ML120690560

Initial Exam 2012-301 Draft RO Written Exam

ML120690560
Person / Time
Site:	Sequoyah
Issue date:	03/08/2012
From:	NRC/RGN-II
To:	Tennessee Valley Authority
Gerald Laska
References
50-327/12-301, 50-328/12-301
Download: ML120690560 (162)

Text

1. 007 EA1.1O 001 Given the following:

- Unit 2 is starting up and is at 2% power with the Steam Dump Control System operating as follows:

- Steam Dump Mode Selector Switch (HS-1-103D): STEAM PRESS Steam Dump Pressure Controller (P10-1-33): AUTO

- A spurious Reactor Trip occurs, however Reactor Trip Breaker B fails to open.

Which ONE of the following identifies the steam generator (SG) pressure setpoint immediately after the reactor trip and the required actions to adjust SG pressure with PlC-I -33 in AUTO?

A. SG pressure corresponding to Tavg of 547°F ( 1005 psig)

Use the lever at the bottom of the P10-1-33 controller.

B. SG pressure corresponding to Tavg of 552°F (- 1047 psig)

Use the lever at the bottom of the P10-1-33 controller.

C SG pressure corresponding to Tavg of 547°F (-- 1005 psig)

Use the setpoint up/down pushbuttons on the P10-1-33 controller.

D. SG pressure corresponding to Tavg of 552°F ( 1047 psig)

Use the setpoint up/down pushbuttons on the P10-1-33 controller.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since first part of distractor is correct, however the controller is in Auto and so positioning the lever will not affect Stm Dump position.

B. Incorrect, Plausible if candidate thinks that following a Rx trip the setpoint of 547°F is in affect, however the Stm Dumps are in Stm Pressure mode so Tavg controller does not affect circuit. Also second part is not correct.

C. Correct, The steam generator pressures would be controlled at no load value of 54 7°F (normal setpoint of controller), and adjustments would be made by adjusting the automatic setpoint up/down as desired.

D. Incorrect, Plausible if candidate thinks that following a Rx trip the setpoint of 54 7°F is in affect, however the Stm Dumps are in Stm Pressure mode so Tavg controller does not affect circuit. Adjustments would be made by adjusting the automatic setpoint up/down as desired.

Sunday, November 20, 2011 3:38:03 PM

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 1 Tier: 1 Group 1 K/A: 007 Reactor Trip EA1.10 Ability to operate and monitor the following as they apply to a reactor trip:

SIG pressure Importance Rating: 3.7 I 3.7 10 CFR Part 55: 41.7 /45.5/45.6 IOCFR55.43.b: Not applicable KIA Match: Applicant must demonstrate the knowledge of how the steam generator pressures would be controlled following a reactor trip and knowledge of required actions if manual control was required Technical

Reference:

ES-0.1, Reactor Trip Response, Rev 32 0-SO-i -2, STEAM DUMP SYSTEM, Rev 0011 0-GO-4, Power Ascension from Less Than 5% Reactor Power to 30% Reactor Power, Rev 0062 Proposed references None to be provided:

Learning Objective: OPT200.SDCS obj. 7 Question Source:

New Modified Bank Bank X Question History: Question used on Feb 2010 NRC exam Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 2/20 10 Last 2 NRC?: YES Sunday, November 20, 2011 3:38:03 PM 2

2. 008 AK2.03 002 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- A leak equivalent to 1/2 inch in diameter has developed on the line connecting the pressurizer to the PORVs.

Which ONE of the following describes the long term response of the pressurizer pressure control and level control systems?

Pressurizer Pressure Controller Pressurizer Level Controller 1 -PIC-68-340A 1 -LIC-68-339 A. output will INCREASE output will INCREASE B. output will INCREASE output will DECREASE C. output will DECREASE output will INCREASE D1 output will DECREASE output will DECREASE Sunday, November 20, 2011 3:56:45 PM 2

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the output of the master controller increases as pressure goes high, not as pressure drops below setpoint to turn heaters on and attempt to raise RCS pressure. Plausible if candidate knows that the heaters need to be turned on but confuses the direction of the change in the output of the master pressure controller or believes that the heaters would be able to maintain system pressure above the trip setpoint. Also, because a LOCA of this size, which is not in the vapor space, would result in the level controller output increasing to raise charging flow.

B. Incorrect, Plausible because the output of the master controller increases as pressure goes high, not as pressure drops below setpoint to turn heaters on and attempt to raise RCS pressure and other controllers do increase their output to lower the process variable Plausible for the applicant to know that the heaters need to be turned on but confuses the direction of the change in the output of the master pressure controller. Also plausible because the change in the output of the level controller would be to decrease.

C. Incorrect, Plausible because the output of the master controller decreasing as the pressure drops below setpoint to turn on heaters is correct, but by design, with a 1/2 in vapor space leak in the PZR the heaters would not be able to maintain system pressure above the trip setpoint. Also, because a LOCA of this size, which is not in the vapor space, would result in the level controller output increasing to raise charging flow.

D. Correct, The pressurizer pressure would be dropping due to the vapor space LOCA. As system pressure compared to setpoint pressure lowers, the output of the master controller will start dropping to turn on heaters. The drop in pressurizer pressure would allow the pressurizer level to expand and additional water to enter the pressurizer, which would cause the level controller to decrease the output to drop charging flow.

Sunday, November 20, 2011 3:56:45 PM 3

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 2 Tier: 1 Group 1 K/A: 008 Pressurizer (PZR) Vapor Space Accident AK2.03 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

Controllers and positioners Importance Rating: 2.5 / 2.4 IOCFRPart55: 41.7/45.7 IOCFR55.43.b: Not applicable KIA Match: This question matches the K/A by having the candidate determine how the pressuirzer master pressure controller and pressurizer level controller would respond to a vapor space LOCA.

Technical

Reference:

EPM-3 E-1 Proposed references None to be provided:

Learning Objective:

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/2011 NRC exam.

