Shared Package
ML14352A136	List: IR 05000286/2015301 ML14363A296 ML15160A409 ML15160A456 ML15160A464 ML15160A470 ML15163A229 ML15163A230 ML15163A231 ML15163A232
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TAC U01908
Download: ML15163A231 (141)

Text

ES-401 Site-Specific RO Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: 5/18/15 Facility/Unit: Indian Point Unit 3 Region: I x II D Ill D IV D Reactor Type: W X CED BW D GED Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant's Sionature Results Examination Value Points Applicant's Score Points Applicant's Grade Percent

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000008AK305 Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: ECCS termination or throttling criteria Importance: 4.0 4.5 Question: #1 The crew is performing ECCS reduction in ES-1.2, Post LOCA Cooldown and Depressurization.

One SI pump has been stopped.

Which of the following describes how the required subcooling changes and the basis for the amount of subcooling required for securing the second SI pump?

a. Required Subcooling decreases due to decreased break flow after stopping the second pump.

b. Required Subcooling decreases to allow pressure to decrease sufficiently after stopping the second pump to place RHR cooling in service.

c. Required Subcooling increases to compensate for the anticipated void formation after the second pump is stopped.

d. Required Subcooling increases to ensure the RCS will remain subcooled after the second SI pump is stopped.

Answer: d Explanation I Justification

a. Incorrect. Plausible because break flow does decrease after stopping each pump; however, subcooling requirement increases.

b. Incorrect. Plausible because the most likely exit from ES-1.2 is starting an RHR pump in injection mode; however, subcooling requirement increases.

c. Incorrect. Plausible because required subcooling increases; however, no void formation is expected or should exist because the RCS is subcooled.

d. Correct because throttling SI flow is not practical, pumps must be stopped. As pumps are stopped, a step decrease in flow is experienced. RCS pressure will decrease to reach equilibrium with break flow and ECCS flow. RCS subcooling and pressure control must be maintained as the reduction takes place. As flow is reduced, a large allowance is needed.

Technical

References:

13LP-ILO-EOPE10, Rev. 2

Proposed References to be provided: None Learning Objective: Objective 15 Question Source: Bank, IP2 2012 NRC Exam Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)5 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000009EA109 Ability to operate and monitor the following as they apply to a small break LOCA: RCP Importance: 3.6 3.6 Question: #2 Indian Point 3 has experienced a Reactor Trip and a Safety Injection. E-0 (Reactor Trip or Safety Injection) Immediate Operator Actions have been completed. The following conditions are observed:

All Control Rods are fully inserted All Reactor Coolant Pumps are running RCS Pressure is 1185 psig and slowly lowering RCS Average Temperature is 547°F The hottest incore thermocouple is reading 570°F Containment Pressure is 2 psig and slowly rising 31 High Head Safety Injection Pump tripped on overcurrent 32 High Head Safety Injection Pump is tagged for motor replacement 33 High Head Safety Injection Pump is running Analyze and describe the required Reactor Coolant Pump configuration by procedure and why?

a. All Reactor Coolant Pumps running, to delay the onset of inadequate cooling

b. All Reactor Coolant Pumps stopped, to prevent excessive core uncovery

c. All Reactor Coolant Pumps stopped, less than two HHSI pumps are running

d. All Reactor Coolant Pumps running, adequate RCS subcooling exists Answer: b Explanation I Justification

a. Incorrect. Plausible because if there were no HHSI pumps running, the RCPs are kept running to delay the onset of inadequate core cooling. At least one HHSI pump must be running to stop the RCPs. Student may believe that a minimum of 2 HHSI pumps are required.

b. Correct. At least one HHSI pump is running and subcooling is less than specified in table.

The trip criteria purpose is to minimize high PCTs due to excessive core uncovery caused by continued RCP operation during a SBLOCA.

c. Incorrect. Plausible because the design bases minimum Engineered Safeguards Systems would be two HHSI pumps running and subcooling is less than specified in table.

d. Incorrect. Plausible because the student may believe that adequate subcooling exists.

Question requires the use of steam tables to determine the amount of subcooling and knowledge of the subcooling required by the EOP.

Technical

References:

13LP-ILO-EOPE10, Rev. 2 Proposed References to be provided: None Learning Objective: Objective 4.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000011EK101 Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA: Natural circulation and cooling, including reflux boiling Importance: 4.1 4.4 Question: #3 During which of the following accident scenarios are natural circulation and reflux boiling not important.

a. A LOCA with leak rate such that SI flow is less than break flow and core boiling is occurring.

b. A LOCA with a leak rate such that the reactor coolant pressure stabilizes at a pressure higher than SG pressure.

c. A LOCA with break flow cooling less than decay heat and the enthalpy of the RCS is rising.

d. A LOCA with break flow cooling greater than decay heat and the pressure of the RCS is rapidly lowering.

Answer: d Explanation I Justification

a. Incorrect. Plausible because natural circulation and reflux cooling are required when ECCS flow and break flow are not adequate to remove decay heat. These conditions are indicative of a SBLOCA that requires supplemental heat removal via the steam generators to adequately cool the core.

b. Incorrect. Plausible because natural circulation and reflux cooling are required when ECCS flow and break flow are not adequate to remove decay heat. These conditions are indicative of a SBLOCA that requires supplemental heat removal via the steam generators to adequately cool the core.

c. Incorrect. Plausible because natural circulation and reflux cooling are required when ECCS flow and break flow are not adequate to remove decay heat. These conditions are indicative of a SBLOCA that requires supplemental heat removal via the steam generators to adequately cool the core.

d. Correct. A LBLOCA (design bases) will result in break flow cooling greater than decay heat, resulting in 1. Slowdown (rapid depressurization & core uncovery), 2. Refill (Accumulator injection), 3. Reflood (ECCS Flow), and 4. Long-term recirculation. Both break flow and ECCS flow are adequate for core cooling.

Technical

References:

13LP-ILO-EOPE10, Rev. 1 Proposed References to be provided: None Learning Objective: Objective 1.0 & 2.0 Question Source: Modified Bank, Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 10 & 14

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000015AK208 Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: CCWS Importance: 2.6 2.6 Question: #4 Indian Point 3 is at 100% Power. 3-AOP-RCP-1; Reactor Coolant Pump Malfunction has been entered due to elevated shaft vibration and rising lower motor bearing temperatures on 31 Reactor Coolant Pump. The cause of the elevated vibration and temperature readings are being investigated. The following conditions are presently observed:

Two CCW Pumps are in service.

31 RCP Shaft Vibration has stabilized at 15 mils.

31 RCP Lower Motor Bearing Temperature is 185°F and rising at approximately 1°/hour.

32, 33, and 34 RCP vibration and motor bearing temperature readings are all normal.

Based on the present observed conditions, what actions are required In accordance with the AOP?

a. Trip the reactor, stop 31 RCP, and Initiate E-0; Reactor Trip or Safety Injection due to 31 RCP Shaft Vibration Readings 2 15 mils.

b. Initiate a normal plant shutdown (POP-2.1 Operation at Greater than 45% Power, POP-3.1 Plant Shutdown from 45% Power) and secure 31 RCP prior to reaching the high vibration limit.

c. Trip the reactor, stop 31 RCP, and Initiate E-0; Reactor Trip or Safety Injection due to 31 RCP Lower Motor Bearing Temperature Readings 2 185°F.

d. Initiate a normal plant shutdown (POP-2.1 Operation at Greater than 45% Power, POP-3.1 Plant Shutdown from 45% Power) and secure 31 RCP prior to reaching the high bearing temperature limit.

Answer: d Explanation I Justification

a. Incorrect. Plausible because IAAT Shaft Vibration is >15 mils and rising greater than 1.0 mils/hour, these action are taken IAW the AOP. However, shaft vibration is equal to 15 mils and has stabilized there.

b. Incorrect. Plausible because a plant shutdown is initiated if Shaft Vibration is > 15 mils and is trending ~ 0.5 mils/hour rise over a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period. However, shaft vibration is equal to 15 mils and has stabilized there.

c. Incorrect. Plausible because at> 200°F, these actions are taken IAW the AOP.

d. Correct. Per the AOP; IAAT motor bearing temperature for the affected RCP is~ 185°F, and rising, then initiate plant shutdown with goal of stopping affected RCP prior to reaching 200°F motor bearing temperature. Operationally valid as the lower bearing CCW flow could be below the nominal 5 gpm and still not cause a total low bearing coolant flow alarm at 155 gpm. Low CCW flow to the lower bearing cooler alone could cause both elevated lower bearing temperatures and elevated shaft vibration.

Technical

References:

13LP-I LO-AOPRCP Proposed References to be provided: None Learning Objective: Objective 3.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000022AK 101 Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Consequences of thermal shock to RCP seals.

Importance: 2.8 3.2 Question: #5 Operators have transitioned to 3-ECA-0.0; Loss of All AC Power due to the loss of all the 480V AC Safeguards Buses. Efforts to restore AC Power have been unsuccessful and the crew has determined the need to isolate RCP seal cooling per the procedure. How is RCP cooling isolated IAW ECA-0.0 and why is it isolated?

a. Close CH-MOV-222, Seal Water Return Isolation Valve only. To prevent hot RCS inventory from returning to the VCT.

b. Close both FCV-625, CCW Return from RCP Thermal Barrier and CH-MOV-222, Seal Water Return Isolation Valve. To ensure any RCS leakage from the seals remains in the containment.

c. Close CH-MOV-222, Seal Water Return Isolation Valve, FCV-625, CCW Return from RCP Thermal Barrier, and all four RCP Seal Injection (needle) Valves. To ensure that seal cooling remains isolated when power is restored and charging is re-established, this will prevent seal and shaft damage due to thermal shock.

d. Close all four RCP Seal Injection (needle) Valves and CH-MOV-222, Seal Water Return Isolation Valve. To prevent vapor binding due to potential elevated VCT and charging suction temperatures when power is restored and charging is re-established.

Answer: c Explanation I Justification

a. Incorrect. Plausible because the student may believe that seal cooling need only be isolated on the return to ensure no leakage exits the containment. Student may believe that since component cooling system is intact, isolation is not required. This would help prevent hot RCS inventory from returning to the CVCS system via seal return.

b. Incorrect. Plausible because the student may believe that seal cooling need only be isolated on the return to ensure no leakage exits the containment.

c. Correct. If neither thermal barrier cooling nor seal injection exists (as would be the case on a loss of all 480V Vital AC Buses), the procedure has you isolate all four individual seal injection valves, seal return, and component cooling return from the thermal barrier.

Following restoration of AC power, it is desirable to keep the RCP seal cooling isolated to prevent thermal shock of the RCP seals and shaft.

d. Incorrect. Plausible because the student may believe that since component cooling system is intact, isolation is not required. A procedure caution states that "High VCT temperature, as indicated on seal water return Tl-117, may cause vapor binding when charging pump is started."

Technical

References:

13LP-ILO-EOPCOO, 3-ECA-0.0 Proposed References to be provided: None Learning Objective: Objectives 8.0 & 11.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 3 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 00002582413 Loss of RHR System: Knowledge of crew roles and responsibilities during EOP usage Importance: 4.0 4.6 Question: #6 The unit is performing a plant cool down on RHR with 34 RCP operating in preparation for entering Cold Shutdown. A human error by an l&C Technician has caused PT-403 to fail high.

What is the effect of the error on the plant and what actions are required to mitigate the effect?

a. AC-MOV-730 will close resulting in a loss of RHR. 3-AOP-RHR-1 will immediately sound the containment evacuation alarm, initiate containment closure, and stop the running RHR pump.

b. AC-MOV-731 will close resulting in a loss of RHR. 3-AOP-RHR-1 will immediately stop the running RHR pump, reopen AC-MOV-731, and restart the standby RHR pump.

c. AC-MOV-730 and AC-MOV-731 will close resulting in a loss of RHR. 3-AOP-RHR-1 will immediately stop the running RHR pump, reopen both AC-MOV-730 and AC-MOV-731, and restart the standby RHR pump.

d. AC-MOV-731 will close resulting in a loss of RHR. The operator should trip the running RHR pump in accordance with the alarm response procedure. 3-AOP-RHR-1 will direct the operator to maintain RCS temperature < 350°F using atmospheric or condenser steam dumps.

Answer: d Explanation I Justification

a. Incorrect. Plausible because these actions would be performed if the loss of RHR was to occur at::; 200°F in accordance with 3-AOP-RHR-1. Plant is in Mode 4 however, therefore RCS temperature is> 200°F and the steam generators are still available. Also PT-403 is interlocked with MOV-731 not 730, student may not remember which pressure transmitter is interlocked with which valve.

b. Incorrect. Plausible because PT-403 is interlocked with MOV-731. Also plausible because the steps make sense if restoring RHR is a priority. However, with PT-403 still failed high the valve will not reopen. Also, these steps are not taken if RCS temperature is > 200°F.

c. Incorrect. Plausible because the student may believe that both RHR suction valves close whether PT-402 or PT-403 fails. Also plausible because the steps make sense if restoring RHR is a priority.

d. Correct. PT-403 is interlocked with MOV-731. 3-ARP-010 directs the operator to trip the running RHR pump and directs the operator to 3-AOP-RHR-1. If RCS temperature is>

200°F, step 4.47 & 4.48 direct the operator to maintain RCS temperature< 350°F using atmospheric or condenser steam dumps.

NOTE: Reactor Operator fundamental knowledge of the valve interlocks and RHR pump operation plus comprehension understandings (question stem) regarding the fact that the unit is MODE 4 will allow the RO to answer this question.

Technical

References:

13LP-ILO-AOPRHR, 3-AOP-RHR-1 System Description 4.2, RHR 3-ARP-010 (Panel SGF)

Proposed References to be provided: None Learning Objective: Objective 1.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)5, 7, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000026AA203 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The valve lineups necessary to restart the CCWS while bypassing the portion of the system causing the abnormal condition Importance: 2.6 2.9 Question: #7 The Unit was in MODE 3 at normal operating temperature and pressure when a large Component Cooling Water Leak occurred resulting in the implementation of 3-AOP-CCW-1; Loss of Component Cooling Water. The Component Cooling Water Headers were originally combined and actions were taken based on both surge tank levels lower than 5%.

Following leak identification and isolation, the Component Cooling System is being restarted in accordance with 3SOP-CC-001 B; Component Cooling System Operation. What precautions and limitations are applicable during startup of the system?

a. ADJUST SW flow through both HXs to maintain CCW outlet temperature below 120°F while limiting the HX Outlet Isolation Valves to a maximum opening of 27.5° and 27° respectively.

b. ADJUST SW flow through both HXs to maintain CCW outlet temperature below 110°F while limiting the HX Outlet Isolation Valves to a maximum opening of 27.5° and 27° respectively.

c. ADJUST SW flow through both HXs to maintain CCW outlet temperature below 110°F while limiting the HX Outlet Isolation Valves to a maximum opening of 37.5° and 37° respectively.

d. ADJUST SW flow through both HXs to maintain CCW outlet temperature below 120°F while limiting the HX Outlet Isolation Valves to a maximum opening of 37.5° and 37° respectively.

Answer: b Explanation I Justification

a. Incorrect. Plausible because when placing RHR in service, then a CCW inlet temperature of 120°F is allowable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The maximum opening of the valves is correct also. Unit is in MODE 3, RHR is not being placed in service.

b. Correct. The procedure states to adjust SW flow through the HXs to maintain CCW outlet temperature below 110°F, while limiting the HX outlet valves to a maximum opening of 27.5° and 27° respectively.

c. Incorrect. Plausible because the temperature requirement of 110°F is correct, but the valve opening limitation is incorrect.

d. Incorrect. Plausible because when placing RHR in service, then a CCW inlet temperature of 120°F is allowable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The maximum opening of the valves is correct also. Unit is in MODE 3, RHR is not being placed in service. The valve opening limitation is also incorrect.

Technical

References:

3-SOP-CC-001 B, Rev. 33, 13LP-ILO-CCW001 Proposed References to be provided: None Learning Objective: Objectives 1.0 & 2.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 4 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000027 AK303 Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: Actions contained in EOP for PZR PCS malfunction Importance: 3.7 4.1 Question: #8 Indian Point 3 has experienced a Reactor Trip and a Safety Injection. E-0 (Reactor Trip or Safety Injection) has been completed through step 13. Step 14 is checking the status of Pressurizer PORVs, Safeties and Spray Valves. The following conditions are observed:

All Reactor Coolant Pumps are running RCS Pressure is 1700 psig and lowering Both PZR PORVs are closed PZR Safety Tailpipe Temperatures and Acoustic Monitors are normal Both PZR Spray Valves are opened The ATC Operator has attempted to manually close the Spray Valves, but they remain full open.

Describe the next action attempted by the procedure to mitigate these malfunctioning spray valves.

a. If valve(s) NOT closed, THEN STOP ALL RCP(s) to stop spray flow.

b. If valve(s) NOT closed, THEN STOP ANY RCP required to stop spray flow.

c. If valve(s) NOT closed, PULL CONTROLLER FUSE for failed spray valve(s) to stop spray flow.

d. If valve(s) NOT closed, THEN GO TO 3-E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

Answer: c Explanation I Justification

a. Incorrect. Plausible because stopping the RCPs will stop the RCS depressurization caused by the failed open spray valves. However, the EOP deviation document states that the procedure reflects IP3's design by addressing removal of control power fuses. This addition promotes the ERG's intent for this substep's RNO and attempts to do so without forcing an RCP be removed from service.

b. Incorrect. Plausible because stopping the required RCPs will stop the RCS depressurization caused by the failed open spray valves. However, the EOP deviation

document states that the procedure reflects IP3's design by addressing removal of control power fuses. This addition promotes the ERG's intent for this substep's RNO and attempts to do so without forcing an RCP be removed from service. This step is also performed if the valves fail to close by removal of the control power fuses. Question asked for the "next" step after attempting to

c. Correct. The EOP deviation document states that the procedure reflects IP3's design by addressing removal of control power fuses. This addition promotes the ERG's intent for this substep's RNO and attempts to do so without forcing an RCP be removed from service.

a. Incorrect. Plausible because this is the action taken if the PORVs and their associated Block Valves cannot be closed.

Technical

References:

13LP-ILO-EOPEOO, 3-E-Or4 Proposed References to be provided: None Learning Objective: Objective 12.0 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 5, 7, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000029EK206 Knowledge of the interrelations between the and the following an ATWS: Breakers, relays, and disconnects Importance: 2.9 3.1 Question: #9 Given the following:

• Reactor Trip testing is in progress on Train "A".

• Reactor Trip Breaker "A" is open.

• Reactor Trip Bypass Breaker "A" is closed.

• A transient occurs requiring a reactor trip.

• The RO attempts to manually trip the reactor but the reactor does NOT trip.

Which ONE of the following describes a malfunction that has contributed to the reactor trip failure?

a. Undervoltage Trip coils failed to energize; Shunt Trip coils failed to de-energize.

b. Undervoltage Trip coils failed to energize; Shunt Trip coils failed to energize.

c. Undervoltage Trip coils failed to de-energize; Shunt Trip coils failed to energize.

d. Undervoltage Trip coils failed to de-energize; Shunt Trip coils failed to de-energize.