Comments:

Source: BANK Source If Bank: WBN Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Sunday, November 20, 2011 3:56:45 PM 4

3. 009 EK2.03 003 Given the following plant conditions:

- A reactor trip and safety injection has occurred.

- Off-site power is lost at the time of the trip.

- The crew has transitioned to E-1, Loss of Reactor or Secondary Coolant.

- Step 3 of E-1 directs the crew to maintain between 10% and 50% narrow range level in all intact S/Gs.

Which ONE of the the following is the reason for maintaining this level in the intact S/Gs for small break LOCA type accidents?

A. Provides a static head of water to prevent primary to secondary leakage.

B Ensures adequate S/G inventory to ensure a secondary heat sink.

C. Ensures water level is above the top of the U-tubes to prevent pressurizing the S/Gs.

D. Provides adequate S/G inventory to support restarting reactor coolant pumps.

DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible since according to EPM-3-E-1 basis document, this is the criteria for maintaining S/G level but for Large break LOCA events.

B. Correct, Per EPM-3-E- 1, Basis document for E- I Loss of Reactor or Secondary Coolant, the purpose of maintaining normal level or AFW flow to the S/Gs during a SBLOCA event is to ensure that a secondary heat sink is maintained.

C. Incorrect, Plausible since this is the reason for maintaining S/G level Th the intact S/Gs but for a Steam generator tube rupture event.

D. Incorrect, Plausible if the candidate thinks that normal level in a S/G is the criteria for starting a reactor coolant pump.

Monday, November 21, 2011 8:51:42 AM 3

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 3 Tier: 1 Group I KIA: 009 Small Break LOCA EK2.02 Knowledge of the interrelations between the small break LOCA and the following:

Steam Generators Importance Rating: 3.0 I 3.3 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate recall the purpose for maintaining level in the intact S/Gs during a SBLOCA event.

Technical

Reference:

EPM-3-E-1 rev 6 Proposed references None to be provided:

Learning Objective: 0PL271 E-1 Loss of Reactor or Secondary Coolant obj 5 Question Source:

New X Modified Bank Bank Question History: This question written for 12-01 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12-0 1 Last 2 NRC?: NO Monday, November 21, 2011 8:51:42 AM 4

4. 011 EK2.02 004 Given the following plant conditions:

- A large break LOCA has occurred on Unit 1.

- ES-i .3, Transfer to RHR Containment Sump, was completed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago.

- A loss of offsite power has just occurred.

Which ONE of the following describes a manual action required to be taken in response to this event?

A. STOP/PULL TO LOCK the Containment Spray pumps.

B. STOP/PULL TO LOCK Safety Injection Pumps.

C. STOP/PULL TO LOCK the RHR pumps.

Dv STOP/PULL TO LOCK the CCP5.

DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible since the alignment made during ES-1.3 would have the Containment Spray pumps being supplied from the RHR pump discharge and RHR pump will not restart following a blackout, however the SI pumps will be stripped when the blackout occurs, but will not restart since the SI signal was reset during performance of ES-1.3.

B. Incorrect, Plausible since the alignment made during ES-1.3 would have the SI pumps being supplied from the RHR pump discharge and RHR pump will not restart following a blackout, however the SI pumps will be stripped when the blackout occurs, but will not restart since the SI signal was reset during performance of ES- 1.3.

C. Incorrect, Plausible since the RHR pumps will be stripped when the blackout occurs, but will not restart since SI was reset during performance of ES-1.3. Also is candidate gets confused on direction of procedure as to whether to restart RHR or place in Pull-to-Lock.

D. Correct, During the performance of ES-1.3, the suction of the CCPs are aligned to the discharge of the RHR pumps which are taking a suction on the containment sump. The CCPs receive a blackout start signal, the RHR pumps do not. Thus the CCPs would be running with no suction source for the given plant conditions. The opeartors are directed to place the control switches in Pull-to-Lock until an RHR pump can be restarted.

Monday, November21, 2011 9:11:32 AM 4

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 4 Tier: 1 Group 1 KIA: 011 Large Break LOCA EK2.02 Knowledge of the interrelations between the following and Large Break LOCA:

Pumps Importance Rating: 2.6 / 2.7 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of how the ECCS pumps that have been realigned to support continued operation to mitigate a large break LOCA event are effected by a loss of power during the event.

Technical

Reference:

ES-I .3, Step 18 Proposed references None to be provided:

Learning Objective: 0PL271 ES-i .3, obj 5 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12-01 Last 2 NRC?: NO Monday, November21, 2011 9:11:32 AM 5

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam

5. 015 AK2.08 005 Given the following plant conditions:

- Unit I is operating at 100% power.

- Alarms indicate a loss of ALL CCS flow to the RCPs

- Seal injection flow rate to each RCP is 8 gpm.

Which ONE of the following identifies how the operation of the RCFs will be affected if the operators do not respond to the alarms?

A. The RCPs are designed to operate without CCS indefinitely.

B. The RCPs will experience seal failure.

C. The RCP stator windings will overheat.

D The RCP motor bearings will overheat.

DIS TRACTOR ANALYSIS:

A. incorrect, The RCP motors cannot operate without component cooling water because the motor bearings will overheat. Plausible to conclude that due to the seal injection flow, the loss of thermal barrier cooling would not be an issue.

B. Incorrect, The RCPs will NOT experience seal failure as long as seal injection flow is present, but plausible because seal damage would occur if the CCS were lost and seal flow was not present.

C. Incorrect, The RCP stator windings are not cooling by CCS. The motor coolers use ERCW to cool the air leaving the RCP motors. Plausible if the RCP motor cooling is confused with the RCP motor bearing cooling.

D. Correct, The loss of CCS cooling to the motor bearings will cause overheating of the motor bearings and damage to the RCP motor.

Sunday, November 27, 2011 2:44:41 PM 5

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 5 Tier: 1 Group 1 KIA: 015 RCP Malfunctions AK2.08 Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions and the following:

CCws Importance Rating: 2.6 I 2.6 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the KIA by testing the candidates knowledge of the interrelationship between CCWS (CCS) and the RCPs and the affect of lossing CCS flow to the RCPs.

Technical

Reference:

AOP-M.03 AOP-R.04 FSAR 5.5.1.2 1 -47W859-2 Proposed references None to be provided:

Learning Objective: OPL271AOP-M.03 obj 13 OPT200.RCP obj 4.c Question Source:

New Modified Bank Bank X Question History: SQN NRC EXAM 1/2008, SEQ Bank AOP-M.03-B.4 8 Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Sunday, November 27, 2011 2:44:41 PM 6

6.022 AA1.03 006 Given the following plant conditions:

AT 10:00, Unit 1 plant conditions were as follows:

- 100% steady state power.

- The following parameters are observed:

- Letdown flow: 75 gpm

- Total charging flow: 100 gpm

- Seal Injection flow: 10 gpm per RCP

- Seal return flow: 3 gpm per RCP At 12:00, Unit 1 plant conditions were as follows:

- 100% steady state power.

- 1-FCV-62-93, Charging Flow Control Valve, has failed closed.

- Normal letdown cannot be established.

- Excess letdown has been placed in service.

- The following parameters are observed:

- Excess Letdown flow: at maximum capacity of heat exchanger: 25 gpm

- Total charging flow: 40 gpm

- Seal injection flow: 10 gpm per RCP

- Seal return flow: 3 gpm per RCP Which ONE of the following identifies the trend in Pressurizer level at 10:00 and 12:00?

Pressurizer level at 10:00 was _(1) and pressurizer level at 12:00 was _(2)_.

(1) (2)

A Increasing Increasing B. Increasing Decreasing C. Decreasing Increasing D. Decreasing Decreasing Monday, November 21, 2011 9:32:56 AM 6

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DISTRA CTOR ANAL YSIS:

A. Correct, With normal charging and letdown in service, letdown flow + seal return flow charging flow + seal injection flow. Letdown flow is given at 75 gpm. Seal return flow is given at 3 gpm/RCP, 12 gpm total. Seal injection flow is given at 10 gpm/RCP, 40 gpm total. So, in order to have a constant pressurizer level, charging flow must equal 47 gpm. Since charging flow is greater than 47 gpm, pressurizer level must rise. Excess letdown flow rate is 25 gpm. Seal return flow is given at 3gpm/RCP, 12 gpm total. So, in order to have a constant pressurizer level, total seal injection flow must be equal to 37 gpm. Seal injection flow is given at 10 gpm/RCP, 40 gpm total. Since seal injection flow is greater than 37 gpm, pressurizer level must rise.

B. Incorrect, Correct trend at 10:00 (see above). Incorrect trend at 12:00. Plausible if applicant does not understand mass balance with excess letdown in service.

C. Incorrect, Incorrect trend at 10:00 (see above). Plausible if applicant does not understand mass balance with normal letdown and charging in service. Correct trend at 12:00.

D. Incorrect, Incorrect trend at 10:00 (see above). Plausible if applicant does not understand mass balance with normal letdown and charging in service. Incorrect trend at 12:00. Plausible if applicant does not understand mass balance with excess letdown in service.

Monday, November 21, 2011 9:32:56 AM 7

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 6 Tier: 1 Group 1 KIA: 022 Loss of Reactor Coolant Makeup AA1 .03 Ability to operate and/or monitor the following as they apply to the Loss of Reactor Coolant Makeup:

Pressurizer Level Trend Importance Rating: 3.2 / 3.2 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the KJA by having the candidate evaluate the data given and determine the trend in Pressurizer level with and without normal charging flow, since normal charging is isolated during times when excess letdown is in service.

Technical

Reference:

1 ,2-SO-62-1 Proposed references None to be provided:

Learning Objective: OPL271AOP-M.09, obj. 3 Question Source:

New Modified Bank Bank X Question History: Original question from WBN bank.

Comments:

Source: BANK Source If Bank: WBN Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November 21, 2011 9:32:56 AM 8

7. 025 AA2.01 007 Given the following plant conditions:

- Unit 2 RCS has been drained to 695lO during a refueling outage, per 0-GO-13, Reactor Coolant System Drain and Fill Operations.

- The reactor vessel head is in place but bolts are NOT tightened.

- Containment sump level starts increasing.

- Annunciator RCS MID LOOP LEVEL LOW (M6-C, A7) alarms.

- Running RHR pump amps are observed cycling between 20 and 40 amps.

Which ONE of the following identifies the expected operator procedural response?

A. Enter AOP-R.02, Shutdown LOCA.

B. Enter AOP-M.04, Refueling Malfunctions.

C Enter AOP-R.03, RHR System Malfunction.

D. Enter AOP-R.05, RCS Leak and Leak Source Identification.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate determines that the flucuating amps indicates a Loss of RCS coolant is occurring and determines that AOP-R.02, Shutdown LOCA would be the correct procedure to enter. However, AOP-R.2 is not the correct procedure to mitigate the event due to plant being in Mode 6.

B. Incorrect, Plausible if the candidate determines that an RCS leak has developed and being in MODE 6. Also there is guidance in AOP-M.04 for Reactor Vessel Seal leaks. This guidance would not apply for these conditions.

C. Correct, The flucuating amps on the RHR pump indicate that the pump is cavitating thus the level is too low. AOP-R.03 applies due to the reduced inventory condition (RCS @ 696 & the reactor vessel head in place but bolts are NOT tightened).

D. Incorrect, Plausible since AOP-R.05 may assist in identifying the leak, it implements NO strategy to restore the lost shutdown cooling (Running RHR pump amps, discharge pressure, and flow fluctuating/running RHR pump trips).

AOP-R. 03 applies due to the reduced inventory condition (RCS @ 696 & the reactor vessel head in place but bolts are NOT tightened).