Answer: c Explanation I Justification

a. Incorrect. Plausible because the student may believe that the UV coils energize to trip and the Shunt coils de-energize to trip.

b. Incorrect. Plausible because the student may believe that the UV coils energize to trip.

c. Correct. UV coils de-energize to trip and the Shunt coils energize to trip (DC).

d. Incorrect. Plausible because the student may believe that the Shunt coils de-energize to trip.

Technical

References:

13LP-ILO-ICRXP Proposed References to be provided: None Learning Objective: Objective 1.0 Question Source: Bank (Indian Point Common)

Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000054AA206 Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW): AFW adjustments needed to maintain proper T-ave. and SIG level Importance: 4.0 4.3 Question: #10 Indian Point Unit 3 is at 100% when the 31 MBFP trips, operators take actions per AOP-FW-1; Loss of Feedwater. Which of the following describes the status of the Main and Auxiliary Feedwater Systems after the initial procedural actions are taken by the operators?

a. Main Feedwater (32 MBFP) is still in service as operators have reduced load to less than 650MWe. Auxiliary Feedwater is not in service.

b. Main Feedwater (32 MBFP) is still in service as operators have reduced load to less than 650MWe. AOP-FW-1 also starts both 31 and 33 ABFPs.

c. Main Feedwater is isolated due to a Feedwater Isolation Signal and the Auxiliary Feedwater System is feeding the Steam Generators at greater than 365 gpm flow until at least one SG NR level is greater than 9%.

d. Main Feedwater is isolated due to a Feedwater Isolation Signal and the Auxiliary Feedwater System is feeding the Steam Generators at greater than 365 gpm flow until all SG NR levels are greater than 14%.

Answer: c Explanation I Justification

a. Incorrect. Plausible because if initial power level is::;; 80%, per attachment 1, load is reduced to < 650 MWe. However, the AOP also starts the motor driven aux feed pumps as well.

b. Incorrect. Plausible because if initial power level is ::;; 80%, per attachment 1, load is reduced to < 650 MWe. The AOP also starts the motor driven aux feed pumps as well.

c. Correct. Because power is > 80%, the AOP directs the operator to trip the reactor and go to E-0; Reactor Trip or Safety Injection. Once in ES-0.1; Reactor Trip Response, the operator would verify proper auxiliary feedwater flow > 365 gpm until at least one SG NR level is greater than 9%. The main and bypass FRVs receive a FW Isolation Signal(< 554°F & Trip Breakers open)

d. Incorrect. Plausible because power is > 80% and the AOP directs the operator to trip the reactor and go to E-0; Reactor Trip or Safety Injection. Once in ES-0.1; Reactor Trip Response, the operator would verify proper auxiliary feedwater flow > 365 gpm until at least

one SG NR level is greater than 9%, not 14%. 14% is plausible as it is the minimum level for adverse containment conditions. The main and bypass FRVs receive a FW Isolation Signal(< 554°F & Trip Breakers open)

Technical

References:

13LP-ILO-AOPFW1 13LP-I LO-EOPEOO 3-ES-0.1 3-AOP-FW-1 Proposed References to be provided: None Learning Objective: Objective 3.0 and Objective 12.0 respectively Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 {b) 5, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000055EK102 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout: Natural circulation cooling Importance: 4.1 4.4 Question: #11 The unit experiences a reactor trip concurrent with a loss of off-site power. The Emergency Diesel Generators have either failed to start or tripped, a loss of all 480V buses has occurred.

Ten minutes after the loss of all 480V bus event, the following plant conditions exist:

• All SIG Pressures are approximately 1030 psig and stable

• RCS Pressure is 2185 psig and stable

• T-hot is approximately 575°F in all loops and lowering slowly

• Core Exit Thermocouples indicate approximately 580°F

• T-cold is approximately 550°F in all loops and stable Based on the above conditions, what is the present status of RCS cooling and what procedural actions are required to attempt to mitigate the event if a vital 480V AC bus is not restored?

a. Natural Circulation exists. Initiate a rapid SG Depressurization to allow the RCS Accumulators to inject even if voiding exists in the reactor vessel head.

b. Natural Circulation exists. Initiate a rapid RCS Depressurization by opening both PORVs to allow the RCS Accumulators to inject.

c. Natural Circulation does not exist. Initiate a rapid SG Depressurization to allow the RCS Accumulators to inject even if voiding exists in the reactor vessel head.

d. Natural Circulation does not exist. Initiate a rapid RCS Depressurization by opening both PORVs to allow the RCS Accumulators to inject.

Answer: a Explanation I Justification

a. Correct. Based on given plant conditions, natural circulation exists. Saturation temperature for the SG Pressure of 1030 psig is 550°F and the saturation temperature for RCS Pressure of 2185 psig is 649.5°F. Therefore all parameters are trending properly and RCS subcooling is> 40°F (69.5°F). Additionally, step 20 of 3-ECA-0.0 initiates a rapid SG depressurization even if voids form in the head (note prior to step 20).

b. Incorrect. Plausible because based on given plant conditions, natural circulation exists.

See above. However, the Loss of all AC procedure does not initiate a rapid RCS depressurization using the PORVs. Plausible because the Red Path FR-H.1 does initiate a "Bleed and Feed" using PORVs.

c. Incorrect. Plausible because steam tables and simple calculations are required, student also needs to know EOP Natural Circulation criteria. Second part of distractor is correct, step 20 of 3-ECA-0.0 initiates a rapid SG depressurization even if voids form in the head (note prior to step 20).

d. Incorrect. Plausible because steam tables and simple calculations are required, student also needs to know EOP Natural Circulation criteria. However, the Loss of all AC procedure does not initiate a rapid RCS depressurization using the PORVs. Plausible because the Red Path FR-H.1 does initiate a "Bleed and Feed" using PORVs.

Technical

References:

13LP-ILO-EOPCOO Proposed References to be provided: None Learning Objective: Objectives 3.0 & 11.0 Question Source: New Question Cognitive Level: Comprehension I Analysis 10CFR Part 55 Content: 55.41 (b) 8 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000057G2411 Loss of Vital AC Electrical Instrument Bus:

Knowledge of abnormal condition procedures Importance: 4.0 4.2 Question: #12 The plant is in a normal 100% power lineup with Channel B selected for feed flow and steam flow control. Which set of actions will be required per 3-AOP-IB-1, Loss of Power to an Instrument Bus, for a loss of 31 and 31A Instrument Susses?

I. Swap feed flow and steam flow to the A channel

11. Place MBFPs in Manual control Ill. Place MFRVs in Manual control

a. I and II

b. I and Ill

c. II and Ill

d. I, 11, and 111 Answer: a Explanation I Justification

a. Correct. 3-AOP-IB-1, step 4.6 places all steam flow and feedwater flow channel transfer switches in position "A". Step 4.10 places all MBFPs in manual control.

b. Incorrect. Plausible because Ill (MFRVs in manual) is a required action for a loss of 34 Instrument Bus. Selection II not being required is plausible because only one MBFP is affected by this failure and it goes into track and hold, however the procedure specifies going to manual.

c. Incorrect. Plausible because Ill (MFRVs in manual) is a required action for a loss of 34 Instrument Bus. Selection I not being needed is plausible because the plant does not trip if this is not done, however the procedure specifies this action.

d. Incorrect. Plausible because Ill (MFRVs in manual) is a required action for a loss of 34 Instrument Bus.

Technical

References:

13LP-ILO-AOPIB1 R1, 3-AOP-IB-1; Rev. 8 Proposed References to be provided: None Learning Objective: Objective 3.0 Question Source: Bank (IP3)

Question Cognitive Level: Knowledge

10CFR Part 55 Content: 55.41 (b) 10 Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000058AK302 Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: Actions contained in EOP for loss of de power Importance: 4.0 4.2 Question: #13 During the performance of ES-0.1, Reactor Trip Response, the operators are directed to manually open generator breakers 1 and 3 if they do not open automatically 30 seconds following a turbine trip. If a loss of DC control power occurred for generator breakers 1 and 3, the operator is directed by 3-AOP-DC-1, "Loss of a 125V DC Panel", to locally trip generator breakers 1 and 3 if a unit trip has occurred and the generator breakers 1 and 3 have not tripped.

Which of the following is the overriding concern?

a. Reverse power in the Main Transformer

b. Tripping of RCP's on Low Frequency

c. Motorizing the Main Generator

d. Unit Auxiliary Transformer damage Answer: c Explanation I Justification

a. Incorrect. Plausible because student may believe that the main transformer is the concern.

Main transformer can send power either way. It is not a generator.

b. Incorrect. Plausible because student may believe that RCP pumping capability is the concern. Grid will maintain proper frequency.

c. Correct. Breakers must be opened to disconnect from grid.

d. Incorrect. Plausible because student may believe that the Unit Aux Transformer is the concern. UAT will not be damaged by being connected, not a generator.

Technical

References:

13LP-ILO-AOPDC1 Proposed References to be provided: None Learning Objective: Objective 2.0 Question Source: Bank - IP3 2006 NRC Exam Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 {b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 0000622139 Loss of Nuclear Service Water: Knowledge of conservative decision making practices Importance: 3.6 4.3 Question: #14 Unit 3 is operating at 100%. A low pressure condition exists in the Service Water System due to a leak on the essential service water header. The crew has entered 3-AOP-SW-1; Service Water Malfunction.

Given the following conditions:

• Essential Service Water Pressure is 55 psig

• No temperature alarms associated with essential service water header cooled components

• Leak isolation requires the entire Essential Service Water Header to be isolated

• Maintenance has estimated that the repair will take 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> What operator actions should be taken in accordance with 3-AOP-SW-1?

a. Trip the reactor and Initiate E-0, continue in parallel with 3-AOP-SW-1 and transfer essential headers IAW 3-SOP-RW-005; Service Water System Operation. Following transfer of essential headers, isolate the leak.

b. Initiate plant shutdown to MODE 3, initiate transfer of essential headers IAW 3-SOP-RW-005; Service Water System Operation, and when the leak can be isolated, isolate leak.

c. Maintain Reactor Power, transfer essential headers IAW 3-SOP-RW-005; Service Water System Operation, and when leak can be isolated, isolate leak.

d. Initiate plant shutdown to MODE 3 and when the leak can be isolated, isolate leak. Non-essential service water header is required to remain operable in MODE 3, so the procedure does not initiate transfer of the essential headers.

Answer: b Explanation I Justification

a. Incorrect. Plausible because if service water pressure is < 50 psig and system cooling ability is affected, then operator is directed to trip the reactor and initiate E-0. Not required because essential service water header pressure is 55 psig. Remaining actions are IAW procedure.

b. Correct. Because service water pressure is> 50 psig, a reactor trip is not required.

However, since isolation of the leak requires the entire Essential Service Water Header to

be isolated (shutdown), a plant shutdown to MODE 3 is initiated (see steps 4.30-4.35 of 3-AOP-SW-1 ). Because the essential header is the affected header, the procedure first initiates transfer of essential headers IAW 3-SOP-RW-005; Service Water System Operation.

c. Incorrect. Plausible because the student may believe that technical specifications allow inoperability of the essential header (one header) for greater than the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> repair estimate. Student could believe it is a standard 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement. It is actually a 12 action statement.

d. Incorrect. Plausible because the plant shutdown to MODE 3 is correct, but although technical specifications require the operability of both headers in MODE 3, the procedure does initiate the transfer of essential headers (Diesels, Instrument Air, Control Room Ventilation, etc.).

Technical

References:

13LP-ILO-SW001; 3-AOP-SW-1 Background document for 3-AOP-SW-1 Proposed References to be provided: None Learning Objective: Objective 6.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 7 & 10 NOTE: Conservative decision making enters into plant shutdown because of the requirement to isolate entire essential header. Conservative decision making also enters into plant shutdown because of the repair estimate by maintenance being 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This is greater than 12 action time. Trip not required because procedural IAAT condition is not met.

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# 000065AA205 Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to commence plant shutdown if instrument air pressure is decreasing Importance: 3.4 4.1 Question: #15 The Unit is at 100% when a line break in the instrument air header in the PAS requires entry into 3-AOP-AIR-1; Air Systems Malfunction.

Which of the following plant conditions require that the operators trip the reactor and initiate E-0 (Reactor Trip or Safety Injection)?

a. Instrument Air Pressure < 90 psig.

b. Volume Control Tank Level lowers to 9%.

c. Multiple AOVs begin repositioning.

d. Pressurizer Level deviation of 5% from program.

Answer: c Explanation I Justification

a. Incorrect. Plausible because IAAT Instrument Air pressure is< 60 psig, the procedure requires tripping the reactor. Also another decision point exists at < 95 psig to send local operator to restore instrument air locally.

b. Incorrect. Plausible because IAAT VCT Level lowers to< 5%, the procedure requires tripping the reactor.

c. Correct. As per step 4.1 of the procedure, IAAT any; which includes_"Multiple AOVs begin repositioning".

a. Incorrect. Plausible because charging pump speed is adjusted per the AOP, however this is not trip criteria.

Technical

References:

13LP-ILO-IA001, Rev. 3, 3-AOP-AIR-1, Rev. 4 Proposed References to be provided: None Learning Objective: Objective 7.0 Question Source: Modified - Bank U2 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)?

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# OOWE04EA 130 Ability to operate and I or monitor the following as they apply to the (LOCA Outside Containment): Desired operating results during abnormal and emergency situations Importance: 3.8 4.0 Question: #16 The crew has completed the actions of ECA-1.2; LOCA Outside Containment. The following plant conditions now exist:

• Pressurizer level is 35% and slowly lowering.

• Steam Generator narrow range levels are 30%.

• Subcooling is 35°F.

• RCS pressure is 1400 psig and slowly lowering.

Which ONE (1) of the following procedures will be performed next?

a. ES-1.2, Post-LOCA Cooldown and Depressurization

b. ECA-1.1, Loss of Emergency Coolant Recirculation

c. E-0, Reactor Trip or Safety Injection

d. E-1, Loss of Reactor or Secondary Coolant Answer: b Explanation I Justification

a. Incorrect. Plausible because ECA-1.1 actions include initiating a cooldown to cold shutdown and depressurizing the RCS to minimize subcooling. Student may remember required actions and assume they could be accomplished with a transition to ES-1.2, Post-LOCA Cooldown and Depressurization.

b. Correct. After all potential leak isolations have been completed, ECA-1.2 checks if RCS Pressure is rising. The RNO is to transition to ECA-1.1, Loss of Emergency Coolant Recirculation because a leak outside containment still exists.

c. Incorrect. Plausible because ECA-1.2 can be entered from E-0 (step 22) and student may believe that the transition is back to procedure and step in effect, however there is no transition to E-0.

a. Incorrect. Plausible because a transition to E-1 is correct if the leak was isolated and RCS Pressure is rising.

Technical

References:

13LP-ILO-EOPC10, 3-ECA-1.2 Proposed References to be provided: None Learning Objective: Objective 7.0

Question Source: Modified IP Common Bank (IP2 2012 Exam)

Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# OOWE05A102 Ability to operate and I or monitor the following as they apply to the (Loss of Secondary Heat Sink): Operating behavior characteristics of the facility Importance: 3.7 4.0 Question: #17 Given the following conditions:

• A reactor trip with Safety Injection (SI) occurs from 100% power

• All SG NR levels are offscale low

• Minimum AFW flow cannot be established

• The crew has entered FR-H.1; "Response to Loss of Secondary Heat Sink"

• Reactor Coolant System pressure is 300 psig and lowering

• Steam Generator pressures are all 600 psig and stable

• The operators are directed by FR-H.1 to return to procedure and step in effect Based on this information, which ONE of the following statements correctly summarizes plant conditions?

a. Steam line rupture in progress and secondary heat sink is NOT required.

b. Large-break Loss of Coolant Accident in progress and secondary heat sink is NOT required.

c. Steam Generator Tube Rupture in progress and secondary heat sink is required.

d. Small-break Loss of Coolant Accident in progress and secondary heat sink is required.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe that feedwater flow has purposely been reduced I isolated by procedural action. The first "Caution" in FR-H.1 states the following; "If total feedflow is less than 365 GPM due to operator action, this procedure should not be performed."

b. Correct. RCS pressure < non faulted SG indicates a larger LOCA, SGs no longer act as a heat sink; core decay heat is being removed by break flow.

c. Incorrect. Plausible because a SGTR is a small break LOCA and requires a secondary heat sink. The student may also believe that actions have been taken to reduce RCS pressure less than ruptured SG pressure.

d. Incorrect. Plausible because a SBLOCA does require a secondary heat sink.

Technical

References:

13LP-ILO-EOPFRH Proposed References to be provided: None Learning Objective: Objective 7.0 Question Source: Bank, IP Common Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10 & 14

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 1 KIA# OOWE12EK202 Knowledge of the interrelations between the (Uncontrolled Depressurization of all Steam Generators) and the following: Facility's 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 Importance: 3.6 3.9 Question: #18 During the performance of ECA-2.1, "UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS", the following condition exists:

• Cooldown rate of the RCS is greater than 100°Flhour How is the operator directed to control feedwater flow?

a. Feedwater flow is terminated to all but a single SIG, preferably an intact SIG.

b. Feedwater flow is maintained at least 365 gpm total until a narrow range level is greater than 9% (14%).

c. Feedwater flow is terminated to all SIGs until temperature stabilizes.

d. Feedwater flow is reduced to 100 gpm to all SIGs with narrow range levels less than 9%

(14%).

Answer: d Explanation I Justification

a. Incorrect. Plausible because a similar strategy is used in FR-H.1 for feeding a hot dry SIG.

b. Incorrect. Plausible because 365 gpm is the minimum normal feedwater flow to ensure adequate heat sink if SIG level is less than 9% (14%) per 3-E-O, Reactor Trip or Safety Injection.

c. Incorrect. Plausible because feedwater flow is terminated (isolated) to individual faulted SIGs in EOP 3-E-2, Faulted Steam Generator Isolation to limit energy blowdown to the containment.

d. Correct. 3-ECA-2.1 states in the RNO for step 2.b (cooldown rate < 100°FIHR) to reduce feed flow to 100 gpm to each SG. Caution also states; "A minimum feed flow of 100 gpm must be maintained to each SG with a narrow range level less than 9% (14%)."

Technical

References:

13LP-ILO-EOPE20, 3-ECA-2.1 Proposed References to be provided: None Learning Objective: Objective 4.0 Question Source: Bank (IP3 - # 24602)

Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# 000001AK102 Knowledge of the operational implications of the following concepts as they apply to Continuous Rod Withdrawal: SUR Importance: 3.6 3.9 Question: #19 Given the following:

• The Reactor is critical at 1o-s amps

• The ATC operator withdraws control rods to continue raising power

• Controls rods continue to move after the In-Hold-Out switch is released Which of the following identifies the INITIAL indications for the event in progress?

SUR PRZR Level Tavg

a. Increase Increase Increase

b. Stable Increase Increase C. Increase Stable Stable

d. Stable Stable Increase Answer: c Explanation I Justification

a. Incorrect. Plausible because these indications are correct if the continuous rod withdrawal occurred> POAH.

b. Incorrect. Plausible because candidate may believe that the increase in Tavg will offset the reactivity from the rods. Below the POAH Tavg and PRZR level will not increase

c. Correct. SUR will continue to increase and PZRZ Level and Tavg will not change below the POAH.

a. Incorrect. Plausible because the student must know the reactor is below the point of adding heat.

Technical

References:

13LP-ILO-AOP-ROD-1, 3-AOP-ROD-1 Proposed References to be provided: None Learning Objective: Objective?