Monday, November 21, 2011 9:39:00 AM 7

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 7 Tier: 1 Group 1 KIA: 025 Loss of RHR System AA2.01 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

Proper amperage of running LPI/decay heat removal/RHR pumps.

Importance Rating: 2.7 I 2.9 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the KJA by testing the candidates knowledge of the expected relative value of RHR pump amps for normal operation, and when that value is not observed determine the correct course of action to mitigate the event.

Technical

Reference:

AOP-R.02, AOP-R.03, AOP-R.05, Proposed references None to be provided:

Learning Objective: OPL271AOP-R.03; Obj. 2 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments: Modified distractor B to remove subset issue. Reordered distractors due to pschometric issues.

Source: BANK MOD Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November 21, 2011 9:39:00 AM 8

8. 027 AG2.1.30 008 Given the following plant conditions,

- The Main Control Room has been abandoned.

- The transfer switches per AOP-C.04, Shutdown from the Auxiliary Control Room, have been placed in the AUX position.

- Aux Control Room alarm PS-68-336AC RCS PRZR PRESS HI has just been received.

Which ONE of the following identifies the lowest RCS pressure when the alarm is received and the action taken in response to the alarm?

RCS Pressure Action A 2310 psi 9 Ensure PZR Backup heaters 1 A-A & 1 B-B OFF at 6.9 Ky Shutdown Boards lA-A & lB-B B. 2310 psig Verify PZR PORVs 1-PCV-68-334 & 340 are going open on 1-L-10 C. 2335 psig Ensure PZR Backup heaters IA-A & I B-B OFF at 6.9 Ky Shutdown Boards 1 A-A & I B-B D. 2335 psig Ensure PZR PORVs 1-PCV-68-334 & 340 are going open on 1-L-10 DISTRACTOR ANAL YSIS:

A. Correct, The high pressure alarm setpoint is 2310 psig. This would indicate that the normal pressure control method of the backup heaters IA-A & lB-B not functioning correctly, which is to cycle between 2210 and 2250 while aligned to the Aux SD panel. Also the guidance would be to ensure that the heaters are off by observing their local control indications which is on 6.9 Ky SD board panels not the Aux SD board.

B. Incorrect, Plausible since the setpoint is correct, however the controls for the PZR heaters are located on the 6.9 Ky SD boards not the Aux Control panel.

C. Incorrect, Plausible if the candidate thinks that the high pressure alarm comes in at the same setpoint as the PORV actuation. Also PORVs are controlled from the Aux SD panel not the 6.9 Ky switchboard controls.

D. Incorrect, Plausible if the candidate thinks that the high pressure alarm comes in at the same setpoint as the PORV actuation. Also the second part of the question is correct, the PORVS are controlled from the Aux SD panel.

Monday, November21, 2011 9:45:44 AM 8

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 8 Tier: 1 Group 1 KIA: 027 Pressurizer Pressure Control System Malfunction AG2.1.30 Ability to locate and operate components, including local controls Importance Rating: 4.4 I 4.0 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This quesUon matches the KA by having the candidate determine what action to take if a malfunction of the PZR pressure control system occurs when not in the main control room and where the controls are located to mitigate the malfunction.

Technical

Reference:

AOPC.04 rev 18 1-AR-L-10, D-5, rev 7 Proposed references None to be provided:

Learning Objective: OPL271AOP-C.04 obj 3 & 11 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November 21, 2011 9:45:44 AM 9

9. 029 EG2.2.37 009 Given the following plant conditions:

- Unit 1 is responding to an ATWS per FR-S.1, Nuclear Power Generation I ATWS.

- Power is lost to all of panel i-M-2.

Which ONE of the following would the CR0 utilize for indication of turbine stop valves being closed?

A. P-9 TURBINE TRIP! REACTOR TRIP annunciator on Reactor First Out panel.

B. TV-i through TV-4 CLOSED lights on the EHC Control panel i-XX-47-1 000.

C. Dispatch AUO to verify turbine stop valves closed on EHC local panel.

D All 4 status lights Turb. Stop Valves Closed lit on i-XX-55-6A.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate thought this alarm would indicate that the Turbine Stop valves are closed, however this is only true if during the initial conditions the turbine was above P-9 (50% power) but there is no indications in the stem as to what initial power was.

B. Incorrect, Plausible since this would be a normal means of verifying the stop valves closed, however panel 1-M-2 has lost power (given in the stem) thus this indication would not be available.

C. Incorrect, Plausible if the candidate thought that the AUOs would be dispatched to locally trip the turbine, however they are not dispatched to verify Turbine Stop valve position during an A TWS event.

D. Correct, Each Turbine Stop valve has a valve position limit switch which indicates if the associated Turbine Stop valve has closed or not. This does provide positive indication of the Turbine Stop valves during and Turbine trip event.

Monday, November21, 2011 9:58:29 AM 9

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 9 Tier: 1 Group 1 KIA: 029 ATWS G 2.2.37 Ability to determine operability and/or availability of safety related equipment.

Importance Rating: 3.6 I 4.6 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K)A by testing the candidates knowledge of the availability of alternate indications that can be used to verify/validate information that is used to determine the status of Reactor Protection system equipment as they are applied to an ATWS event.

Technical

Reference:

FR-S.1 Proposed references None to be provided:

Learning Objective: OPL271FR-S.1, Obj. 3 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments: Changed distractors A & B to increase plausibility Reordered distractors to correct pyshometric flaws Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November21, 2011 9:58:29AM 10

10. 038 EA1.07 010 Given the following plant conditions:

- A SGTR has occurred on Unit 1.

- Offsite power was lost on the trip.

- The crew is performing actions in E-3, Steam Generator Tube Rupture.

- While the OATC is depressurizing the RCS to the target pressure the following alarms on panel 1-XA-55-5A are received:

- LS-68-300A!B PRESSURIZER RELIEF TANK LEVEL HI-LOW

- TS-68-309 PRESSURIZER RELIEF TANK TEMP HIGH

- PS-68-301 PRESSURIZER RELIEF TANK PRESS HIGH Which ONE of the following identifies the required actions?