Question Source: Bank (IP3 Question# 25621)

Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)5&6

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# 000003AA203 Ability to determine and interpret the following as they apply to the Dropped Control Rod: Dropped rod, using in-core/excore instrumentation, in-core or loop temperature measurements Importance: 3.6 3.8 Question: #20 Given the following initial plant conditions:

• The operating crew is performing 3-POP-1.3, Plant Startup from Zero to 45% Power

• Indicated NIS Reactor Power is 25%

• Control Bank D is at 149 steps

• Tavg is 552°F The Control Room then receives indication that the Control Bank D Rod located at core location H-8 has dropped into the core. This includes the following indications:

• The Rod Bottom Light for Location H-8 illuminates

• IRPI for Location H-8 indicates 0 inches

• All systems operate as designed in automatic Which of the following additional plant indications will verify that the control rod has dropped and it is not an IRPI failure?

a. Tavg has lowered to 548°F and all four Power Range Nls indicate approximately 23%

Reactor Power.

b. Tavg and Reactor Power have been restored to original values after an Auto Rod withdrawal.

c. Tavg has lowered to 548°F and an asymmetric power distribution is seen by the Power Range Nls, power in one channel lower than the remaining three channels.

d. Both Tavg and Reactor Power Nls indicate an asymmetric power distribution is seen by the Power Range Nls, power in one channel lower than the remaining three channels.

Answer: a Explanation I Justification

a. Correct. Core location H-8 is the actual center of the core, therefore the power distribution will be symmetric, but will lower. Both RCS Temperature and NI power indications will lower.

b. Incorrect. Plausible because the student may believe that rod control will see the temperature deviation and restore it to program value. This cannot happen due to the designed rod stop (Rod Bottom Rod Stop at a minimum).

c. Incorrect. Plausible because in most conditions (rods not in the center of the core), the operator will see an asymmetric power distribution. One quadrant will indicate low, while the other channel will actually indicate higher than original levels.

a. Incorrect. Plausible because in most conditions (rods not in the center of the core), the operator will see an asymmetric power distribution. See FSAR section 14.1.3.1.

Technical

References:

13LP-ILO-ICROD System Description 16.1, Rod Control System Proposed References to be provided: None Learning Objective: Objective G Question Source: Modified Bank Question - # 25643 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 5, 6, 7, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# 000005AK201 Knowledge of the interrelations between the Inoperable I Stuck Control Rod and the following: Controllers and positioners Importance: 2.5 2.5 Question: #21 Given the following Unit 3 conditions:

• The Unit is at 85% reactor power with Control Bank "D" rods at 190 steps.

• A single control rod in Control Bank "D" is at 160 steps.

• The shift has entered 3-AOP-ROD-1; Rod Control and Indicating Systems Malfunction.

• The Rod was verified misaligned, l&C has corrected a blown "movable coil" fuse, and the control rod has been realigned IAW 3-AOP-ROD-1.

• During the implementation of the rod recovery per the procedure, poor procedure place keeping has resulted in the P/A converter not being reset.

After restoring the Rod Control System to Automatic, a plant transient occurs and rods begin to drive in.

Relative to ACTUAL Rod Insertion Limits, at what rod height will the alarm "ROD INSERTION LOW LIMIT" actuate?

a. Alarm will actuate 30 steps ABOVE Rod Insertion Limit

b. Alarm will actuate 24 steps BELOW Rod Insertion Limit

c. Alarm will actuate 30 steps BELOW Rod Insertion Limit

d. Alarm will actuate 24 steps ABOVE Rod Insertion Limit Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe that the 30 step disagreement would translate into 30 step above/below the insertion limit. Student could confuse rod direction and P/A converter readings relative to the rod recovery.

b. Correct. The step counter would have been reset to the stuck rod height and then the stuck rod withdrawn 30 steps in order to realign with the remainder of the bank. Because the PIA converter was not reset, the P/A converter is going to believe that Control Bank "D" is at 220 steps (190+30=220). The ARP 3-ARP-003, page 57 for "Rod Insertion Low Limit" states

that the alarm comes in at 6 steps or 3. 75 inches above variable limit. Therefore the alarm will come in (30-6=24) steps below the actual rod insertion limit.

c. Incorrect. Plausible because the student may believe that the 30 step disagreement would translate into 30 step above/below the insertion limit. Student could confuse rod direction and P/A converter readings relative to the rod recovery.

d. Incorrect. Plausible because 24 steps is the correct answer, but student could confuse rod direction and P/A converter readings relative to the rod recovery.

Technical

References:

13LP-I LO-AOPROD, 3-AOP-ROD-1 Proposed References to be provided: None Learning Objective: Objective 2.0 Question Source: Modified Bank (IP3 Question #20491)

Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 6, 7, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# 000024AK301 Knowledge of the reasons for the following responses as they apply to Emergency Boration: When emergency boration is required Importance: 4.1 4.4 Question: #22 Which of the following conditions would require initiation of EMERGENCY BORATION in accordance with 3-0NOP-CVCS-3; Emergency Boration?

I. Panel SAF - Reactor Coolant System, ROD INSERTION LOW LIMIT Alarm.

II. Panel SAF - Reactor Coolant System, ROD INSERTION LOW-LOW LIMIT Alarm.

Ill. Any control rod greater than or equal to 20 STEPS with the Reactor Trip Breakers open.

IV. Uncontrolled RCS cooldown to less than 547°F with any control rod greater than 20 STEPS with the Reactor Trip Breakers open.

V. More than one control rod is greater than or equal to 20 STEPS with the Reactor Trip Breakers open.

VI. Uncontrolled RCS cooldown to less than 500°F with the Reactor Trip Breakers open.

a. I, IV, V

b. 11, Ill, VI

c. Ill, IV, V

d. II, V, VI Answer: d Explanation I Justification

a. Incorrect. Plausible because the alarm response for 13 requires a "normal" boration to clear the alarm, also IV would be correct if cooldown was less than 540°F, and Vis correct.

Incorrect.

b. Incorrect. Plausible because II is correct, student may believe that any rod stuck out greater than or equal to 20 steps is correct, and VI is correct.

c. Incorrect. Plausible because student may believe that any rod stuck out greater than or equal to 20 steps is correct, also IV would be correct if cooldown was less than 540°F, and V is correct. .

d. Correct. II, V, and VI are all correct. See 3-ES-0.1; Reactor Trip Response, step 5 and 3-0NOP-CVCS-3; Emergency Boration.

Technical

References:

3-ES-0.1; Reactor Trip Response, 3-0NOP-CVCS-3; Emergency Boration, 13LP-ILO-ONPCVC 13LP-ILO-EOPEOO Proposed References to be provided: None Learning Objective: Objective 1.0 & Objective 12.0 respectively Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 5, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# 000059M202 Ability to determine and interpret the following as they apply to the Accidental Liquid Radwaste Release: The permit for liquid radioactive-waste release Importance: 2.9 3.9 Question: #23 In accordance with 3-SOP-WDS-014; Liquid Waste Releases, the Liquid Radioactive Waste Release Permit Form requires documenting the results of the Radiation Monitor Source Check.

While completing the waste release permit for the release of 31 Waste Monitor Tank the R-18 Channel 1 failed its source check.

What actions are required in accordance with the procedure and the ODCM to allow the release authorization of the 31 Waste Monitor Tank?

a. As long as the radioactive liquid contained in the 31 Waste Monitor Tank is verified to be

1O Curies, excluding tritium and dissolved or entrained gases, then the release can be authorized.

b. As long as the volume of the 31 Waste Monitor Tank is verified by tank level indication to be:::; 90% and the liquid waste discharge flow meter (FT-1064) is operable, then the release can be authorized.

c. As long as a satisfactory check-source test of R-18 Channel 2 has been completed, then the release can be authorized.

d. As long as two (2) independent samples have been taken, the release rate calculations independently verified, and the discharge valve alignment independently verified, then the release can be authorized.

Answer: d Explanation I Justification

a. Incorrect. Plausible because ODCM Spec D 3.1.4; Liquid Holdup Tanks states that outdoor liquid storage tanks (like 31 Monitor Tank) shall be limited to:::; 10 Curies. However, the procedure and ODCM Spec 3.3.1 require independent samples, independently verified release rate calculations, and independently verified discharge valve alignments in order to authorize a release with an inoperable rad monitor.

b. Incorrect. Plausible because one way that liquid effluent curie content is limited is by limiting the volume of the tank. Student may believe that knowing the tank volume, activity, and

then discharge flow rate is enough to authorize discharge. ODCM and procedure require independent samples, calculations, and lineup verification.

c. Incorrect. Plausible because the student may believe that R-18 Channel 2 is utilized for liquid waste discharges. See NOTE prior to step 4.1.6 of 3-SOP-WDS-014, which states:

"Do NOT perform a check-source check of R-18 Channel 2 as an inability to remove R-18 Channel 2 from service may occur after test. Channel 2 is NOT used."

d. Correct. See step 4.1.13 of 3-SOP-WDS-014 and ODCM Spec 3.3.1. The release may be authorized if independent samples, independently verified release rate calculations, and independently verified discharge valve alignments are completed.

Technical

References:

13LP-ILO-LWR001, 3-SOP-WDS-014 ODCM D 3.3.1, Radioactive Liquid Effluent Monitoring Instrumentation Proposed References to be provided: None Learning Objective: Objective 7 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41(b)7, 10, & 11

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# 0000762128 High Reactor Coolant Activity: Knowledge of the purpose and function of major system components and controls Importance: 4.1 4.1 Question: #24 Given the following:

• Unit 3 is at 100% power

• R63A and B (Gross Failed Fuel Detectors) are in Alert

• Chemistry has confirmed that high RCS activity exists

• Chemistry has also confirmed that RCS activity is less than TS 3.4.16 limits

• 3-AOP-HIACT-1; High RCS Activity has been entered Which of the following actions should be taken due to the rising RCS activity in accordance with 3-AOP-HIACT-1 and the reason for the actions?

a. Raise letdown flow to a maximum of 120 gpm to raise the rate of activity removal via the eves demineralizers.

b. Lower letdown flow to minimum (45 gpm) to minimize the amount of radioactive letdown that is flowing through the auxiliary building.

c. Conduct a Feed and Bleed of the RCS by diverting Letdown and initiating a continuous makeup to the VCT to rapidly reduce activity levels.

d. Place the Cation bed demineralizer in service and remove the mixed bed demineralizers in order to facilitate removal of fission products from the RCS.

Answer: a Explanation I Justification

a. Correct. 3-AOP-HIACT-1 ensures letdown is aligned through the eves demineralizers and then raises letdown flow up to a maximum of 120 gpm. The background document for 3-AOP-HIACT-1 states; "Step 4.5 increases letdown flow to increase the rate of activity removal via the eves demins."

b. Incorrect. Plausible because minimizing radiation levels in the auxiliary building is prudent from an ALARA perspective, however that will not help lower RCS activity levels.

c. Incorrect. Plausible because the student may believe that diverting the RCS will help lower RCS activity levels and permit waste processing activities.

d. Incorrect. Plausible because placing the Cation bed demineralizer in service is an action in the AOP, but from a system perspective the mixed bed demineralizers cannot be removed from service and still have flow through the Cation bed.

Technical

References:

13LP-ILO-RCSPSS, 3-AOP-HIACT-1 Proposed References to be provided: None Learning Objective: Objectives 4d, 5, and 6 Question Source: Modified Bank (IP3 Question # 24189)

Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# OOWE07EA 120 Ability to operate and I or monitor the following as they apply to the (Saturated Core Cooling): Operating behavior characteristics of the facility Importance: 3.2 3.7 Question: #25 A SATURATED CORE COOLING condition is identified during a review of the Critical Safety Function Status Trees following a small break loss of coolant accident. Operators have transitioned to 3-E-1; Loss of Reactor or Secondary Coolant. The following conditions are noted:

No Safety Injection pumps are available Both RHR pumps are running No RCPs are running Both motor driven AFW pumps are running RCS Pressure is 1100 psig CETs are 700°F and slowly rising RVLIS is 35% and slowly lowering Steam Generator pressure is 1050 psig Is a transition to a Functional Restoration procedure required? What indications are being monitored to verify adequate core cooling?

a. Yes, a transition to 3-FR-C.2; Response to Degraded Core Cooling is required. RVLIS level, Core Exit Thermocouples, and ECCS flow are monitored to verify adequate core cooling.

b. Yes, a transition to 3-FR-C.3; Response to Saturated Core Cooling is required. RVLIS level, Core Exit Thermocouples, RCP operating status, and RCS Subcooling are monitored to verify adequate core cooling.

c. No, a transition to a functional restoration procedure is not required. RVLIS level, RCP operating status, and AFW system status are monitored.

d. No, a transition to a functional restoration procedure is not required. RVLIS level, Core Exit Thermocouples, RCP operating status, and RCS Subcooling are monitored to verify adequate core cooling.

Answer: d Explanation I Justification

a. Incorrect. Plausible because if CETC temperatures were 715°F or RVLIS was 33%, then the transition would be to an Orange Path 3-FR-C.2. The second part of the answer is correct, but not all inclusive. Also ECCS flow will correct an inadequate core cooling situation, but is not a parameter monitored by CSFST 3-F-0.2; Core Cooling.

b. Incorrect. Plausible because the Critical Safety Function Status Tree will direct you to the Yellow Path based on the given plant conditions. However, Yellow conditions do not require immediate operator attention and therefore do not "require transition". The CRS/SM is allowed to decide whether or not to implement any Yellow condition FRP.

c. Incorrect. Plausible because Yellow conditions do not require immediate operator attention and therefore do not "require transition". No is correct, but AFW system status is not a parameter or indication monitored by CSFST 3-F-0.2; Core Cooling.

d. Correct. Yellow conditions do not require immediate operator attention and therefore do not "require transition". The CRS/SM is allowed to decide whether or not to implement any Yellow condition FRP. RVLIS level, Core Exit Thermocouples, RCP operating status, and RCS Subcooling are monitored by CSFST 3-F-0.2; Core Cooling.

Technical

References:

13LP-ILO-EOPFRC, 3-FR-C.3 OAP-012, EOP USERS GUIDE, Section 4.3 Proposed References to be provided: None Learning Objective: Objective 11.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# OOWE10K303 Knowledge of the reasons for the following responses as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS): Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.

Importance: 3.4 3.6 Question: #26 ES-0.4; "Natural Circulation Cooldown with Steam Void in Vessel (With RVLIS), checks RVLIS Full Range greater than 70% after the RCS coodown and depressurization have been started.

Why is maintaining RVLIS Full Range greater than 70% important and what actions are required should RVLIS Full Range lower to less than 70%?

a. Ensures the core is always covered with water to prevent the precipitation of boron in the upper region of the core that could reduce natural circulation flow. If RVLIS lowers to less than 70%, manually ACTUATE SI and GO To E-0, Reactor Trip or Safety Injection.

b. Ensures the reactor vessel head is maintained full of water to aid in the cooldown of the head and to prevent formation of additional steam in the upper vessel head area. If RVLIS lowers to less than 70%, the RCS depressurization is stopped while the cooldown is continued.

c. Ensures reactor vessel level is maintained above the top of the hot legs to prevent the formation of steam in the hot legs and the steam potentially reaching the top of the Steam Generator U-tubes causing a disruption in natural circulation. If RVLIS lowers to less than 70%, re-pressurize the RCS.

d. Ensures the reactor vessel level is maintained above the active fuel region of the core to ensure heat transfer via reflux boiling while controlling void growth in the reactor vessel head. If RVLIS lowers to less than 70%, transition to FR-1.3, Response to Voids in Reactor Vessel.

Answer: c Explanation I Justification

a. Incorrect. Plausible because the student may believe 70% RVLIS level is the level just above active fuel and boron precipitation is a concern following the injection phase during a

LOCA. Also plausible because SI initiation criteria exits in ES-0.3, but it is PRZR level cannot be maintained> 5%.

b. Incorrect. Plausible because the student may believe that 70% RVLIS levels indicate water in the reactor head region and that having level in that region may prevent additional steam formation. Finally, the second part of the answer is partially correct. The procedure would require that the depressurization be stopped while continuing with the cooldown, however re-pressurization to reduce the size of the void would also be required.

c. Correct. Ensures reactor vessel level is maintained above the top of the hot legs to prevent the formation of steam in the hot legs and the steam potentially reaching the top of the Steam Generator U-tubes causing a disruption in natural circulation. If RVLIS lowers to less than 70%, re-pressurize the RCS.

d. Incorrect. Plausible because if level does lower below the top of the hot legs, reflux boiling may be the only method of heat transfer available. The student may believe that ES-0.3 allows a rapid cooldown and depressurization, so therefore it will allow significant void growth to ensure an expedited cooldown and depressurization. Also the student may believe that the functional recovery procedure for voids is applicable. It actually states that if natural circulation is occurring and void growth is expected, do not enter.

Technical

References:

13LP-ILO-EOPEOO, 3-ES-0.3 Proposed References to be provided: None Learning Objective: Objective 12.0 Question Source: Modified Bank Question# 24605 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 1 Group# 2 KIA# OOWE14K201 Knowledge of the interrelations between the (High Containment Pressure) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Importance: 3.4 3.7 Question: #27 Select the statement below which describes the automatic containment spray actuation signal.

a. Containment Pressure of 22 psig sensed on 2 of 3 containment pressure detectors PT-948A, B, C.

b. Containment Pressure of 22 psig sensed by 2 of 3 containment pressure detectors from PT-948A, B, C AND 2 of 3 containment pressure detectors PT-949A, B, C.

c. Containment Pressure of 22 psig sensed by 2 of 3 containment pressure detectors from PT-948A, B, C OR 2 of 3 containment pressure detectors PT-949A, B, C.

d. Containment Pressure of 22 psig sensed on any 4 containment pressure detectors PT-948A, B, C and PT-949A, B, C.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe the coincidence for containment spray actuation is only 2/3 on one channel. Pressure setpoint is correct.

b. Correct. The signal is a coincident 2/3 from two separate channels. See System Description 10.2, page 5.

c. Incorrect. Plausible because the student may believe that for redundancy there are two sets of instruments, but only one channel required to initiate spray.

d. Incorrect. Plausible because the student may believe that for redundancy there are multiple instruments, but only any four are required to initiate spray.

Technical

References:

13LP-ILO-CS001 Proposed References to be provided: None Learning Objective: Objective E-5 Question Source: Bank Question # 23848 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 003000A103 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: RCP motor stator winding temperatures Importance: 2.6 2.6 Question: #28 Given the following conditions:

• The Reactor is in Mode 5

• RCS pressure is 350 psig

• RCS temperature is 190°F

• 34 RCP is in service

• RHR is in service Which of the following conditions if associated with 34 RCP would require the immediate tripping of the pump?

a. Shaft vibration reading of 20 mils and stable.

b. Motor winding (stator) temp. on 4-point recorder> 250°F for> 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

c. Shaft vibration reading of 15 mils and rising at 0.2 mils/hr.

d. #1 seal return flow reading 5.8 gpm.

Answer: b Explanation I Justification

a. Incorrect. Plausible because 3-AOP-RCP-1, STEP 4.1 requires stopping affected RCPs if Shaft Vibration is > 20 mils. In this case, vibration is stable at 20 mils.

b. Correct. 3-AOP-RCP-1, STEP 4.1 requires stopping affected RCPs if Motor winding (stator) temp on 4-point recorder> 250°F for> 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

c. Incorrect. Plausible because 3-AOP-RCP-1, STEP 4.1 requires stopping affected RCPs if Shaft vibration > 15 mils and increasing > 1.0 mils/hr. Also the IAAT for frame vibration uses increasing > 0.2 mils/hr.

a. Incorrect. Plausible because 3-AOP-RCP-1, STEP 4.1 requires stopping affected RCPs if

1. 1 seal return flow> 6 gpm (5.99 gpm digital).

Technical

References:

13LP-ILO-AOPRCP Proposed References to be provided: None Learning Objective: Objective 3.0 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41(b)5&10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 004000K604 Knowledge of the effect of a loss or malfunction on the following eves components: Pumps Importance: 2.8 3.1 Question: #29 Given the following conditions:

• Unit is at 50% Power

• All Control Systems are in Automatic Which of the following malfunctions will result in an increased actual NPSH available for the in service Charging Pump?

a. TCV-130, Non-Regen. Hx. CCW Outlet Flow Valve fails open.

b. Controlling Pressurizer Level Channel fails to 25% indicated level.

c. LT-112, VCT Level Channel fails full scale high.

d. TE-130, Non-Regen. Hx. Outlet temperature device fails full scale low.

Answer: a Explanation I Justification

a. Correct. If the CCW control valve to the NRHX fails open, letdown temperature will lower which will result in a lower VCT temperature and therefore a higher actual NPSH available to the running Charging Pump.

b. Incorrect. Plausible because the student is required to understand PZR Level Control and how that will impact the charging and letdown system. A failed PZR Level (controlling channel) will result in a level deviation causing the level controller to raise charging pump speed to in turn raise charging flow. However, a rise in charging pump speed will result in a decrease in actual NPSH available to the running charging pump. Knowledge of NPSH relationships to temp, flow, pressure, level, and speed also required.

c. Incorrect. Plausible because the student is required to understand VCT level control and what will happen to actual level without operator action. The student may also believe that either makeup or swap over to the RWST occurs automatically. However, with the VCT level channel failed high, the LCV-112A will direct all letdown flow to the CVCS HUT and this will result in a lowering of VCT level and therefore a lowering of actual NPSH available to the running Charging Pump. With LT-112 failed high, no auto makeup or swapping to the RWST will occur.

d. Incorrect. Plausible because the student is required to understand how TE-130 is used for letdown temperature control and then how letdown temperature affects charging pump NPSH. If the TE-130 fails low, it sends a low temperature signal to the HTC-130 which controls the TCV-130 valve (CCW control valve to the NRHX) in the CCW system causing the TCV-130 to close. Closing the TCV-130 results in rising Letdown temperature, which then results in rising VCT temperature and a decrease in actual NPSH available to the running charging pump.

Technical

References:

13LP-ILO-CVC001 3-ARP-009 3-AOP-CVCS-1 System Description 3.0 Proposed References to be provided: None Learning Objective: Objective 6.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 005000K505 Knowledge of the operational implications of the following concepts as they apply to the RHRS: Plant response during "solid plant" pressure change due to the relative incompressibility of water.

Importance: 2.7 3.1 Question: #30 Given the following plant conditions:

• Mode 5

• 31 RHR Pump is in shutdown cooling mode

• Low Pressure purification is in service

• The pressurizer is water solid

• RCS temperature is 166°F

• Letdown Pressure Control Valve PCV-135 is in automatic

• RCS pressure is being maintained at 275 psig The 31 RHR Heat Exchanger Outlet Flow Control Valve (HCV-638) has just stroked from an initial position of 25% open to the full open position due to a circuit failure. How does the Letdown system initially respond to this valve failure?

a. PCV-135 OPENS, Letdown flow through HCV-133 will LOWER

b. PCV-135 CLOSES. Letdown flow through HCV-133 will LOWER

c. PCV-135 OPENS, Letdown flow through HCV-133 will RISE

d. PCV-135 CLOSES, Letdown flow through HCV-133 will RISE Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe that the higher RHR flow rate will cause an increase in RCS pressure during solid plant operations. A lower letdown flow rate is plausible if more RHR flow is directed to the RCS loops instead of the low pressure purification system.

b. Correct. RHR I RCS temperature will initially lower due to greater flow through the RHR Heat Exchanger. In a solid plant condition, this will result in a reduction in RCS pressure.

PCV-135 will respond to the reduction in pressure by closing down in automatic. The

combination of this closing of PCV-135 and the rise in flow to the RCS loops will result in reduced letdown flow through the HCV-133.

c. Incorrect. Plausible because the student may believe that the higher RHR flow rate will cause an increase in RCS pressure during solid plant operations. Also plausible because if the PCV-135 opens, letdown flow through the HCV-133 may also rise.

d. Incorrect. Plausible because PCV-135 will respond to the reduction in pressure by closing down in automatic. Student may be confused on the subsequent flow rate through HCV-133, he may believe that the higher flow through the RHR HX will also result in higher letdown flow rate.

NOTES: Check response? Find out typical values? HCV-638 has a 60% open mechanical stop limiting flow to 3000gpm. Intent of question is to lower temperature I pressure.

Technical

References:

13LP-ILO-RHR001 3-SOP-RHR-001 3-SOP-CVCS-O 12 Proposed References to be provided: None Learning Objective: Objective 3.0 Question Source: Modified Bank - Seabrook 2013 NRC Exam - 030 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 5 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 0060002123 Emergency Core Cooling: Ability to perform specific system and integrated plant procedures during all modes of operation Importance: 4.3 4.4 Question: #31 Which of the following methods of limiting High Head Safety Injection (HHSI) into the RCS to ensure cold RCS over pressurization protection is the preferred method in accordance with integrated operating procedure 3-POP-3.3; Plant Cooldown - Hot to Cold Shutdown?

a. Place 31, 32, and 33 Safety Injection Pump Control Switches in PULL OUT and close specified SI MOVs from the control room.

b. RACK OUT Breakers associated with 31, 32, and 33 Safety Injection Pumps and close and remove power from specified SI MOVs.

c. RACK OUT Breakers associated with 31, 32, and 33 Safety Injection Pumps and Place 31, 32, and 33 Safety Injection Pump Control Switches in PULL OUT.

d. Close the specified SI MOVs associated with the Boron Injection Tank Flowpath and maintain one Safety Injection Pump available to be manually realigned to the RCS Cold Legs.

Answer: a Explanation I Justification

a. Correct. Attachment 3, step 1.0, of 3-POP-3.3 states that placing the Control Switches in Pull Out and closing listed SI MOVs is the preferred method. The pump breakers are not racked out so that two safety injection pumps shall be maintained available for manual realignment to injection in Mode 4 to support a Hot Shutdown LOCA with Accumulators Isolated event.

b. Incorrect. Plausible because Method C (not the preferred method) has the Breakers racked out, but no valves are manipulated.

c. Incorrect. Plausible because this is exactly what is stated as Method C (not the preferred method).

a. Incorrect. Plausible because Attachment 3 footnote (1) states, Two safety injection pumps shall be maintained available for manual realignment to injection in Mode 4 to support ONOP-RCS-8; Hot Shutdown LOCA with Accumulators Isolated. Tech Spec LCO 3.4.12, Note 2 also states; "One HHSI pump may be made capable of injecting into the RCS as needed to support emergency boration or to respond to a loss of RHR cooling." But again, not the preferred method.

Technical

References:

13LP-ILO-POP002, 3-POP-3.3 13LP-ILO-ONPRCS, ONOP-RCS-8 Proposed References to be provided: None Learning Objective: Objectives 4.0 and 2.0 respectively Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 006000K506 Knowledge of the operational implications of the following concepts as they apply to ECCS: Relationship between ECCS flow and RCS pressure Importance: 3.5 3.9 Question: #32 A small break LOCA is in progress with the following conditions:

• All ECCS Pumps are operating except 31 SI Pump which failed to start.

• RCS Pressure is 1200 psig and stable.

• Pressurizer Level is 5% and stable.

• RCS Average Temperature is slowly lowering about 5°F/hour.

• No Charging Pumps are operating.

For this condition what is the approximate break flow?

a. 100 - 200 gpm

b. 300 - 500 gpm

c. 700 - 900 gpm

d. 1100 - 1300 gpm Answer: c Explanation I Justification

a. Incorrect. Plausible because student needs to know that design pressure and flow of the SI Pumps is 400 gpm @ 1200 psig. He then needs to know that only two pumps are operating in parallel, with something close to almost 2x flow depending on system resistance. For the leak to be in this range, then PZR Level would be rising.

b. Incorrect. Plausible because the student may confuse parallel and series pump operations.

c. Correct. Student needs to know that design pressure and flow of the SI Pumps is 400 gpm

@ 1200 psig. He then needs to know that only two pumps are operating in parallel, with something close to almost 2x flow depending on system resistance.

a. Incorrect. Plausible because this would be the approximate flow range for all three SI pumps operating.

Technical

References:

13LP-ILO-SIS001 Proposed References to be provided: None

Learning Objective: Objective 3.0 Question Source: Bank Question # 23949 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)8

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 007000A102 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining quench tank pressure Importance: 2.7 2.9 Question: #33 Given the following:

• The plant is operating at 100% power.

• The following indications are reported by the Reactor Operator:

o The PRESSURIZER RELIEF TANK LIQUID HIGH TEMPERATURE Annunciator has alarmed and PRT temperature is currently 135°F and slowly rising.

o The PRESSURIZER RELIEF TANK HIGH PRESSURE Annunciator has alarmed and PRT pressure is currently 7 psig and slowly rising.

o The field operator has reported that the PRT N2 pressure regulator is set at less than 5 psig.

Responding to the Alarm Response Procedures for both alarms, the Reactor Operator is taking action in accordance with 3-SOP-RCS-007, Pressurizer Relief Tank Operations. Which of the following procedural actions will restore both alarmed conditions to their normal operating parameters and at what pressure are the PRT Rupture Disks designed to rupture to containment?

a. Venting the PRT directly to Containment atmosphere while filling the PRT to 90% will reduce both temperature and pressure. The Rupture Disk Design Pressure is 85 psig.

b. Spraying Primary Water into the PRT will reduce both temperature and pressure. The Rupture Disk Design Pressure is 100 psig.

c. Spraying Primary Water into the PRT will reduce both temperature and pressure. The Rupture Disk Design Pressure is 85 psig.

d. Venting the PRT directly to Containment atmosphere while filling the PRT to 90% will reduce both temperature and pressure. The Rupture Disk Design Pressure is 100 psig.

Answer: b Explanation I Justification

a. Incorrect. Plausible because venting will lower PRT pressure and adding water volume to the PRT will lower temperature. However, procedural guidance does not exist in 3-SOP-RCS-007 to vent the PRT directly to the Containment atmosphere. Filling and draining of the PRT would also be done within normal operating limits of 67 - 77%, 90% is a value used for purging the PRT. Additionally, Rupture Disk rupture pressure is not 85 psig.

b. Correct. The procedure states that both temperature and pressure control can be achieved by using the FILL and DRAIN PRT instructions under section 4.3, PRT Level Control. This section uses Primary Water Spray to control both temperature and pressure. Also plausible because 100 psig is the correct rupture disk rupture pressure

c. Incorrect. Plausible because the procedure states that both temperature and pressure control can be achieved by using the FILL and DRAIN PRT instructions under section 4.3, PRT Level Control. This section uses Primary Water Spray to control both temperature and pressure. Additionally, Rupture Disk rupture pressure is not 85 psig.

a. Incorrect. Plausible because venting will lower PRT pressure and adding water volume to the PRT will lower temperature. However, procedural guidance does not exist in 3-SOP-RCS-007 to vent the PRT directly to the Containment atmosphere. Filling and draining of the PRT would also be done within normal operating limits of 67 - 77%, 90% is a value used for purging the PRT. Additionally, Rupture Disk rupture pressure is 100 psig.

Technical

References:

13LP-ILO-RCSPZR Proposed References to be provided: None Learning Objective: Objective E-8 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 5, 7, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 008000K401 Knowledge of the CCWS design feature(s) and/or interlock(s) which provide for the following: Automatic start of standby pump Importance: 3.1 3.3 Question: #34 Given the following:

• The Unit is at 100% Power.

• CCW Pumps 32 and 33 are running. CCW Pump 31 is in Auto.

• 6.9 KV Bus 6 normal feeder inadvertently opens.

Which ONE of the following describes the AUTOMATIC operation of CCW Pumps for this event?

a. CCW Pump 31 starts when Bus 6A is de-energized and CCW Header Pressure lowers to 105 psig. CCW Pump 33 restarts 19 seconds after Bus 6A is re-energized.

b. CCW Pump 31 starts when Bus 6A is de-energized and CCW Header Pressure lowers to 100 psig. CCW Pump 33 restarts 28 seconds after Bus 6A is re-energized.

c. CCW Pump 31 starts when Bus 6A is de-energized and CCW Header Pressure lowers to 100 psig. CCW Pump 33 restarts 34 seconds after Bus 6A is re-energized.

d. CCW Pump 31 starts when Bus 6A is de-energized and CCW Header Pressure lowers to 105 psig. CCW Pump 33 does NOT automatically restart.

Answer: c Explanation I Justification

a. Incorrect. Plausible because student may believe that the standby CCW pump starts on low supply header pressure at 105 psig. Also plausible to believe that CCW loads at 19 seconds which is the loading time for service water pumps.

b. Incorrect. Plausible because the standby pump does start at 100 psig and the CCW pump on Bus 5A restarts 28 seconds after the bus is re-energized on a single bus loss of voltage.

Buses 2A and 6A load CCW Pumps at 34 seconds into the blackout loading sequence.

c. Correct. The standby CCW Pump starts on low supply header pressure (discharge) at 100 psig. The single bus loss of voltage sequence for Bus 6A starts the 33 CCW Pump at 34 seconds into the loading sequence.

d. Incorrect. Plausible because student may believe that the standby CCW pump starts on low supply header pressure at 105 psig. Also plausible for student to believe that CCW Pumps

do not load on a "Blackout" or single bus loss of voltage. CCW Pumps are not started for a "Blackout and SI" signal, but are started on a "Blackout" or single bus loss of voltage.

Technical

References:

13LP-I LO-CCWOO 1 System Description 4.1; Component Cooling Proposed References to be provided: None Learning Objective: Objective 4 Question Source: Modified Bank - Indian Point 3 - Question # 24303 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 7, 8 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 010000K602 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PZR Importance: 3.2 3.5 Question: #35 The plant is operating at 100% power with the following conditions:

• Pressurizer Pressure Control is in AUTO

• Pressurizer Pressure Defeat Switch is in position "Defeat 3&4"

• Pressurizer Pressure Channel PT-455 fails full scale low

• NO operator action is taken for the event Which of the following describes the Pressurizer Pressure Control System response?

a. All PZR Heaters energize, Spray Valves PCV 455A and 4558 go full closed, and PORV RC-PCV-456 opens.

b. All PZR Heaters energize, Spray Valves PCV 455A and 4558 go full closed, and both PORVs RC-PCV-455C and 456 open.

c. All PZR Heaters energize, Spray Valves PCV 455A and 4558 then subsequently open to offset the Pressure rise, both PORVs RC-PCV-455C and 456 remain closed.

d. All PZR Heaters energize, Spray Valves PCV 455A and 4558 then subsequently open to offset the Pressure rise, PORV RC-PCV-456 opens.

Answer: a Explanation I Justification

a. Correct. PZR Control configuration is such that Channel I (PT-455) is the controlling channel and Channel II (PT-456) is the alarm channel. All heaters energize due to the failed low controlling channel, causing actual pressure to rise. Both spray valves close, and PORV RC-PCV-455C is inhibited from auto opening. If no operator action is taken, actual pressurizer pressure rises slowly until PZR PORV RC-PCV-456 opens (Ch II & Ch Ill).

b. Incorrect. Plausible because student may believe that only one channel is required to allow the PORVs to open. There may also be some misunderstanding of the configuration of the PZR PCS with the defeat switch in the "Defeat 3&4" position.

c. Incorrect. Plausible because the student may believe that the PZR Spray valves will respond to the pressure rise forgetting that the controller is selected to the failed channel.

d. Incorrect. Plausible because may believe that the PZR Spray valves will respond to the pressure rise forgetting that the controller is selected to the failed channel. Plausible because PORV RC-PCV-456 does open.

Technical

References:

13LP-ILO-ICPZRC 13LP-ILO-AOPINT System Description 1.4; Pressurizer & Pressurizer Relief Tank Proposed References to be provided: None Learning Objective: Objective 5.0 & 3.0 respectively Question Source: New Question Cognitive Level: Comprehension I Analysis 10CFR Part 55 Content: 55.41 (b) 5, 7, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 012000K502 Knowledge of the operational implications of the following concepts as they apply to the RPS: Power density Importance: 3.1 3.3 Question: #36 Which of the following Reactor Protection System Trips ensures that the allowable heat generation rate (kW/ft) of the fuel is not exceeded?

a. Overtemperature L1 T (OTL1 T)

b. Power Range Neutron Flux - High

c. Steam Generator Water Level - Low Low

d. Overpower L1 T (OPL1 T)

Answer: d Explanation I Justification

a. Incorrect. Plausible because OTL1T ensures the design limit DNBR is met and the inputs to the setpoint include coolant temperature and axial power distribution.

b. Incorrect. Plausible because this trip ensures protection against a positive reactivity excursion leading to DNB.

c. Incorrect. Plausible because this trip protects against a loss of heat sink and actuates the AFW System prior to uncovering the SG tubes.

d. Correct. The OPL1 T trip ensures that the allowable heat generation rate (kW/ft) of the fuel is not exceeded. See Tech Spec 3.3.1 bases.

Technical

References:

13LP-ILO-ICRXP Proposed References to be provided: None Learning Objective: Objective E-1 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 0120002145 Ability to identify and interpret diverse indications to validate the response of another indication Importance: 4.3 4.3 Question: #37 Which of the following malfunctions will cause the Overtemperature L1 T Channel Trip or Rod Stop alarm to annunciate?

a. PR Channel N41 failure high

b. Loop 32 Thot microprocessor fails low

c. PT-456 fails high

d. Loop 31 Tcold RTD fails low Answer: d Explanation I Justification

a. Incorrect. Plausible because NIS does provide an input to OTL1 T; however input is from delta flux. If the upper detector for N41 failed high, the alarm would annunciate.

b. Incorrect. Plausible because the candidate must determine how the failure will affect actual L1 T and how the calculated setpoint will be affected.

c. Incorrect. Plausible because candidate must recall that Channel 2 (PT-456) cannot be the controlling channel. If a controlling channel failed high, actual pressure would decrease and the alarm would annunciate when the first OTL1 T channel setpoint was reached.

d. Correct. The failure causes the indicated L1 T to fail high. The alarm will annunciate when the calculated setpoint is less than the indicated L1 T. Setpoint should also decrease based on a higher Tavg.

Technical

References:

13LP-ILO-ICRXP System Description 28.0 Proposed References to be provided: None Learning Objective: Objective 1.0 Question Source: Bank (IP3 - Question # 25208)

Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 013000K301 Knowledge of the effect that a loss or malfunctions of the ESFAS will have on the following: Fuel Importance: 4.4 4.7 Question: #38 Which of the following choices contains ONLY items that are described in 10 CFR 50.46 as Acceptance Criteria for Emergency Core Cooling Systems for light-water nuclear power reactors?