A. Immediately lineup to cool the PRT by spraying primary water.

B. Close the PORV and lineup Aux Spray to complete the depressurization.

C. Depressurize the PRT by opening the vent to waste gas vent header.

D Continue depressurizing the RCS using the PORV to the required pressure.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible since this is the action that would normally be performed if a high pressure condition in the PRT is achieved. However the need to depressurize the RCS overrides the high pressure condition in the PRT.

B. Incorrect, Plausible if the candidate gets confused as the need to use Aux Spray.

However the use of Aux spray would only be used if the PORV was not available.

In this case the operators would continue to use the PORV.

C. Incorrect, Plausible since this is an action that could be performed during normal plant operation for high pressure in the PRT, however for the conditions stated in stem, the operators are to continue to depressurize the RCS and not spend any time attempting to vent the PRT.

D. Correct, For conditions when the PZR PORV is used to depressurize the RCS to reduce the leak rate into the SG, the actions necessary to minimize the potential for offsite release are of greater priority than the potential for lifting a relief or rupturing the PRT. Thus even if the PRT rupture disk were ready to blow, the RCS depressuirization will continue, unless other parameters idenfitied in E-3 are met.

Monday, November21, 201110:09:29 AM 10

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 10 Tier: 1 Group 1 K!A: 038 Steam Generator Tube Rupture EA1 .07 Ability to operate or monitor the following as they apply to a SGTR:

PRT tank temperature, pressure, and setpoints.

Importance Rating: 2.5 / 2.6 10 CFR Part 55: 41.7 IOCFR55A3.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the interactions of the PRT with depressurizing the RCS during a SGTR event.

Technical

Reference:

EPM-3 for E-3, Steam Generator Tube Rupture E-3 caution step 20.

Proposed references None to be provided:

Learning Objective: 0PL271 E-3 obj 5 Question Source:

New X Modified Bank Bank Question History: New question developed for 1201 ILT exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November 21, 2011 10:09:29 AM 11

II. W/E12 EK3.2 011 Given the following plant conditions:

- A steam line break has occurred in the Unit I West Valve Vault Room.

- The crew was unable to isolate any of the SGs.

- ECA-2.1, Uncontrolled Depressurization of All Steam Generators, is in progress.

- The crew has taken action to minimize the plant cooldown.

- AFW flow to S/Gs 1, 2, 3 and 4 are currently at 50 gpm.

- RCS Hotleg temperatures are decreasing slowly.

- The following alarms are received:

- 1-XA-M3-C3, STM GEN #1 LEVEL LOW

- 1-XA-M3-C4, STM GEN #2 LEVEL LOW

- 1-XA-M3-C5, STM GEN #3 LEVEL LOW

- 1-XA-M3-C6, STM GEN #4 LEVEL LOW Which ONE of the following identifies the action and reason for the action, that is required in accordance with ECA-2.1?

A. Raise AFW flow to #1, #2, #3, and #4 SGs. The minimum NR level, per ECA-2.1, is 10%.

B. Raise AFW flow to #1, #2, #3, and #4 SGs. The minimum NR level, per ECA-2.1, is 25%.

C Maintain AFW flow at its current value, If Thot starts to rise, raise AFW flow to stabilize RCS temperature.

D. Maintain AFW flow at its current value. When 3 of 4 SGs are at the applicable setpoint, transition to FR-H.1, Response to Loss of Secondary Heat Sink.

Monday, November 21, 2011 10:32:14AM 11

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DIS TRA CTOR ANAL YSIS:

A. Incorrect, Plausible because even after throttling AFW to minimize RCS cooldown, even if levels are low, AFW remains throttled until Thot begins to rise. At that point, AFW is throttled just enough to stabilize temperature. Also 10% is the minimum level required in the SGs to maintain a secondary heat sink during other EOP mitigating actions.

B. Incorrect, Plausible because even after throttling AFW to minimize RCS cooldown, even if levels are low, AFW remains throttled until Thot begins to rise. At that point, AFW is throttled just enough to stabilize temperature. Also 25% is the minimum level required in the SGs during adverse containment condlitions to maintain a secondary heat sink during other EOP mitigating actions.

C. Correct, The actions directed in ECA-2. I has the crew throttle AFW to minimum value to limit the cooldown on the RCS. As long as T-hot temperatures are going down, adequate heat removal is being maintained.

D. Incorrect, Plausible since this would normally be entry conditions for FR-H. 1, however since this is an operator induced reduction in AFW flow, FR-H. I actions would not apply.

Monday, November21, 2011 10:32:14AM 12

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 11 Tier: 1 Group 1 KIA: W/E12 Steam Line Rupture Excessive Heat Transfer EK3.2 Knowledge of the reasons for the following responses as they apply to the (Uncontrolled Depressurization of all Steam Generators):

Normal, abnormal, and emergency operating procedures associated with (Uncontrolled Depressurization of all Steam Generators)

Importance Rating: 3.3 I 3.9 10 CFR Part 55: 41.5, 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the reason for actions that are performed to mitigate an uncontrolled heat extration event (ECA-2.1) occuring on all SGs.

Technical

Reference:

ECA-2.1, AR M3 Proposed references None to be provided:

Learning Objective: OPL271ECA-2.1, obj 5 Question Source:

New X Modified Bank Bank Question History:

Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12-0 1 Last 2 NRC?: NO Monday, November21, 2011 10:32:14 AM 13

12. 054AA2.03 012 Given the following plant conditions:

- Unit 2 is operating at 52% power.

- 2B MFP is in service.

- Due to excessive vibrations, the 2B MFP needs to be shutdown.

- Just as the 2A MFP has been reset and is ready to be placed in service the 2B MFP trips.

- The following SG NR levels are observed:

- #1 10%

- #2 15%

- #3 20%

- #4 18%

Immediately following the Reactor Trip, which ONE of the following identifies the AFW pump(s) that will be automatically started and the reason why?

A. The 2A MDAFW pump, ONLY, starts due to low level in #1 SG.

B Both MDAFW pumps, ONLY, start due to the level in #1 SG.

C. Both MDAFW pumps and the TDAFW pump start due to trip of MFW pump.

D. Both MDAFW pumps and the TDAFW pump start due to Rx trip and Lo Tave.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible since low level in I SG will start the MDAFW pumps, however both are started not just one. Also plausible since the 2A MDAFW pump feeds SG 1 & 2 so it would seem logical that a low level on #1 SG would start the 2A MDAFW pump.

B. Correct, For the conditions in the stem both MDAFW pumps will start on low level (10.7%) in any SG. It takes low level in 2 SGs for auto start of the TDAFW pump.

C. Incorrect, Plausible since a trip of both MFW pumps is an auto start for all of the AFW pumps, however only I MFP is tripped (it is given in the stem that the A MFP is reset in preparation for start-up). Also a FWI signal will trip both of the MFW pumps however the RCS temperature has remained above the 550°F which would needed for the FWI logic, thus the A MFP would not have tripped.

D. Incorrect, Plausible if the candidate thinks that a FWI isolation would have been generated (which is normal after a trip with maximum AFW flow) however with RCS temperature at 552°F, the FWI would not have been generated.

Monday, November21, 2011 10:40:13AM 12

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 12 Tier: 1 Group 1 KIA: 054 Loss of Main Feedwater AA2.03 Ability to determine and interpret the following as they apply to the Loss of Main Feedwater:

Conditions and reasons for AFW pump startup.

Importance Rating: 4.1 / 4.2 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate determine the AFW system response to a Loss of Main Feedwater at reduced power level (52%) with only 1 Main Feedwater Pump in service. Also requires the candidate to determine why the AFW pumps, that did start, are running.

Technical

Reference:

Tl-28, Proposed references None to be provided:

Learning Objective: OPT200.AFW obj 8 Question Source:

New X Modified Bank Bank Question History: SQN bank question with the initial power level changed and different unit. Distractors have been revised to match wording changes in stem.

Comments:

Source: BANK MOD Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November21, 201110:40:13 AM 13

13. 055 EK1.02 013 Given the following plant conditions:

- The crew is in ECA-0.0, Loss of All AC Power.

- During the rapid depressurization of all intact SGs to reduce RCS pressure to 100-200 psig, an overshoot occurs.

- RCS is reduced to 50 psig before the depressurization is stabilized.

Which ONE of the following identifies the potential implication that could result from this overshoot in SG depressurization?

A Natural circulation may be impeded.

B. Unacceptable upper head voiding may occur.

C. The integrity of the S/G U-tubes may be challenged.

D. The integrity of the Reactor Vessel may be challenged.

DISTRA CTOR ANAL YSIS:

A. Correct, Nitrogen gas being discharged from the CLAs could accumulate in the RCS hotlegs and SG U-tubes, will inhibit natural circulation.

B. Incorrect, Plausible since upper head voiding will occur after pressurizer level is lost. However this is an acceptable consequence to minimize RCS inventory loss through RCP seal degradation.

C. Incorrect, Plausible since the rate of cooldown may be a concern; but final pressure/temperature evaluated.

D. Incorrect, Plausible since the rate of cooldown may be a concern; but final pressure/temperature evaluated. Pressurized thermal shock a concern when RCS remains pressurized, however the RCS is intentionally being de-pressurized.

Monday, November21, 201111:02:56 AM 13

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 13 Tier: 1 Group 1 KIA: 055 Station Blackout EK1 .02 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout:

Natural Circulation Cooling Importance Rating: 4.1 I 4.4 10 CFR Part 55: 41.8, 41.10 IOCFR55.43.b: not applicable KIA Match: This question matched the K/A by testing the candidates knowledge of the effect of lowering RCS pressure below the minimum value during Natural Circulation cool down.

Technical

Reference:

ECA-0.0, Loss of All AC Power EPM-3-ECA-0.0 Proposed references None to be provided:

Learning Objective: 0PL271 ECA-0.0, Obj 5 Question Source:

New Modified Bank Bank X Question History: SQN bank question used on 2008 NRC exam Comments: Changed distractor A to increase plausibility and reordered the distrators due to pychometric problems.

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November21, 201111:02:56 AM 14

14. 056 AG2.4.34 014 Given the following plant conditions:

- The I B-B DG is tagged for maintenance

- A loss of off-site power has occurred.

- Unit 1 is responding per AOP-P.01, Loss of Offsite Power.

- Unit 2 is responding per ECA-0.0, Loss of All AC Power.

- You are an extra UO and have been directed by U2 SRO to perform EA-250-1, Load Shed of Vital Loads After Station Blackout.

Which ONE of the following identifies when:

(1) the required actions are be completed and (2) which 125 Vdc loads are no longer available after the load shedding is complete?

A (1) 45 minutes (2) MCR annunciators, permissive lights and SSPS status lights.

B. (1) 45 minutes (2) DC air side seal oil pump.

C. (1) 90 minutes (2) MCR annunciators, permissive lights and SSPS status lights.

D. (1) 90 minutes (2) DC air side seal oil pump.

DISTRA CTOR ANAL YSIS:

A. Correct, ECA-0. 0 states that to conserve battery capacity, 125V battery board breakers with pink SBO tags must be opened within 45 minutes. Load shedding of 125 Vdc loads will result in loss of all MCR annunciators, permissive lights and SSPS status lights.

B. Incorrect, Plausible since the action to load shed is to be completed within 45 minutes, however the DC air side seal oil pump is removed from service using EA-250-2, but only after the generator hydrogen pressure has been reduced to less than 3 psig and is performed within 90 minutes.

C. Incorrect, Plausible since the DC air side seal oil pump is to be load shed within 90 minutes, thus the candidate may get confused as to when the load shed actions of EA-250-1 are to be performed.