I. The calculated total amount of hydrogen generated shall not exceed .17 times the hypothetical amount that would be generated if all the fuel cladding were to react.

II. The calculated maximum fuel element cladding temperature shall not exceed 2000°F.

Ill. The calculated total oxidation of the cladding shall not exceed 0.17 times the total cladding thickness before oxidation.

IV. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

V. The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

VI. The calculated total oxidation of the cladding shall not exceed 0.1 O times the total cladding thickness before oxidation.

VII. The zirc-water reaction postulated to occur at 1800°F will not become self-sustaining under any circumstance.

VIII. The calculated total amount of hydrogen generated shall not exceed 0.01 times the hypothetical amount that would be generated if all the fuel cladding were to react.

a. I, II, VI, VII

b. II, Ill, IV, VIII

c. Ill, IV, V, VIII

d. I, V, VI, VII Answer: c Explanation I Justification

a. Incorrect. Plausible because the answer includes hydrogen production, cladding reduction, and cladding temperature numbers. The numbers are incorrect for their individual acceptance criteria, but include some actual numbers from 50.46.

b. Incorrect. Plausible because Ill, IV, and VIII are correct. Student may believe the peak cladding temperature to be 2000°F.

c. Correct. See 10 CFR 50.46. Cladding - 17%, Hydrogen - 1%, Coolable Geometry, 2200°F.

d. Incorrect. Plausible because the answer includes hydrogen production, cladding reduction, and cladding temperature numbers. The numbers are incorrect for their individual acceptance criteria, but include some actual numbers from 50.46.

Technical

References:

13LP-ILO-ESS001 Proposed References to be provided: None Learning Objective: Objective 1.0 Question Source: Modified Bank - Salem 2004 NRC Exam Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 022000A405 Containment Cooling System - Ability to manually operate and/or monitor in the control room: Containment readings of temperature, pressure, and humidity system Importance: 3.8 3.8 Question: #39 Given the following plant conditions:

• Unit is at 100% Power

• RWST Temperature =95°F

• Containment Temperature = 125°F In accordance with Technical Specifications, Containment Pressure shall be maintained below what upper operating limit and why?

a. Containment pressure shall be maintained ::; +1.5 psig to ensure that in the event of a OBA, the resultant containment accident pressure will remain below the containment design pressure.

b. Containment pressure shall be maintained ::; +2.5 psig to maximize the effectiveness of the ECCS during the core reflood phase of a LOCA and ensure motor heating concerns are addressed.

c. Containment pressure shall be maintained ::; +2.5 psig to ensure that in the event of a OBA, the resultant containment accident pressure will remain below the containment design pressure.

d. Containment pressure shall be maintained~ +1.5 psig to maximize the effectiveness of the ECCS during the core reflood phase of a LOCA and ensure motor heating concerns are addressed.

Answer: c Explanation I Justification

a. Incorrect. Plausible because if RWST Temperature was greater than 95°F and Containment Temperature was greater than 125°F, then this would be the correct answer.

b. Incorrect. Plausible because pressure shall be maintained ::; +2.5 psig in accordance with TS LCO 3.6.4, but the bases is incorrect. Higher Containment Pressures (not limiting them) will cause higher containment backpressure and therefore increase the effectiveness of the ECCS during the reflood phase. Also motor heating concerns are associated with ensuring pressure is greater than the lower limit of -2.0 psig. ECCS effectiveness and motor heating

are discussed in the bases, but not with respect to limiting containment pressure to less than or equal to the upper operating limit.

c. Correct. If RWST temperature is:::; 95°F and Containment temperature is:::; 125°F, Containment pressure shall be ;::: -2.0 psig and :::; +2.5 psig. See TS 3.6.4 LCO and bases.

d. Incorrect. Plausible because if RWST Temperature was greater than 95°F and Containment Temperature was greater than 125°F, the first part of this distractor is correct. Also plausible because ECCS effectiveness and motor heating are discussed in the bases, but not with respect to limiting containment pressure to less than or equal to the upper operating limit.

Technical

References:

Technical Specification LCO 3.6.4 & Bases 13LP-ILO-VCVCB Proposed References to be provided: None Learning Objective: Objectives 7.0 & 9.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 000022K201 Knowledge of power supplies to the following: Containment cooling fans Importance: 3.0 3.1 Question: #40 Fan cooler Unit 31 is tagged out for maintenance. During a Safety Injection with a Loss of Off-Site Power, Emergency Diesel Generator 32 fails to start. Which of the following indicates the containment cooling equipment available?

a. 32, 33, and 34 Fan Cooler Units 31 Containment Spray Pump Minimum safeguards equipment for containment cooling is satisfied.

b. 33 and 35 Fan Cooler Units 31 Containment Spray Pump Minimum safeguards equipment for containment cooling is NOT satisfied.

c. 33 and 35 Fan Cooler Units 31 and 32 Containment Spray Pumps Minimum safeguards equipment for containment cooling is satisfied.

d. 32, 33, and 34 Fan Cooler Units 32 Containment Spray Pump Minimum safeguards equipment for containment cooling is NOT satisfied.

Answer: a Explanation I Justification

a. Correct. #31 FCU is tagged in the stem. 32 EDG feeds bus 6A which supplies 32 Containment Spray Pump and #35 FCU. All other equipment related to containment cooling remains available. Minimum design requirements are also met, 1 Containment Spray Pumps and 3 CFUs.

b. Incorrect. Plausible because the student may believe that #32 and #34 FCU are supplied by 32 EOG and the combination of one spray pump and only two FCUs does not satisfy minimum design requirements.

c. Incorrect. Plausible because the student may believe that 32 EDG does not supply a containment spray pump and two containment spray pumps does satisfy minimum design requirements.

d. Incorrect. Plausible because the available FCUs are correct, but the containment spray pump available is incorrect. The student may also believe more than three FCUs are required for minimum design requirements to be met.

Technical

References:

13LP-ILO-VCCARC 13LP-ILO-CS001 Proposed References to be provided: None Learning Objective: Objectives 3.0, 4.0 and E-4 Question Source: Modified Bank Question # 23864 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)?

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 000026K202 Knowledge of bus power supplies to the following: MOVs Importance: 2.7 2.9 Question: #41 The unit has experienced a large break LOCA and a Loss of Off-Site Power. Emergency Diesel Generator 33 has tripped. Containment pressure has reached the HI-HI setpoint and a Containment Spray actuation signal has occurred. What is the present status of the Containment Spray System?

a. Both Containment Spray Pumps are running and discharging to the spray header via their respective open pump discharge MOVs (866A and 866B).

b. Both Containment Spray Pumps are running, however only 31 Containment Spray Pump is discharging through its open pump discharge MOV (866A).

c. Only 31 Containment Spray Pump is running and discharging through its open pump discharge MOV (866A).

d. Only 32 Containment Spray Pump is running and discharging through its open pump discharge MOV (866B).

Answer: d Explanation I Justification

a. Incorrect. Plausible because the student may believe that the 33 EOG does not supply either of the Containment Spray Pumps and their associated MOVs.

b. Incorrect. Plausible because the student may believe that the 33 EOG does not supply either of the Containment Spray Pumps, however they may think that the 33 EOG feeds one of the MOVs through MCC-36A or MCC-36B.

c. Incorrect. Plausible because the student may believe that 32 Containment Spray Pump and its respective discharge MOV are supplied by the 33 EOG.

d. Correct. 31 Containment Spray Pumps and its respective discharge MOV (MCC-36A) are powered from the 5A Bus and the 33 EOG.

Technical

References:

13LP-ILO-CS001 Proposed References to be provided: None Learning Objective: Objective E-4 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 000026A405 Ability to manually operate and/or monitor in the control room: Containment spray reset switches Importance: 3.5 3.5 Question: #42 A large break LOCA has occurred and containment pressure has risen to the automatic containment spray setpoint. Both Containment Spray Pumps start and 4 Fan Cooler Units start.

The fifth CFU cannot be started. If containment pressure remains above 20 psig; which of the following describes when the Containment Spray Pumps are stopped?

a. Both Containment Spray Pumps are stopped when Recirc Switch 1 is taken to ON.

b. 31 Containment Spray Pump is stopped immediately after the RWST reaches 11.5 feet and 32 Containment Spray Pump is stopped when the RWST reaches 1.5 feet.

c. 32 Containment Spray Pump is stopped when Recirc Switch 1 is taken to ON and 31 Containment Spray Pump is stopped after the RWST reaches 1.5 feet.

d. 31 Containment Spray Pump is stopped when Recirc Switch 1 is taken to ON and 32 Containment Spray Pump is stopped after the RWST reaches 1.5 feet.

Answer: c Explanation I Justification

a. Incorrect. Plausible because the student may believe that recirculation switch 1 stops both Containment Spray Pumps. Also Attachment 6 of ES-1.3 uses containment pressure of 16 psig or 5 FCU running to stop both Containment Spray Pumps after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Stem includes information on containment pressure and number of FCU running.

b. Incorrect. Plausible because the entry condition for ES-1.3, Transfer to Cold Leg Recirculation is RWST level < 11.5 feet. RWST level of 1.5 feet is also the level at which operators stop the second containment spray pump when transferring to recirculation mode.

c. Correct. See ES-1.3, Transfer to Cold Leg Recirculation step 4.0 and Attachment 6, steps 1.0 & 2.0. Also see System Description 10.2, pages 4-6.

d. Incorrect. Plausible because the student may believe the order of stopping pumps is 31 first, followed by 32.

Technical

References:

13LP-ILO-EOPE10, 3-ES-1.3 System Description 10.2 Proposed References to be provided: None Learning Objective: Objective 22.0

Question Source: Bank IP3 - Question # 2435 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 039000K406 Knowledge of MRSS (Main and Reheat Steam System) design feature(s) and/or interlock(s) which provide the following:

Prevent reverse steam flow on a steam line break Importance: 3.3 3.6 Question: #43 The Main Steam Non-Return Checks Valves are:

a. Located upstream of the MS IVs and provide steam line isolation in the event of a downstream rupture. They are designed to limit the blowdown rate of the steam generators in the event of a steam line break.

b. Located downstream of the MS IVs and provide steam line isolation in the event of an upstream rupture. They are designed to preclude the blowdown of more than one steam generator inside containment.

c. Located upstream of the MS IVs and provide steam line isolation in the event of an upstream rupture. They are designed to preclude the blowdown of more than one steam generator inside containment.

d. Located downstream of the MSIVs and provide isolation in the event of a downstream rupture. They are designed to limit the blowdown rate of the steam generators in the event of a steam line break.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe that they are installed upstream of the MSIVs and may believe they function similar to the MSIVs in that they open into the steam flow and steam flow assists in their closure. They may also believe they function like an "excess flow" check valve. The flow restrictor is designed to limit the blowdown rate in the event of a downstream rupture. The valve will limit the total blowdown, but not the rate.

b. Correct. They are located downstream of the MS IVs and provide steam line isolation in the event of an upstream rupture. They are designed to preclude the blowdown of more than one steam generator inside containment.

c. Incorrect. Plausible because everything in the statement is correct except the physical location of the check valve respective to the MS IVs.

d. Incorrect. Plausible because the location respective to the MSIVs is correct. Also plausible because the student may believe they function similar to the MSIVs in that they open into the steam flow and steam flow assists in their closure.

Technical

References:

13LP-ILO-MSS001 FSAR Section 14.2.5 System Description 18.0 Proposed References to be provided: None Learning Objective: Objective 2.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)?

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 059000A206 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW (Main Feedwater System); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of steam flow to MFW system Importance: 2.7 2.9 Question: #44 Given the following:

• The Plant is at 100% Power

• All Control Systems are in Automatic

• All Steam Flow and Feed Flow Channel Transfer Switches are selected to Channel A

• 31 SG Steam Flow Compensating Pressure Transmitter PT-419A fails to zero.

First predict the plant response to the failed pressure transmitter with NO operator action, then secondly describe what procedural operator actions would correct or mitigate the consequences of the failed pressure transmitter.

a. 31 SG Steam Flow Channel I fails low causing a steam flow I feed flow mismatch signal which then closes the Feed Regulating Valve reducing feed flow. Actual Steam Generator Level lowers causing a level program error, but without operator action, SG level lowers until the reactor trips due to low SG level. Mitigating actions would be taken in accordance with 3-AOP-INST-1. Operator would take manual control of the FRV and select the SG Transfer Switch to Steam Flow Channel II.

b. 31 SG Steam Flow Channel I fails low causing a steam flow I feed flow mismatch signal which then closes the Feed Regulating Valve reducing feed flow. Actual Steam Generator Level lowers causing a level program error with a 3 times gain which then overcomes the steam flow I feed flow mismatch signal. Because of the level program error gain, the Feed Regulating Valve opens restoring Steam Generator level to approximately 5% below program. Mitigating actions would be taken in accordance with 3-AOP-FW-1. Operator would take manual control of the FRV and select the SG Transfer Switch to Steam Flow Channel II.

c. 31 SG Steam Flow Channel I fails high causing a steam flow I feed flow mismatch signal which then opens the Feed Regulating Valve raising feed flow. Actual Steam Generator

Level rises causing a level program error, but without operator action, SG level rises until a turbine trip and feedwater isolation occurs on high SG level. With reactor power >

10%, the turbine trip causes a reactor trip. Mitigating actions would be taken in accordance with 3-AOP-INST-1. Operator would take manual control of the FRV and select the SG Transfer Switch to Steam Flow Channel II.

d. 31 SG Steam Flow Channel I fails high causing a steam flow I feed flow mismatch signal which then opens the Feed Regulating Valve raising feed flow. Actual Steam Generator Level rises causing a level program error with a 3 times gain which then overcomes the steam flow I feed flow mismatch signal. Because of the level program error gain, the Feed Regulating Valve closes restoring Steam Generator level to approximately 5%

below program. Mitigating actions would be taken in accordance with 3-AOP-FW-1.

Operator would take manual control of the FRV and select the SG Transfer Switch to Steam Flow Channel II.

Answer: a Explanation I Justification

a. Correct. See SGWLC Lesson Plan and System Description. 3-AOP-INST-1 provides mitigating actions.

b. Incorrect. Plausible because the steam flow channel does fail low and the level channel error does have a 3X gain. However the negative signal from the failed low steam flow will actually still cause the overall signal to remain negative. Mitigating actions are correct, but wrong procedure.

c. Incorrect. Plausible because the student may believe if the compensating pressure channel fails to zero, the steam flow indication will fail high. Mitigating actions and the procedure are correct.

d. Incorrect. Plausible because the student may believe if the compensating pressure channel fails to zero, the steam flow indication will fail high. Mitigating actions are correct, but the procedure is incorrect.

Technical

References:

13LP-ILO-ICSGL System Description 21.1 3-AOP-INST-1 Proposed References to be provided: None Learning Objective: Objective 2.0 & 5.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 061000K301 Knowledge of the effect that a loss or malfunction of the AFW will have on the following: RCS Importance: 4.4 4.6 Question: #45 Given the following plant conditions:

The Unit is at 100% Power All Control Systems are in Automatic All Systems are in their normal 100% Power configuration Control Rods are at their fully withdrawn position Which of the following describes the effects of a spurious start of the 32 ABFP and why?

a. Tave will lower, Reactor Power will rise due to the steam demand required to start the 32 ABFP.

b. Tave will lower, Reactor Power will rise due to the injection of cold auxiliary feedwater from the 32 ABFP.

c. Tave will lower, Reactor Power will rise due to both the steam demand required to start the 32 ABFP and the injection of cold auxiliary feedwater from the 32 ABFP.

d. Tave and Reactor Power will both be restored to normal by the automatic rod control system.

Answer: a Explanation I Justification

a. Correct. Tave will lower, Reactor Power will rise due to the steam demand required to start the 32 ABFP. The normal operations lineup has the discharge feed regulating valves for the 32 ABFP aligned closed to prevent their injecting water.

b. Incorrect. Plausible because the student may believe that the discharge feed regulating valves for the turbine driven auxiliary feed pump are normally open as is the case for the motor driven pumps.

c. Incorrect. Plausible because the student may believe that the discharge feed regulating valves for the turbine driven auxiliary feed pump are normally open as is the case for the motor driven pumps.

d. Incorrect. Plausible because the student may believe that the steam demand for the 32 ABFP will effect turbine impulse pressure, lowering Tret. causing control rods to insert to restore temperature and power.

Technical

References:

13LP-ILO-AFW001 Proposed References to be provided: None Learning Objective: Objective 3.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 5, 6, & 7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 062000A304 Ability to monitor automatic operation of the ac distribution system, including: Operation of inverter (e.g., precharging synchronizing light, static transfer)

Importance: 2.7 2.9 Question: #46 Failure of the 31 Static Inverter has caused the Automatic Static Transfer Switch to transfer to the alternate source of power. Electrical Maintenance has investigated and has requested that 31 Static Inverter be taken out of service. In accordance with 3-SOP-EL-002, Instrument Bus and Plant Computer Static Inverter Operation, how will Operations take the static inverter out of service?

a. Operations will verify that the static transfer switch is in the "Forward Transfer" position and then place the Maintenance Bypass Switch (MBS) in the BYPASS position.

b. Operations will verify that the static transfer switch is in the "Reverse Transfer" position and then place the Maintenance Bypass Switch (MBS) in the BYPASS position.

c. Operations will verify that the static transfer switch is in the "Reverse Transfer position and the Maintenance Bypass Switch (MBS) is in the INVERTER position.

d. Operations will verify that the static transfer switch is in the "Forward Transfer" position and the Maintenance Bypass Switch (MBS) is in the INVERTER position.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe that the "Forward Transfer" position is correct. When the static transfer switch transfer to the alternate source of power, it is considered to have "reverse transferred". The MBS in the BYPASS position is correct.

b. Correct. The static transfer switch automatically transfers to reverse transfer on inverter failure. Then per the procedure the Maintenance Switch is placed in BYPASS. See procedure section 4.3 and also Caution before step 4.1.16. See lesson plan section 11.D.

c. Incorrect. Plausible because the static transfer switch does automatically transfers to reverse transfer on inverter failure. Student may believe that the Maintenance Bypass switch is placed in the INVERTER position to take it out of service.

d. Incorrect. Plausible because the student may believe that the "Forward Transfer" position is correct. When the static transfer switch transfer to the alternate source of power, it is considered to have "reverse transferred". Student may believe that the Maintenance Bypass switch is placed in the INVERTER position to take it out of service. The MBS in the BYPASS position is correct.

Technical

References:

13LP-ILO-EDS118, 3-SOP-EL-002 System Description 27.5 Proposed References to be provided: None Learning Objective: Objectives E-3 and E-9 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)7&10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 063000A201 Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Importance: 2.5 3.2 Question: #47 Given the following:

• Unit 3 is operating at 100% power

• A ground on 34 Battery Charger causes a loss of power to 34 DC Power Panel.