D. Incorrect, Plausible since the DC air side seal oil pump is to be load shed within 90 minutes, however the time limit for load shed in accordance with EA-250-1 is within 45 minutes.

Monday, November21, 201111:09:10 AM 14

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 14 Tier: 1 Group 1 KIA: 056 Loss of Offsite Power AG2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Importance Rating: 4.2 I 4.1 10 CFR Part 55: 41.10 10CFR55.43b: not applicable K/A Match: This question matches the K/A since it requires actions to be performed which are outside the control room. As an extra UO, it would reasonable that the UO would be sent out to perform these actions.

Technical

Reference:

ECA-0.0 rev 23, EA-250-1 rev 14 Proposed references None to be provided:

Learning Objective: 0PL271 ECA-0.0 obj 5 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November21, 201111:09:10 AM 15

15. 057 AK3.O1 015 Given the following plant conditions:

- Unit I is at 100% RTP.

- MCR alarms received indicate that an electrical board has failed.

- All trip status lights on Panel 1-XX-55-5 (1-M-5) are OFF.

- The crew responds in accordance with the appropriate procedure.

Which ONE of the following identifies (1)the electrical board that failed and (2) the reason that manipulation of Auxiliary Feedwater (AFW) controls will be required?

L21 A 120 VAC Vital Instrument Board 1-I. To prevent overcooling caused by excessive AFW flow due to the LCVs failing open.

B. 120 VAC Vital Instrument Board 1-I. To allow the turbine driven AFW pump to be operated above minimum speed.

C. 120 VAC Vital Instrument Board 1-Il. To prevent overcooling caused by excessive AFW flow due to the LCVs failing open.

D. 120 VAC Vital Instrument Board 1-li. To allow the turbine driven AFW pump to be operated above minimum speed.

Monday, November 21, 201111:16:35 AM 15

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam DISTRA CTOR ANAL YSIS:

A. Correct, The board failure is correct, and the MD AFW pump lA-A LCVs do fail open causing excessive flow requiring the pump to be stopped in accordance with AOP-P. 03.

B. Incorrect, The board failure is correct, but the TD AFW pump is not affected by the 1-I board failure. Plausible because the TD AFW pump speed control would be affected if the failure was the 120 VAC Vital Instrument Board 1-Ill or if the 120 VAC Vital Instrument Board 1-I failure had occurred on Unit 2. If the pump was operating at minimum speed it would not develop enough pressure to be able to pump water to the steam generator C. Incorrect, The board failure is incorrect, but the MD AFW pump lB-B LCVs do fail open causing excessive flow requiring the pump to be stopped in accordance with AOP-P.03. Plausible because the excessive flow would occur due to the 120 VAC Vital Instrument Board 1-li failure and the candidate however the I-Il board failure would not caused all Status Lights to be off D. Incorrect, The board failure and the AFW manipulation are incorrect. Plausible because the failure was the 120 VAC Vital Instrument Board 1-Il would cause required manipulations of the AFW controls and a different board failure would require the TD AFW pump speed control power supply to be swapped to restore allow speed to be controlled above minimum speed in order to allow speed to rise sufficiently to pump water into the steam generators.

Monday, November21, 201111:16:35 AM 16

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 13 Tier: 1 Group 1 K/A: 057 Loss of Vital AC Instrument Bus AK3.0l Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus:

Actions contained in EOP for loss of vital AC electrical instrument bus.

Importance Rating: 4.1 / 4.4 10 CFR Part 55: 41.5, 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge fo the reasons for the action associated with AFW pump controls associated with a loss of a Vital Instrument Board.

Technical

Reference:

AOP-P.03, Loss of Unit 1 Vital Instrument Power Board, Rev. 21 Proposed references None to be provided:

Learning Objective: OPL271AOP-P.03 & P.04 obj. 3 Question Source:

New Modified Bank Bank Question History: SQN bank question used on Jan 2009 NRC exam Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November21, 201111:16:35 AM 17

16. 062 AA2.03 016 Given the following plant conditions:

- The crew entered AOP-M.O1, Loss of Essential Raw Cooling Water, due to high flow on ERCW Supply Header 2A.

- The leak was isolated by closing 2-FCV-67-223, Hdr 2A to Hdr 1 B lsol Valve.

- The crew completes the applicable section of AOP-M.O1.

Which one of the following describes the ERCW Supply Header to the CCS heat exchangers after the remaining actions of AOP-M.O1 are complete?

Reference Provided 2A1/2A2 Hx OBIIOB2 Hx IAIIIA2 Hx A. 2A ERCW Supply 2B ERCW Supply 1 B ERCW Supply Header Header Header B. 2A ERCW Supply ERCW Isolated lB ERCW Supply Header to Hx Header C 2A ERCW Supply 2B ERCW Supply ERCW Isolated Header Header to Hx D. ERCW Isolated 2B ERCW Supply lB ERCW Supply to Hx Header Header DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since closing 2-FCV-67-233 isolates the 2A header from the lB header. Closing this valve isolated the leak between the 2A header and 1-FCV-67-424, Hdr I B to Hdr 2A CCS Hx Iso Valve. However this section of the header contains the 1AI/1A2 Hx and is isolated.

B. Incorrect, Plausible since closing 2-FCV-67-233 isolates the 2A header from the lB header. Closing this valve isolated the leak between the 2A header and 1-FCV-67-424, Hdr I B to Hdr 2A CCS Hx Iso Valve. This section of the header contains the IA I/1A2 Hx and is isolated. Also the OBI/0B2 Hx is normally supplied by the 2 B ERCW supply header.

C. Correct, Closing 2-FCV-67-233 isolates the 2A header from the lB header.

Closing this valve isolated the leak between the 2A header and l-FCV-67-424, Hdr I B to Hdr 2A CCS Hx Iso Valve. This section of the header contains the 1A I/IA 2 Hx and is isolated.