• 34 Battery Charger is damaged by the ground

• 34 DC Power Panel and 34 Battery are NOT damaged Which of the following is correct regarding the impact of this event and what actions, if any, can be taken?

a. 34 Instrument Bus will lose power. 3-AOP-IB-1; Loss of Power to an Instrument Bus, will address this by powering the bus from MCC-36C.

b. 34 Instrument Bus will lose power. 3-AOP-IB-1; Loss of Power to an Instrument Bus, will address this by powering the bus from MCC-36B.

c. 34 Instrument Bus will remain energized. 3-AOP-DC-1; Loss of a 125V DC Panel, will provide direction to re-energize 34 DC Panel using 35 Battery Charger.

d. 34 Instrument Bus will remain energized. 3-AOP-DC-1; Loss of a 125V DC Panel, will NOT allow re-energizing 34 DC Panel using 35 Battery Charger because the plant is above Mode 5.

Answer: c Explanation I Justification

a. Incorrect because 34 Instrument Bus will remain energized. Plausible because 34 Instrument Bus used to not automatically switch to a backup power source. Additionally, there are two back up feeds; one from MCC-36B and the other from MCC-36C.

b. Incorrect because 34 Instrument Bus will remain energized. Plausible because 34 Instrument Bus used to not automatically switch to a backup power source. Additionally, there are two back up feeds; one from MCC-36B and the other from MCC-36C.

c. Correct. 34 Instrument Bus will remain energized. 3-AOP-DC-1 will utilize 35 Battery Charger. Instructions and P&Ls in 3-SOP-EL-003 are applicable.

d. Incorrect. Plausible because there are restrictions on using 35 Battery Charger above 200°F due to the potential for cross tying buses. 3-AOP-DC-1 will utilize 35 Battery Charger. Instructions and P&Ls in 3-SOP-EL-003 are applicable.

Technical

References:

13LP-ILO-AOPDC1, 3-AOP-DC-1 3-SOP-EL-003 Proposed References to be provided: None Learning Objective: Objectives 1.0 & 2.0 Question Source: Bank - Indian Point 3 - Question # 25118 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 064000K405 Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following: Incomplete-start relay Importance: 2.8 3.2 Question: #48 An operator is manually starting the 31 Emergency Diesel Generator (EOG) for a Technical Specification Surveillance run. The operator depresses the start pushbutton which completes the start circuit and applies 125 VDC to the engine start relay coil. During this manual start, the diesel trips and the operator receives an Overcrank Alarm. The OVERCRANK alarm is actuated if the engine fails to attain speed within _ _ seconds of receiving the start signal as sensed by _ __

a. 10, lube oil pressure at 30 psig

b. 15, jacket water pressure at 8 psig

c. 10, jacket water pressure at 8 psig

d. 15, lube oil pressure at 30 psig Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may remember the design start requirement of 10 seconds and also remember interlocks associated with lube oil pressure being 30 psig.

Actually the Low Oil Pressure trip is 30 psig with a time delay of 20 seconds to allow for sufficient time for the oil pressure to increase to 30 psig. The 20 second timer is energized when jacket water pressure is > 8 psig as well.

b. Correct. The OVERCRACK alarm is actuated if the engine fails to attain speed within 15 seconds of receiving the start signal as sensed by jacket water pressure (set at 8 psig). See System Description 27.3, section 2.6.3.

c. Incorrect. Plausible because the student may remember the design start requirement of 10 seconds and the second part of the distractor is correct.

d. Incorrect. Plausible because the first part is correct. Also plausible because the student may remember interlocks associated with lube oil pressure and engine start. Actually the Low Oil Pressure trip is 30 psig with a time delay of 20 seconds to allow for sufficient time for the oil pressure to increase to 30 psig. The 20 second timer is energized when jacket water pressure is > 8 psig as well.

Technical

References:

13LP-ILO-EDSEDG System Description 27.3; Emergency Diesels

Proposed References to be provided: None Learning Objective: Objectives 4 & 6 Question Source: Modified - Indian Point 3 - Question # 23776 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 7 & 10 NOTE: Could also associate this with KIA 4.02

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 0640002132 Ability to explain and apply system limits and precautions Importance: 3.8 4.0 Question: #49 Given the following conditions:

• The Unit has experienced a Safety Injection and a Loss of Offsite Power

• The 32 EOG is carrying 1740 KW

• It is desired to start the 33 CCW pump

• The 33 CCW pump is a 220 KW load Which of the following is the MINIMUM amount of load that must be reduced before starting the 33 CCW pump that would ensure the final diesel loading is at or below the CONTINUOUS run load limit for 32 EOG?

a. No load reduction required

b. 10 KW

c. 110 KW

d. 210 KW Answer: d Explanation I Justification

a. Incorrect. Plausible because the student may believe that the continuous load limit is 2000 KW. 2000 KW is the~ hour limit.

b. Incorrect. Plausible because the student may believe that the continuous load limit is 1950 KW. 1950 KW is the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit.

c. Incorrect. Plausible because the student may believe that the continuous load limit is 1850 KW. 1850 is 100 KW below the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit.

d. Correct. 1740 KW plus 220 KW is 1960 KW, which is 210 KW above the continuous load limit of 1750 KW. See 3-SOP-EL-015, P&L 2.3.

Technical

References:

13LP-ILO-EOSEOG 3-SOP-EL-001, Attachment 4 3-SOP-EL-015, P&L 2.3 and Attachment 3 Proposed References to be provided: None Learning Objective: Objective 5

Question Source: Bank - IP3 - Question# 1578 (slightly modified)

Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 7, 8, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 073000K301 Knowledge of the effect that a loss or malfunction of the PRM system will have on the following: Radiative effluent releases Importance: 3.6 4.2 Question: #50 A loss of power has occurred to the following process radiation monitors, R-12; Containment Radiogas Monitor, R-14; Plant Vent Radiogas Monitor, and R-27; Wide Range Plant Vent Gas Monitor. Which of the following are common system responses to a loss of power to these three radiation monitors.

I. CHANNEL FAILURE Alarm II. Containment Evacuation Alarm Ill. Containment Purge Supply and Exhaust Valves will Close IV. PAB Exhaust will be Diverted Through Charcoal Filters V. Flow Control Valve (RCV-014) in the Waste Gas Discharge Line will Close VI. Containment Pressure Relief Valves will Close

a. I, II, VI

b. II, Ill, IV

c. Ill, IV, V

d. I, Ill, VI Answer: d Explanation I Justification

a. Incorrect. Plausible because the student may believe that the Plant Vent radiation monitors also cause the initiation of the containment evacuation alarm since they are a backup to the Containment monitor and provide containment isolation. I and IV are correct.

b. Incorrect. Plausible because the student may believe that the Plant Vent radiation monitors also cause the initiation of the containment evacuation alarm since they are a backup to the Containment monitor and provide containment isolation. Ill and IV are correct.

c. Incorrect. Plausible because Ill and IV are correct and two of the three monitors cause V to be correct.

d. Correct. The common system responses are a "CHANNEL FAILURE" ALARM due to the loss of power and all three cause containment ventilation isolation.

Technical

References:

13LP-ILO-RMSPRM Proposed References to be provided: None

Learning Objective: Objective F Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)7, 10, & 11

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 078000A401 Ability to manually operate and/or monitor in the control room: SWS pumps Importance: 2.9 2.9 Question: #51 Given the following:

• The plant is at 100% power

• 31, 32, and 35 Service Water Pumps are running

• 33, 34, and 36 Service Water Pumps are in AUTO

• The Service Water Pump Mode Selector Switch is in the 1-2-3 position.

• The 31 Emergency Diesel Generator is tagged for maintenance.

Assuming no operator action, which of the following describes the status of the SW Pumps after a Reactor Trip coincident with a Loss of Off-Site power occurs?

a. 31, 33, 34, and 36 Service Water Pumps will be running

b. 31 and 33 Service Water Pumps will be running

c. Only 33 Service Water Pump will be running

d. 34, 35, and 36 Service Water Pumps will be running Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may assume that all service water pumps will start except those associated with 31 EOG, that being 32 and 35 SW Pumps.

b. Correct. Only 31 and 33 SW Pumps start because 32 is powered from 31 EOG which is tagged for maintenance. Only essential service water header pumps start (1-2-3).

c. Incorrect. Plausible because the student may believe that 31 EOG powers both 31 and 32 Service Water Pumps.

d. Incorrect. Plausible because the student may believe that the non-essential pumps or the pumps not selected will start on a non-SI blackout. Student may also believe that 31 EOG powers both 31 and 32 Service Water Pumps.

Technical

References:

13LP-ILO-SW001, System Description 24.0 Proposed References to be provided: None Learning Objective: Objectives 2.0, 3.0, and 6.0 Question Source: Modified Bank Question# 25927

Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)?

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 078000K107 Knowledge of the physical connections and/or cause-effect relationships between SWS and the following systems: Secondary closed cooling water Importance: 2.5 2.3 Question: #52 During normal operating conditions, which of the following loads is cooled directly from the Conventional Plant Non-essential Service Water Header?

I. Main Boiler Feed Pump L.O. Coolers II. Turbine Hall Closed Cooling Ill. Instrument Air Closed Cooling IV. Exciter Air Coolers V. Hydrogen Coolers VI. Heater Drain Pumps VII. Component Cooling Heat Exchangers

a. I, IV, VI

b. II, Ill, VII

c. II, IV, V

d. l,V, VI Answer: c Explanation I Justification

a. Incorrect. Plausible because Main Boiler Feed Pumps and Heater Drain Pumps are secondary non-essential components. The Main Boiler Feed Pump pedestals and the Heater Drain Pumps are also both cooled by Turbine Hall Closed Cooling, but not service water directly. Exciter Air Coolers are a correct answer.

b. Incorrect. Plausible because Turbine Hall Closed Cooling is correct, the #33 Instrument Air Compressor is cooled by Turbine Hall Closed Cooling, and the Component Cooling Heat Exchangers are "Non-essential" loads but not on the conventional header.

c. Correct. Turbine Hall Closed Cooling, Exciter Air Coolers, and Hydrogen Coolers are all Conventional Non-essential Service Water Header loads.

d. Incorrect. Plausible because Main Boiler Feed Pumps and Heater Drain Pumps are secondary non-essential components. The Main Boiler Feed Pump pedestals and the Heater Drain Pumps are also both cooled by Turbine Hall Closed Cooling, but not service water directly. Hydrogen Coolers are a correct answer.

Technical

References:

13LP-ILO-SW001, System Description 24.0 Proposed References to be provided: None Learning Objective: Objective 1.0 Question Source: Modified Bank Question #14315 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41(b)4

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 078000K103 Knowledge of the physical connections and/or cause-effect relationships between IAS and the following systems: Containment Air Importance: 3.3 3.4 Question: #53 With the Unit at 100%, a spurious Safety Injection has been initiated due to human error during maintenance functional testing. The operators have transitioned to 3-ES-1.1; SI Termination. In accordance with 3-ES-1.1; SI Termination, what actions are required from the control room to restore Instrument Air to Containment

a. Rest SI, Reset Phase A, Press Instrument Air Reset Pushbutton, Open IA-PCV-1228

b. Reset SI Only, valve IA-PCV-1228 will return to the pre-SI position Automatically

c. Reset SI, Reset Phase A, Reset Phase B, Open IA-PCV-1228

d. Reset Phase A Only, valve IA-PCV-1228 will return to the pre-SI position Automatically Answer: a Explanation I Justification

a. Correct. See 3-ES-1.1; SI Termination procedure steps 1 through 3. PCV-1228, supply to VC closes directly on a Phase A signal.

b. Incorrect. Plausible because the student may believe that because it was a spurious SI signal, only SI is required to be reset. Also plausible because original plant design provided for valves to automatically reset to their previous position upon reset of the containment Phase A isolation signal (System Description 10.7, page 4).

c. Incorrect. Plausible because the procedure resets Phase B if actuated, student may believe Phase B needs to be reset also. Student may also believe that PCV-1228 is a Phase B valve.

a. Incorrect. Plausible because PCV-1228, supply to VC closes directly on a Phase A signal.

Also, plausible because original plant design provided for valves to automatically reset to their previous position upon reset of the containment Phase A isolation signal.

Technical

References:

13LP-ILO-EOPE10, 3-ES-1.1 System Description 10. 7; Containment Isolation 13LP-ILO-IA001 System Description 29.2; Instrument Air Proposed References to be provided: None Learning Objective: Objectives 20.0 and 8.0 respectively

Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41(b)7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 103000A205 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Emergency containment entry Importance: 2.9 3.9 Question: #54 The unit is in Mode 3, 547°F, 2235 psig, preparing to startup following a refueling. Containment Integrity has been established.

The control room has determined that there is a small RCS leak inside the containment. The Shift Manager has determined that an expedited emergency containment entry should be made to determine the leak location.

After discussion with Health Physics, Maintenance, and the Operations Manager, it is determined that three work parties will be needed to investigate the RCS leak.

In accordance with OAP-007; "Containment Entry and Egress", which of the following is the minimum number of people in each work party and the maximum number of personnel permitted in the containment under the established conditions.

a. 2,24

b. 3, 12

c. 2, 12

d. 3,24 Answer: a Explanation I Justification

a. Correct. Precaution and Limitation 2.14 states; "for personnel safety, a 24 person guideline is the limit for VC entries when in Modes 3 or 4." P&L 2.22 also states; "When Containment Integrity is established in Modes 1, 2, 3, or 4, Then Entry Parties SHALL consist of at least 2 people, and individuals SHALL NOT separate into parties of less than two people."

b. Incorrect. Plausible because as per P&L 2.13 and the note prior to step 4.4.13; "If RX is critical in Modes 1 or 2, then the number of people in VC at any one time SHALL NOT exceed 12."

c. Incorrect. Plausible because the first part is correct per P&L 2.22 and because as per P&L 2.13 and the note prior to step 4.4.13; "If RX is critical in Modes 1 or 2, then the number of people in VC at any one time SHALL NOT exceed 12.

d. Incorrect. Plausible because the second part of the answer is correct per P&L 2.14 which states; "for personnel safety, a 24 person guideline is the limit for VC entries when in Modes 3 or4."

Technical

References:

13LP-ILO-VCVCB, OAP-007 Proposed References to be provided: None Learning Objective: Objective 8.0 Question Source: Modified Bank Question #3755 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)9

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 1 KIA# 103000A301 Ability to monitor automatic operation of the containment system, including: Containment Isolation Importance: 3.9 4.2 Question: #55 Given the following conditions:

• The Unit was at 100% when a human error by an l&C Technician caused a Safety Injection Signal to be generated.

• The crew is performing 3-E-O, Reactor Trip or Safety Injection and has not yet transitioned to 3-ES-1.1, SI Termination.

• The BOP Operator is performing the actions of 3-R0-1, BOP Operator Action During Use of EOPs.

The following valves are identified as being in the OPEN position:

a. AOV-204B, Normal Charging Isolation

b. TCV-1104, Service Water Containment Cooling Full Flow Isolation

c. AC-AOV-791, Excess Letdown Heat Exchanger CCW Isolation

d. IV-AOV-1410, Isolation Valve Seal Water System Flow Control Valve For the current plant conditions, which ONE of the following valves must be repositioned to place it in the required accident configuration?

Answer: c Explanation I Justification

a. Incorrect. Plausible because the student may believe that charging flow needs to be isolated for the spurious Safety Injection Signal, but 3-E-O step 11 has the reactor operator establish charging flow and verify AOV-204B is open.

b. Incorrect. Plausible because the student may believe that the Containment Fan Cooling Units are not needed for the spurious safety injection signal. The TCV-1004 opens automatically from a safety injection signal.

c. Correct. Excess Letdown Heat Exchanger is considered a nonessential process and is isolated on a Phase A signal caused by the Safety Injection Signal. This valve should be closed. Student may think that because it is a CCW valve, it is only isolated by a Phase B Signal.

a. Incorrect. Plausible because the student may be confused on how the Isolation Valve Seal Water System operates and expect any valves associated with containment isolation to be closed.

Technical

References:

13LP-ILO-EOPEOO, 3-E-Or4 Proposed References to be provided: None Learning Objective: Objective 12.0 Question Source: Modified Bank - IP2 Question # 24353 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41(b)7 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 001000K205 Knowledge of bus power supplies to the following: M/G sets (Control Rod Drive System)

Importance: 3.1 3.5 Question: #56 Given the following conditions:

• The Unit is at 100% power.

• #31 Rod Drive MG Set is cleared and tagged for motor replacement.

• #32 Emergency Diesel Generator is cleared and tagged for a fuel line leak repair.

• A fault occurred in the Buchanan switchyard resulting in a loss of 138KV power.

What is the immediate impact on the plant?

a. The reactor will remain at approximately 100% power because #32 Rod Drive MG Set still energized from 480 Volt Bus 5A.

b. The reactor will trip because power is lost to the only remaining Rod Drive MG Set with the loss of 480 Volt Bus 6A.

c. The reactor will trip because power is lost to the only remaining Rod Drive MG Set with the loss of 480 Volt Bus 3A.

d. The reactor will remain at approximately 100% power because #32 Rod Drive MG Set still energized from 480 Volt Bus 2A.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe that #32 Rod Drive MG Set is powered from 480 Volt Bus 5A via the 33 EOG.

b. Correct. #32 Rod Drive MG Set is powered from 480 Volt Bus 6A which is de-energized due to the loss of 138KV and #32 EOG being cleared and tagged.

c. Incorrect. Plausible because the student may believe that #32 Rod Drive MG Set is powered from 480 Volt Bus 3A. Also plausible because Bus 3A does not lose power as of the result of the switchyard fault.

a. Incorrect. Plausible because the student may believe that #32 Rod Drive MG Set is powered from 480 Volt Bus 2A. Also plausible because Bus 2A does not lose power as of the result of the switchyard fault. Bus 2A is actually the power supply for the #31 Rod Drive MG Set, but that MG Set is cleared and tagged.

Technical

References:

13LP-ILO-ICROD Proposed References to be provided: None Learning Objective: Objective D Question Source: Modified Bank Question# 25961 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 011000K401 Knowledge of PZR LCS design feature(s) and/or interlock(s) which provide for the following: Operation of PZR heater cutout at low PZR level Importance: 3.3 3.7 Question: #57 The plant is at 100% power steady with the following conditions:

• All Control Systems are in Automatic.

• Pressurizer Level Channel Defeat Switch is selected to the "Defeat I" position.

• Backup Heater Group #32 is ON.

Subsequently, Pressurizer Level Channel LT-461 fails to zero. After five minutes, which of the following describes the response to the instrument failure assuming no operator action is taken?

a. Actual Pressurizer level rises, charging flow lowers, all Pressurizer Heaters de-energize.

b. Actual Pressurizer level rises, charging flow rises, all Pressurizer Heaters de-energize.

c. Actual Pressurizer level rises, charging flow rises, Modulating Heater Group OFF and Backup Heater Group #32 remains ON.

d. Pressurizer level remains on program, charging flow remains constant, Backup Heater Group #32 remains ON.

Answer: a Explanation I Justification

a. Correct. With the PZR Level Channel Defeat Switch in the "Defeat I" position, LT-461 is the alarm channel and LT-460 is the controlling channel. When the alarm channel fails to zero, letdown isolates via LCV-460 causing a charging/letdown mismatch with actual pressurizer level rising. The controlling channel will see actual level greater than program level and lower charging flow. The alarm channel failing to zero will also de-energize all pressurizer heaters.

b. Incorrect. Plausible because the student may believe that LT-461 is the controlling channel based on the position of the channel defeat switch. This would be the correct response for the controlling channel. The controlling channel would see a low level and raise charging flow to compensate.

c. Incorrect. Plausible because the student may believe that LT-461 is the controlling channel.

The student may also believe that because the Backup Heater Group #32 was in manual

"ON" that the failed level channel would not affect the operation of the heater. Incorrect understanding of the heater cut-out circuit.

d. Incorrect. Plausible because this would be the effect if, LT-461 was the defeated channel.

The defeated channel has no input to the Pressurizer Level Control System.

Technical

References:

13LP-ILO-ICPZLV Proposed References to be provided: None Learning Objective: Objective E-5 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41(b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 014000K202 Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of power to the RPIS Importance: 3.1 3.6 Question: #58 Unit 3 is operating at 100% power with the IRPI System aligned to its backup power supply.

Offsite power is lost, the plant remains on line and all equipment functions as designed. How does this affect power to the IRPls and what Tech Spec actions are required, if any?

a. The IRPls will lose power since the backup power supply is MCC-39. Rod Control must immediately be placed in Manual to comply with Technical Specification 3.1. 7.B.

b. The IRPls will lose power since the backup power supply is MCC-39. There are no applicable Technical Specification requirements associated with a loss of power to the IRPls.

c. The IRPls will have power because the backup power supply is 36C. If the IRPls had been on normal power (MCC-39), they would have lost power, and Rod Control would have had to be placed in Manual to comply with Technical Specification 3.1.7.B.

d. The IRPls will have power because the backup power supply is 36C. If the IRPls had been on normal power (MCC-39), they would have lost power. However, there are no applicable Technical Specification requirements associated with a loss of power to the IRPls.

Answer: a Explanation I Justification

a. Correct. The IRPI backup power supply is MCC-39 which is fed from 480 Volt Bus 5A. Bus 5A will lose power due to the Loss of Off-Site Power; therefore power has been lost to the IRPI System. Technical Specification 3.1.7.B requires Control Rods in Manual with a completion time of immediately if more than one IRPI per group is inoperable. See Note prior to 3-AOP-ROD-1, step 4.136.

b. Incorrect. Plausible because the IRPI backup power supply is MCC-39 which is fed from 480 Volt Bus 5A. Bus 5A will lose power due to the Loss of Off-Site Power; therefore power

has been lost to the IRPI System. Student may not recognize that all IRPls are inoperable due to loss of indication.

c. Incorrect. Plausible because the student may believe that MCC-36C is the backup power supply to the IRPI System. Plausible because MCC-36C is fed from 480 Volt Bus 2A which still has power following the Loss of Off-Site Power event. Technical Specification requirement is also true.

d. Incorrect. Plausible because the student may believe that MCC-36C is the backup power supply to the IRPI System. Plausible because MCC-36C is fed from 480 Volt Bus 2A which still has power following the Loss of Off-Site Power event. Student may not recognize that all IRPls are inoperable due to loss of indication.

Technical

References:

13LP-ILO-ICRPI Proposed References to be provided: None Learning Objective: Objectives C, F, and H Question Source: Modified Bank Question# 25119 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 033000A302 Ability to monitor automatic operation of the Spent Fuel Pool Cooling System including:

Spent fuel leak or rupture Importance: 2.9 3.1 Question: #59 Design features of the Spent Fuel Cooling and Purification Loop ensure the fuel stored in the Spent Fuel Pool will not become uncovered as a result of any postulated loss of system integrity.

Which one of the following correctly states these design features?

a. The Spent Fuel Cooling Pump suction pipe has an anti-siphon hole, and the Spent Fuel Cooling Pump will automatically trip on Low Spent Fuel Level.

b. The Spent Fuel Cooling Pump suction pipe has an anti-siphon hole, and the return line outlet is above the fuel level.

c. The Spent Fuel Cooling loop does not have drains located below the level of the fuel, and the makeup flow capacity exceeds the System flow.

d. The Spent Fuel Cooling Pump suction pipe inlet is above the fuel level, and there is an anti-siphon hole on the return pipe.

Answer: d Explanation I Justification

a. Incorrect. Plausible because the return line has a 'X inch hole to prevent siphoning and the student may confuse suction and discharge piping design features.

b. Incorrect. Plausible because the return line has a 'X inch hole to prevent siphoning and the student may confuse suction and discharge piping design features.

c. Incorrect. Plausible because there are no drain lines in the spent fuel pit.

d. Correct. The SFP cooling pump suction is approximately 6 feet below the surface of the pit (remember 23 feet of water is maintained above the fuel assemblies) and the SFP pump discharges into the pit approximately 7 feet above the top of the fuel assemblies. An anti-siphon hole in the discharge pipe prevents draining due to return line failure.

Technical

References:

13LP-ILO-SFP001 Proposed References to be provided: None Learning Objective: Objective 7.0 Question Source: Bank Question # 23900

Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)?

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 0340002140 Knowledge of refueling administrative requirements Importance: 2.8 3.9 Question: #60 In accordance with 3-REF-003-GEN, Fuel Movement Requirements - Core Reload, WHEN the transfer system is unattended, THEN the transfer cart

a. SHALL always be stored in the Fuel Storage Building (FSB) to facilitate closing of the fuel transfer systems gate valve.

b. SHALL always be stored in the Containment Building (VC) to facilitate closing of the fuel transfer systems gate valve.

c. SHALL always be stored in the Containment Building (VC) to ensure the UPENDER CLEAR interlock is met.

d. SHALL always be stored in the Fuel Storage Building (FSB) to ensure the UPENDER CLEAR interlock is met.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may believe that the transfer cart's home position is the FSB and that in order to close the gate valve, the cart must be positioned in the FSB.

b. Correct. 3-REF-003-GEN, precautions and limitations steps 2.21 and 2.33 both state that when the transfer system is left unattended, the transfer cart shall always be stored in the containment building to facilitate the closing of the fuel transfer systems gate valve.

c. Incorrect. Plausible because the containment is the transfer cart's home position and the procedure precaution states that it should be stored in its home position when the transfer system is left unattended. Also plausible because the UPENDER CLEAR interlock does exist and it ensures that the respective upender is clear of either the manipulator crane or the fuel handling crane. Student may believe that having this interlock made up allows manipulations in both respective buildings without use of the transfer system.

d. Incorrect. Plausible because the student may believe that the transfer cart's home position is the FSB. Also plausible because the UPENDER CLEAR interlock does exist and it ensures that the respective upender is clear of either the manipulator crane or the fuel handling crane. Student may believe that having this interlock made up allows manipulations in both respective buildings without use of the transfer system.

Technical

References:

13LP-ILO-FHD-001, 3-REF-003-GEN System Description 17.0

Proposed References to be provided: None Learning Objective: Objective C (E-3)

Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 035000K113 Knowledge of the physical connections and/or cause-effect relationships between the S/Gs and the following systems:

Condensate System Importance: 2.7 2.8 Question: #61 Operators are responding to a Loss of Secondary Heat Sink Red Path in accordance with 3-FR-H .1. The only feed source that is available is 31 Condensate Pump. RCS pressure is greater than Steam Generator Pressure. Steam Generator narrow range (NR) levels are off scale low and wide range (WR) levels are 45% and lowering. What actions are required to mitigate the event?

a. Depressurize at least one SG to less than 380 psig to establish feed flow. Closely monitor SG WR levels for Bleed and Feed criteria during the SG depressurization.

Bleed and Feed is initiated if the average of the 3 lowest WR SG levels is less than 20%.

b. Depressurize all but one SG to less than 380 psig to establish feed flow. Closely monitor SG WR levels for Bleed and Feed criteria during the SG depressurization.

Bleed and Feed is initiated if the average of all WR SG levels is less than 30%.

c. Depressurize at least one SG to less than 380 psig to establish feed flow. Closely monitor SG WR levels for Bleed and Feed criteria during the SG depressurization.

Bleed and Feed is initiated if the average of all WR SG levels is less than 30%.

d. Depressurize all but one SG to less than 380 psig to establish feed flow. Closely monitor SG WR levels for Bleed and Feed criteria during the SG depressurization.

Bleed and Feed is initiated if the average of the 3 lowest WR SG levels is less than 20%.

Answer: a Explanation I Justification

a. Correct. See EOP 3-FR-H-1, Attachment 3; Establishing Feedwater Flow from Secondary Plant, steps 3 & 4. SG pressure must be lowered to less than 380# to allow a condensate pump to feed. Step states "at least one". Bleed and Feed criteria is stated in step 2 of the procedure and foldout page as "average of the lowest WR SG levels less than 20% (30%

adverse)."

b. Incorrect. Plausible because the student may believe that he needs to leave one SG untouched for future needs. Also plausible because the student may remember the adverse SG level of 30%. Containment conditions are normal.

c. Incorrect. Plausible because the first part of the distractor is correct and because the student may remember the adverse SG level of 30%. Containment conditions ae normal.

d. Incorrect. Plausible because the student may believe that he needs to leave one SG untouched for future needs. Also plausible because the second part of the distractor is correct.

Technical

References:

13LP-ILO-EOPFRH 3-FR-H.1 System Description 20.0; Condensate Proposed References to be provided: None Learning Objective: Objectives 4.0 & 6.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 4, 8, & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 045000A106 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including: Expected response of secondary plant parameters following T/G trip Importance: 3.3 3.7 Question: #62 The following plant conditions exist:

• The Plant is operating at 100% power.

• All systems are lined up in their normal lineups.

• All control systems are in automatic.

• Main Generator output breakers 1 and 3 trip due to a pilot wire trip.

Which of the following describes the expected immediate plant response?

a. SIG pressure initially RISES as main turbine is lost, SIG levels initially LOWER due to shrink, feed flow initially RISES.

b. SIG pressure initially RISES as main turbine is lost, SIG levels initially LOWER due to shrink, feed flow initially LOWERS.

c. SIG pressure initially LOWERS as main turbine is lost, SIG levels initially LOWER due to shrink, feed flow initially RISES.

d. SIG pressure initially LOWERS as main turbine is lost, SIG levels initially RISE due to lower steam pressure, feed flow initially LOWERS.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the student may think that an ESF actuation for AFW will cause levels to initially rise, however that is not an immediate response. Steam Flows initially rapidly decrease due to the loss of the turbine causing initial feed flow to decrease as well.

b. Correct. SIG pressures rapidly increase to - 1000 psig due to the immediate drop in steam flow caused by the turbine trip. Continued steam dump system action will maintain SG pressure at - 1000 psig. SIG NR Levels decrease rapidly off-scale low due to the increase in SG pressure and the drop in SG recirculation flow causing shrink in the SG. Steam Flows

initially rapidly decrease due to the loss of the turbine causing initial feed flow to decrease as well.

c. Incorrect. Plausible because the student may believe that less steam production caused by the loss of turbine load will result in lower SIG pressures.

a. Incorrect. Plausible because the student may believe that less steam production caused by the loss of turbine load will result in lower SIG pressures.

Technical

References:

13LP-ILO-EOPEOO, 3-E-Or4 Proposed References to be provided: None Learning Objective: Objective 1.0 Question Source: Bank Question # 17368 IP3 NRC Exam 2006 Seabrook NRC Exam 2003 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)5

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 071000K305 Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following: ARM and PRM systems Importance: 3.2 3.2 Question: #63 Given the following conditions:

• The Waste Gas System is aligned for normal operation

• 31 Large Gas Decay Tank is in-service

• 32 Large Gas Decay Tank is in standby

• R-20, Gas Decay Tank Activity Monitor is NOT in alarm Which one of the following describes the system response (assuming no operator action) if the 31 Large Gas Decay Tank relief valve were to fail open?

a. Radiation monitor R-27, Wide Range Plant Vent Gas Activity Monitor will alarm causing the plant vent discharge valve, RCV-014 to close isolating the release.

b. Radiation monitor R-14, Plant Vent Gas Activity Monitor and R-27, Wide Range Plant Vent Gas Activity Monitor will both alarm causing the plant vent discharge valve, RCV-014 to close isolating the release.

c. Radiation monitor R-20, Gas Decay Tank Activity Monitor will alarm causing the plant vent discharge valve, RCV-014 to close isolating the release.

d. Radiation monitor R-14, Plant Vent Gas Activity Monitor and R-27, Wide Range Plant Vent Gas Activity Monitor may alarm, however the release will continue.

Answer: d Explanation I Justification

a. Incorrect. Plausible because the R-27, Wide Range Plant Vent Gas Monitor is used to monitor Gas Decay Tank releases and when in alarm, it will isolate a normal gas decay tank release path via the RCV-014.

b. Incorrect. Plausible because both the R-14, Plant Vent Gas Activity Monitor and R-27, Wide Range Plant Vent Gas Activity Monitor are used to monitor Gas Decay Tank releases and when in alarm, they will isolate a normal gas decay tank release path via the RCV-014.

c. Incorrect. Plausible because the R-20, Gas Decay Tank Activity Monitor is used to monitor the radioactive gas moving through the Waste Gas System and accumulating in the in

service gas decay tank. However, there is no automatic isolation I actions associated with the monitor in alarm. It is used to warn the operators of the potential for reaching the tank curie limit of 19,400 curies.

d. Correct. Although activity in the in service gas decay tank may cause plant vent radiation monitors to alarm, the release via the relief valve will continue because the relief valve has a direct lease path to the plant vent downstream of the RCV-014. It is not capable of being isolated.

Technical

References:

13LP-ILO-GWR001 System Description 5.2; Gaseous Waste Disposal -

Figure 5.2-4 System Description 12.0; Radiation Monitoring Proposed References to be provided: None Learning Objective: Objectives 9e & 11 e Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 7 & 13

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 079000A401 Knowledge of SAS (Station Air System) design feature(s) and/or interlock(s) which provide the following: Cross-connect with IAS (Instrument Air System)

Importance: 2.9 3.2 Question: #64 Which of the following describes how Station Air is automatically cross-tied to the Instrument Air System?

a. When instrument air system pressure falls to 95 psig, PCV-1142 is automatically opened to admit air to the system. Supplied downstream of the heatless desiccant dryers, this air is not conditioned the same as the air normally supplied by the Instrument Air compressors.

b. When instrument air system pressure falls to 90 psig, PCV-1142 is automatically opened to admit air to the system. Supplied upstream of the heatless desiccant dryers, this air is conditioned the same as the air normally supplied by the Instrument Air compressors.

c. When instrument air system pressure falls to 95 psig, PCV-1142 is automatically opened to admit air to the system. Supplied upstream of the heatless desiccant dryers, this air is conditioned the same as the air normally supplied by the Instrument Air compressors.

d. When instrument air system pressure falls to 90 psig, PCV-1142 is automatically opened to admit air to the system. Supplied downstream of the heatless desiccant dryers, this air is not conditioned the same as the air normally supplied by the Instrument Air compressors.

Answer: b Explanation I Justification

a. Incorrect. Plausible because the standby IA compressor starts at 95 psig. Student may also believe that the station air cross-tie line comes in downstream of the dryers.

b. Correct. When instrument air system pressure falls to 90 psig, PCV-1142 is automatically opened to admit air to the system. Supplied upstream of the heatless desiccant dryers, this air is conditioned the same as the air normally supplied by the Instrument Air compressors.

See System Description 29.2, page 19, section 2.9.1. Also see lesson plan.

c. Incorrect. Plausible because the standby IA compressor starts at 95 psig. The discussion regarding conditioned air is correct.

d. Incorrect. Plausible because the valve opens at 90 psig, first part of distractor is correct.

Student may believe that the station air cross-tie line comes in downstream of the dryers.

Technical

References:

13LP-ILO-IA001 System Description 29.2 Proposed References to be provided: None Learning Objective: Objective 3.0 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b)7

Exam Outline Cross

Reference:

Level: RO SRO Tier# 2 Group# 2 KIA# 086000K604 Knowledge of the effect of a loss or malfunction on the Fire Protection Systems following will have on the: Fire, smoke, and heat detectors Importance: 2.6 2.9 Question: #65 Given the following conditions and TRM references:

• The Unit is in Mode 4.

• A loss of 125VDC has caused a loss of power to the C02 Fire Protection Control Panels.

• All Diesel Generator Room(s) fire detectors and fire barriers are operable.

• The 31 Emergency Diesel Generator Wet Pipe Sprinkler System is isolated for maintenance to replace faulty sprinkler heads.

What are the required compensatory actions for the inoperable fire suppression equipment and can C02suppression be manually discharged into the Diesel Generator Room(s)?

a. Establish a continuous fire watch and backup fire suppression equipment. Yes, on a loss of DC power, a C02suppression discharge can be manually initiated.

b. Establish an hourly fire watch patrol with backup fire suppression equipment. No, a loss of DC power causes the master pilot valve to fail closed precluding a manual initiation of C0 2suppression.

c. Establish an hourly fire watch patrol with backup fire suppression equipment. Yes, on a loss of DC power, a C0 2suppression discharge can be manually initiated.

d. Establish a continuous fire watch and backup fire suppression equipment. No, a loss of DC power causes the master pilot valve to fail closed precluding a manual initiation of C02 suppression.

Answer: a Explanation I Justification

a. Correct. Student is given various fire protection TRMs. TRM 3.7.A.2; Fire Protection Water Spray and Sprinkler Systems Action B, states if the Diesel Generator Building Water Spray System is inoperable, then establish a continuous fire watch and backup fire suppression equipment. Procedure 3-SOP-FP-003; Fire Protection C0 2System Operation discusses operation with DC power unavailable.

b. Incorrect. Plausible because the student may believe that an hourly fire watch is acceptable because all the fire detectors and fire barriers are operable. If the 31 EDG sprinkler wasn't also inoperable, this could be true. The student may also see action A of TRM 3. 7.A.2 and believe it is applicable. Also plausible because the student may believe that a loss of DC renders the C0 2Fire Protection System completely unavailable, however, the master pilot valve fails open on a loss of DC power which enables manual actuation per 3-SOP-FP-003.

c. Incorrect. Plausible because the student may believe that an hourly fire watch is acceptable because all the fire detectors and fire barriers are operable. If the 31 EDG sprinkler wasn't also inoperable, this could be true. The student may also see action A of TRM 3. 7.A.2 and believe it is applicable. Also plausible because the second part of the distractor is true, Procedure 3-SOP-FP-003; Fire Protection C02System Operation discusses operation with DC power unavailable.

d. Incorrect. Plausible because the student may believe that a loss of DC renders the C02 Fire Protection System completely unavailable, however, the master pilot valve fails open on a loss of DC power which enables manual actuation per 3-SOP-FP-003. The first part of the distractor is correct.

Technical

References:

13LP-ILO-FPS001 System Description 29.6 3-SOP-FP-003 Proposed References to be provided: TRM 3.7.A.2; Fire Protection Water Spray and Sprinkler Systems, TRM 3.7.A.4; Fire Detection Systems, TRM 3.7.A.7; C02 Fire Protection System, and the respective bases documents.

Learning Objective: Objective 5.a Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b)7 NOTE:

Fire Protection Lesson Plan, 13LP-ILO-FPS001, Objective, 5.a states:

"Describe the effect that a loss or malfunction of the following will have on the Fire Protection system, a. Loss of Power to C02 System." Student needs to know that manual capability still exists and that fire detection equipment (fire, smoke, and heat detectors) are unaffected by the DC power loss. Automatic actuation actuated by at least two heat detectors, but auto actuation is effected by the loss of DC power.

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# Conduct of Operations KIA# 1940012129 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

Importance: 4.1 4.0 Question: #66 In accordance with OAP-019, Component Verification and System Status Control; Verification SHALL be performed when component manipulations are performed on certain safety systems to restore these systems to a functional or operable status. Which of the systems below require Verification in accordance with OAP-019?

I. Carbon Dioxide Fire Protection System

11. Condensate Ill. Fire Pump House IV. Gaseous Disposal System V. Instrument Air VI. Liquid Disposal System VI I. Station Air

a. I, Ill, V, VII

b. 11, IV, V, VI

c. I, Ill, IV, VI

d. I, II, Ill, IV Answer: c Explanation I Justification

a. Incorrect. Plausible because the student may believe that instrument air is an important safety system listed on Attachment 1, Safety Systems Requiring Verification, of OAP-019.

All other listed systems are correct.

b. Incorrect. Plausible because two of the systems listed are correct and the student may believe that both instrument air and condensate are important safety systems as both are required to maintain the unit on-line. Loss of either system can lead to a reactor trip and plant transient.

c. Correct. All four systems listed are included on Attachment 1, Safety Systems Requiring Verification, of OAP-019.

d. Incorrect. Plausible because the student may believe that condensate is an important safety system listed on Attachment 1, Safety Systems Requiring Verification, of OAP-019. Loss of the condensate system can lead to a reactor trip and plant transient. All other listed systems are correct.

Technical

References:

13LP-ILO-Admin-COO?, OAP-019 Proposed References to be provided: None Learning Objective: Objective (Generic KIA - Admin-COO?)

Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# coo KIA# 1940012144 Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation Importance: 3.9 3.8 Question: #67 Given the following conditions:

• The Unit is in a "De-Fueled" condition

• The Reactor Core has been completely off loaded to the Spent Fuel Pool

• Today is day 20 of the refueling outage

• A loss of Spent Fuel Cooling has just occurred

• The NPO reports from the field that Spent Fuel Pool Temperature is currently 140°F

• 3-AOP-SF-1; Loss of Spent Fuel Cooling has been entered Given reference 3-AOP-SF-1, Attachment 1, determine the Spent Fuel Pool heat-up rate.

a. 5°F/HR

b. 7.3°F/HR

c. 8.2°F/HR

d. 12°F/HR Answer: d Explanation I Justification

a. Incorrect. Plausible because the student may read graph incorrectly. The graph provides a time of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to reach 200°F from the current temperature of 140°F. Student may read the graph as 5°F/HR heat-up.

b. Incorrect. Plausible because the student may read the graph incorrectly. The graph provides a time of 8.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to reach 200°F from the temperature of 100°F, not the reported 140°F current temperature. Student could then still use the 8.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the 60°F rise from 140°F and come up with 7.3°F/HR.

c. Incorrect. Plausible because the student may read the graph incorrectly. The graph provides a time of 8.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to reach 200°F from the temperature of 100°F, not the reported 140°F current temperature. Student may then read the graph as 8.2°F/HR heat-up.

d. Correct. The graph provides a time of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to reach 200°F from the current temperature of 140°F. Student will note that it takes 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the pool to rise 60°F, therefore 12°F/HR.

Technical

References:

13LP-ILO-SFP001, 3-AOP-SF-1 Proposed References to be provided: 3-AOP-SF-1, Attach 1 (3-GRAPH-ACS-1 F & 1G)

Learning Objective: Objective 8a Question Source: Modified, IP3 Question# 18683 Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# EC KIA# 0000002223 Ability to track Technical Specification limiting conditions for operations Importance: 3.1 4.6 Question: #68 The Unit is operating at 100% when Reactor Engineering informs the Shift Manager that they have determined via the 31 EFPD Flux Map that the Heat Flux Hot Channel Factor (Fq(Z)) has exceeded its limit specified in the COLR.

Which of the following describes the required actions in accordance with Technical Specifications?

a. Within 15 minutes reduce THERMAL POWER 2 1% RTP for each 1% Fq(Z) exceeds limit.

b. Within 60 minutes reduce THERMAL POWER to< 50% RTP.

c. Within 30 minutes reduce THERMAL POWER to< 90% RTP.

d. Within 15 minutes reduce THERMAL POWER 2 3% RTP for each 1% Fq(Z) exceeds limit.

Answer: a Explanation I Justification

a. Correct. Technical Specification LCO 3.2.1 states that Fq(Z) shall be within the limits specified in the COLR. Action A states to reduce THERMAL POWER 2 1% RTP for each 1% Fq(Z) exceeds limit within 15 minutes.

b. Incorrect. Plausible because the AFD Tech Spec reduces THERMAL POWER to< 50%

within 30 minutes when the cumulative penalty time has exceeded 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Student may confuse Power Distribution Technical Specification actions and only remember the 60 minutes of penalty time.

c. Incorrect. Plausible because the AFD Tech Spec reduces THERMAL POWER to < 90%

within 15 minutes when AFD is outside the target band. Student may confuse Power Distribution Technical Specification actions.

d. Incorrect. Plausible because the 15 minute action time is correct per Tech Spec 3.2.1, but the 3% RTP for each 1% is the same reduction required by the QPTR Tech Spec for QPTR not within limit (after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). Student may confuse Power Distribution Technical Specification actions.

Technical

References:

Technical Specification 3.2.1 13LP-ILO-ICNXC

Proposed References to be provided: None Learning Objective: Objective E-3 Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 2 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# Equipment Control KIA# 1030002238 Knowledge of conditions and limitations in the facility license Importance: 3.6 4.5 Question: #69 In Modes 1, 2, 3, and 4 Containment Penetration flow paths with manual isolation valves may be intermittently opened under administrative control. What administrative controls are required to open a normally locked closed manual containment isolation valve while still complying with Technical Specifications and are there any time limitations regarding the opening on the valve

a. The administrative controls consist of using two operators to concurrently verify the position of the valve before and after manipulating the valve. The valve SHALL not remain open for greater than one hour.

b. The administrative controls consist of entering a "Tracking LCO" to document the time the valve is open. The valve SHALL not remain open for greater than four hours.

c. The administrative controls consist of entering the applicable Technical Specification Action Statement and prior to opening the valve, ensure that the penetration affected has been isolated by the use of one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve. The valve SHALL not remain open for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

d. The administrative controls consist of stationing a dedicated operator at the valve controls, who is in continuous communication with the control room. The valve SHALL only remain open as long as necessary to perform the intended function.

Answer: d Explanation I Justification

a. Incorrect. Plausible because in accordance with OAP-19 the valve may require concurrent verification. The student may believe that Tech Spec 3.6.1 is applicable during the valve opening (Containment inoperable), therefore allowing only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with the specification.

b. Incorrect. Plausible because a "Tracking LCO may be applicable from a configuration control process requirement. Also plausible because Tech Spec 3.6.3 action A has a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time.

c. Incorrect. Plausible because the actions for an inoperable containment isolation valve are exactly what is stated. Also plausible because Tech Spec 3.6.3 action Chas a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time.

d. Correct. OAP-019, step 4.1.1 and Technical Specification 3.6.3 BASES state; "The administrative controls consist of stationing a dedicated operator at the valve controls, who

is in continuous communication with the control room. The valve SHALL only remain open as long as necessary to perform the intended function."

Technical

References:

13LP-ILO-VCCIS OAP-019 Technical Specification 3.6.3 BASES Proposed References to be provided: None Learning Objective: Objective 4.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)7&10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# EC KIA# 1030002240 Ability to apply Technical Specifications for a system Importance: 3.4 4.7 Question: #70 Which one of the following individual RCS leakages will require a unit shutdown in accordance with Technical Specifications?

a. 0.5 gpm leak to containment atmosphere from an unknown source.

b. 8.0 gpm measured leakage to the PRT.

c. 0.8 gpm leakage past the inner reactor vessel flange a-ring.

d. 0.11 gpm primary to secondary leakage on 31 Steam Generator.

Answer: d Explanation I Justification

a. Incorrect. Plausible because student may believe any leakage to the containment atmosphere from an unknown source may require a plant shutdown because it could interfere with future leakage detection by the containment process radiation monitors.

b. Incorrect. Plausible because the student may not know that leakage to the PRT is considered identified leakage or they may believe that the tech spec only allows a maximum of 1 gpm identified leakage.

c. Incorrect. Plausible because the student may believe that reactor vessel flange leakage is considered pressure boundary leakage.

d. Correct. Technical Specification LCO 3.4.13 states that RCS operational LEAKAGE shall be limited to: 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG). 150 gallons per day is equal to .104 gpm. Or .11 gpm is equal to 158.4 gpd.

Technical

References:

Technical Specification 3.4.13 13LP-I LO-RCS001 Proposed References to be provided: None Learning Objective: Objective 9.0 Question Source: Modified Bank Question # 1671 Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 3 & 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# RC KIA# 1940012304 Knowledge of the radiation exposure limits under normal or emergency conditions Importance: 3.2 3.7 Question: #71 A plant worker has received 700 mrem TEDE dose for the current year. This worker has been assigned to work in a 200 mrem/hr radiation field.

How long will it take for this worker to reach the 10CFR20 limits and the Site Administrative limits for whole body dose without any dose extensions being authorized?

10CFR20 SITEADMIN

a. 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

b. 19.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 19.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />

c. 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

d. 19.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Answer: c Explanation I Justification

a. Incorrect. Plausible because student may believe that both the site and federal 10CFR20 limit are the same, 5 Rem per year.

b. Incorrect. Plausible because the student may believe that the total administrative annual limit (including dose extensions) applies, 4.5 Rem per year. The student may also believe that since dose extensions allow up to 4.5 Rem per year that it is also the federal limit.

c. Correct. The individual is allowed up to 5 Rem per year per 10CFR20.1201 and up to 2 Rem per year without any dose extensions in accordance with EN-RP-201; Dosimetry Administration.

d. Incorrect. Plausible because student may believe that since dose extensions allow up to 4.5 Rem per year that it is also the federal limit. The student may also believe that personnel are limited to 1 Rem per year without any dose extensions. 1000 mrem is a limit if the individual's lifetime dose is greater than their age.

Technical

References:

IOWKB-ILO-ADMOO EN-RP-201 10CFR20.1201 Proposed References to be provided: None Learning Objective: Objectives 2,3,4 ?

Question Source: Modified Bank Question # 24102

Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41(b)10 & 12

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# RC KIA# 1940012315 Knowledge of the radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance: 2.9 3.1 Question: #72 Which of the following Area Radiation Monitors is one that is checked in 3-E-1; Loss of Reactor or Secondary Coolant, to determine if a LOCA Outside Containment is occurring and if transition to 3-ECA-1.2, LOCA Outside of Containment is required?

a. R-69; 54' Elevation Pipe Penetration Area Monitor

b. R-70; 80' Elevation Fan House Area Monitor

c. R-6; Sampling Room Monitor

d. R-34A; 31 CVCS HUT Monitor Answer: a Explanation I Justification

a. Correct. See 3-E-1, Loss of Reactor or Secondary Coolant, Attachment 1; Plant Status Evaluation, steps 5 and 6. R-64 through R-69 are the listed radiation area monitors used to determine if PAB radiation conditions are normal. Deviation document states that the ERG's intent is that radiation monitors be used to locate the source of the leakage into the PAB.

b. Incorrect. Plausible because R-70 is an area radiation monitor in the PAB at elevation 80'.

It would not meet the intent of determining the source of the leak. It is not listed in 3-E-1, Attachment 1, step 5.

c. Incorrect. Plausible because R-6 is an area radiation monitor located on the 55' elevation of the PAB. It would not meet the intent of determining the source of the leak. It is not listed in 3-E-1, Attachment 1, step 5.

d. Incorrect. Plausible because R34A is an area radiation monitor located on the 33' elevation of the PAB. It would not meet the intent of determining the source of the leak. It is not listed in 3-E-1, Attachment 1, step 5.

Technical

References:

13LP-ILO-EOPE10, 3-E-1 r4 & Dev. Document Proposed References to be provided: None Learning Objective: Objective 19.0 Question Source: Modified Bank (Salem 2012 NRC Exam # 72)

Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 10 & 12

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# EOP-EP KIA# 1940012412 Knowledge of general operating crew responsibilities during emergency operations Importance: 4.0 4.3 Question: #73 The crew has entered EOP 3-E-O; Reactor Trip or Safety Injection and the CRS has directed the BOP Operator to implement 3-R0-1; "BOP Operator Actions during the Use of EOPs." Which of the following describes conditions which will result in the termination of the 3-R0-1 procedure?

a. The BOP Operator has completed or initiated all the actions of 3-R0-1 or if 3-E-O; Reactor Trip or Safety Injection has been exited.

b. The BOP Operator has completed or verified completed all the actions of 3-R0-1 and 3-E-O; Reactor Trip or Safety Injection has been exited.

c. The CRS has transitioned to 3-E-1, Loss of Reactor or Secondary Coolant and SI Initiation has been verified.

d. The CRS has transitioned to 3-ES-1.1; SI Termination or the CRS has transitioned to 3-ECA-0.0, Loss of All AC Power Answer: d Explanation I Justification

a. Incorrect. Plausible because the first statement is correct, completion of R0-1 is one exit point. However, 3-R0-1, step 16 Response Not Obtained directs the BOP Operator to continue with the procedure if 3-E-O has been exited.

b. Incorrect. Plausible because the first statement is correct, however actions only need to be initiated. However, 3-R0-1, step 16 Response Not Obtained directs the BOP Operator to continue with the procedure if 3-E-O has been exited.

c. Incorrect. Plausible because the student may believe that once diagnosis of the accident has been made, the transition procedure will continue any required verification steps.

d. Correct. See third bullet note prior to step 1 of 3-R0-1 which states that; "If CRS subsequently transitions to 3-ECA-0.0, Loss of All AC Power, then this procedure SHALL be terminated. Also see step 16 of 3-R0-1 to see that the procedure is terminated if the CRS has transitioned to 3-ES-1.1; SI Termination. Also see note prior to step 17.

Technical

References:

13LP-ILO-EOPEOO, 3-E-Or4, 3-R0-1 & Dev Doc OAP-012; EOP Users Guide Proposed References to be provided: None Learning Objective: Objectives 11.0 & 13.0

Question Source: New Question Cognitive Level: Knowledge 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# EOP-EP KIA# 0000742423 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations Importance: 3.4 4.4 Question: #7 4 Operators have responded to a design large break LOCA in containment. Operators have transitioned from E-0; Reactor Trip or Safety Injection to E-1; Loss of Reactor or Secondary Coolant. While in E-1, RWST Level lowers to less than 11.5 feet. At the same time that RWST Level lowers to less than 11.5 feet, an Orange Path Critical Safety Function for 3-FR-Z.1 ~

Response to Containment High Pressure is announced by the STA. There is no other Red or Orange Path identified. Which procedure transition is required?

a. Transition immediately to 3-FR-Z.1; Response to Containment High Pressure. If an Orange Path condition arises, suspend any E-Set procedure in progress and transition to the FRP required by the Orange Path.

b. Perform 3-FR.Z.1; Response to Containment High Pressure in parallel with ES-1.3; Transfer to Cold Leg Recirculation. Only Red Path conditions require suspension of any E-Set procedure in progress.

c. Transition immediately to ES-1.3; Transfer to Cold Leg Recirculation. FRPs shall not be implemented prior to the completion of transfer to cold leg recirculation.

d. Perform 3-FR.Z.1; Response to Containment High Pressure in parallel with E-1; Loss of Reactor or Secondary Coolant. The only E-Set procedure that has priority over an FRP is ECA-0.0; Loss of All AC Power.

Answer: c Explanation I Justification

a. Incorrect. Plausible because for most conditions this statement is true. However certain procedures take priority over FRPs because of their treatment of specific initiating events.

See step 4.4.8.2 of OAP-012; EOP Users Guide.

b. Incorrect. Plausible because the student may believe that only Red Path conditions require suspension of any E-Set procedures.

c. Correct. See step 4.4.8.2 of OAP-012; EOP Users Guide. The steps of ES-1.3; Transfer to Cold Leg Recirculation must be completed, even if challenges to a Critical Safety Function occur at this time, since these steps ensure core cooling.

d. Incorrect. Plausible because the student may not recognize that less than 11.5 feet in the RWST requires transition to ES-1.3; Transfer to Cold Leg Recirculation. The student may

also believe that only Red Path conditions require suspension of any E-Set procedures.

ECA-0.0; Loss of All AC Power is not the only E-Set procedure that takes priority over FRPs.

Technical

References:

13LP-ILO-EOPROU OAP-012; EOP Users Guide Proposed References to be provided: None Learning Objective: Objective 12.0 Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level: RO SRO Tier# 3 Group# EOP-EP KIA# 1940012430 Knowledge of events related to system operation/status that must be reported to internal organization or external agencies, such as the State, the NRC, or the transmission system operator Importance: 2.7 4.1 Question: #75 Given the following:

• Unit 3 is at 100% power

• The Unit 3 Shift Manager has just been notified by the Security Shift Supervisor that a hostile action is occurring within the Owner Controlled Area.

The Unit 3 Shift Manager confers with the Unit 2 Shift Manager and agrees that both units are affected. The Shift Manager classifies the event and declares an ALERT.

Prior to initiating the ALERT notification, the Security Shift Supervisor informs both Unit Shift Managers that a hostile action is now occurring in the Protected Area. The event has been reclassified and declared a Site Area Emergency.

As the Control Room Communicator, you will

a. Unit 2, disregard previous classification and continue notification with highest current classification. Follow-up notification SHALL include details of all conditions/emergency classifications.

b. Unit 3, continue notification, then state at the end the following; "Changes in plant conditions indicate a potential for escalating the Emergency Classification."

c. Unit 3, disregard previous classification and continue notification with highest current classification. Follow-up notification SHALL include details of all conditions/emergency classifications.

d. Unit 2, continue notification, then state at the end the following; "Changes in plant conditions indicate a potential for escalating the Emergency Classification."

Answer: a Explanation I Justification

a. Correct. IP-EP-210 states that both Shift Managers shall confer with each other for any event of condition which may affect both Units such as security or natural events. If it is agreed that both units are affected, then the Unit 2 Shift Manager SHALL classify and declare the emergency. Step 1.4.F also states that if plant condition/emergency classification changes prior to initiating notification, disregard previous classification and continue with highest current classification.

b. Incorrect. Plausible because student may believe that since the Unit 3 Shift Manager was the first one notified, he would assume the role for classification and declaration of the event. Also plausible because IP-EP-210, step 1.4.G tells the communicator to continue with the notification if the plant condition/emergency classification changes while performing notification. The question stem states that the ALERT notification had not been initiated yet.

c. Incorrect. Plausible because student may believe that since the Unit 3 Shift Manager was the first one notified, he would assume the role for classification and declaration of the event. Second part of distractor is correct.

d. Incorrect. Plausible because IP-EP-210, step 1.4.G tells the communicator to continue with the notification if the plant condition/emergency classification changes while performing notification. The question stem states that the ALERT notification had not been initiated yet.

First part of distractor is correct.

Technical

References:

IP-EP-210 Proposed References to be provided: None Learning Objective: E-Plan Admin Question Source: New Question Cognitive Level: Comprehension 10CFR Part 55 Content: 55.41 (b) 10