D. Incorrect, Plausible since closing 2-FCV-67-233 isolates the 2A header from the lB header. Closing this valve isolated the leak between the 2A header and I-FCV-67-424, Hdr I B to Hdr 2A CCS Hx Iso Valve. This section of the header contains the IA I/IA 2 Hx and is isolated. Also the 2A I/2A2 Hx is still supplied by the 2A supply header.

Monday, November21, 201111:28:24 AM 16

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 16 Tier: 1 Group 1 KIA: 062 Loss of Nuclear Service Water (ERCW)

AA2.03 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:

The value lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition.

Importance Rating: 2.6 / 2.9 10 CFR Part 55: (43.5, 45.13)

IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the ERCW system and its interconnections with CCS and if part of the system is isolated, which heat exchangers would be affected.

Technical

Reference:

AOP-M.01 47W845-2 Proposed references One line diagram of ERCW supply headers to be provided:

Learning Objective: OPL271AOP-M.01, Obj 5 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November21, 201111:28:24 AM 17

17. 065 AK3.08 017 Given the following plant conditions:

- Both Units were operating at 100% power when a Station Blackout occurs.

- Both unit crews are reponding using ECA-0.0, Loss of All AC Power.

Which ONE of the following is the reason for the backup air supply for the TD AFW Pump LCVs and the action/condition required to align the backup supply?

A Allows the LCVs to be CLOSED during a Station Blackout event and will require manual alignment locally when needed.

B. Allows the LCVs to be CLOSED during a Station Blackout event and is automatically supplied when air pressure drops below regulator setpoint.

C. Allows the LCVs to be OPENED during a Station Blackout event and will require manual alignment locally when needed.

D. Allows the LCV5 to be OPENED during a Station Blackout event and is automatically supplied when air pressure drops below regulator setpoint.

DIS TRACTOR ANAL YSIS:

A. CORRECT, the backup supply is from high pressure air cylinders that allow the valves to be closed a limited number of times after the normal air pressure is lost during a station blackout and its use requires manual valve alignment in accordance with EA-3-4, Local Alignment of TD AFW LCV Backup Air Supply.

B. Incorrect, the backup supply is to allow the valves to be closed during a station blackout where the normal air is lost but it requires a manual valve to be opened to enable its use.

Plausible because its purpose is to allow the valve to be closed and there are regulators to maintain pressure to the LCVs at 75 psig when using the backup supply C. Incorrect, the backup supply is to allow the valves to be closed, not opened, during a station blackout where the normal air is lost. The manual alignment of the supply is correct. Plausible because other AFW LCVs do fall closed and the manual alignment is required to use the backup supply.

D. Incorrect, the backup supply is to allow the valves to be closed, not opened, during a station blackout where the normal air is lost and while there is a regulator, the backup supply is not automatically until it is manually aligned. Plausible because other AFW LCVs do fail closed and there are regulators to maintain pressure to the LCVs at 75 psig when using the backup supply.

Monday, November21, 201111:44:26 AM 17

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 17 Tier: 1 Group 1 KIA: 065 Loss of Instrument Air AK3.08 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air:

actions contained in EOP for loss of instrument air.

Importance Rating: 3.7 I 3.9 10 CFR Part 55: 41.5, 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the reasons for backup air supply for the TD AFW LCVs.

Technical

Reference:

1 ,2-47W848-1 2 R41 EA-3-4, Local Alignment of TD AFW LCV Backup Supply, Rev 4 FSAR amendment 20, Section 10.4.7.2.

Proposed references None to be provided:

Learning Objective: OPT200.AFW obj. 14 Question Source:

New X Modified Bank Bank Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November21, 201111:44:26 AM 18

18. W/E05 EK1.3 018 Given the following plant conditions:

- Unit 1 is responding to a Loss of Heat Sink per FR-Hi, Response to Loss of Secondary Heat Sink.

- All Steam Generator Wide Range levels are Off-Scale low.

- RCS temperature is approximately 588°F and rising slowly.

- Core Exit Thermocouples are 605°F and rising slowly.

Which ONE of the following describes the preferred method of initiating Auxiliary Feed flow for these conditions?

A. Feed at the highest possible rate to one S/G to preclude initiation of RCS Feed and Bleed.

B. Feed at the minimum required flow to prevent possible SG tube failures.

C Feed at the highest possible rate to one S/G to reestablish SG inventory and secondary heat sink.

D. Feed at the minimum required flow to establish a controllable cooldown rate and prevent RCS pressure from reaching the PORV setpoint.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible if the candidate did not recognize that with all SG wide range levels off-scale, the RCS would already be on Feed and Bleed.

B. Incorrect, Plausible since SG tube failures are the primary concern when initiating aux. feed, but for these conditions, restoration of I SG as soon as possible is the priority C. Correct, If RCS temp is rising with no inventory, AFW flow should be directed to one SG at the max rate in an attempt to recover heat sink. This minimizes the chance for multiple tube failures as well as the quickest way to recover at least I SG as heat sink. At this point, bleed and feed should already be initiated.

D. Incorrect, Plausible, since the operators normally are to limit RCS cooldown rate, however on a loss of heat sink, cooldown rate is not the priority. The RCS has already heated up. Loss of inventory is a concern due to potential tube failures, but addressed by feeding only I SG Monday, November 21, 2011 12:03:51 PM 18

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 18 Tier: 1 Group 1 KIA: W/E05 Inadequate Heat Transfer Loss of Secondary Heat Sink EK1 .3 Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink):

annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Secondary Heat Sink)

Importance Rating: 3.9 /4.1 IOCFRPart55: 41.8,41.10 IOCFR55.43b: not applicable KIA Match: This question matches the KIA by testing the candidates knowledge of the procedural actions needed to mitigate a Loss of Secondary Heat Sink and use that knowledge to assess the given plant conditions to determine the correct action.

Technical

Reference:

EPM-3 FR-H.1 Proposed references None to be provided:

Learning Objective: OPL271FR-H.1 Obj. 5 & 6 Question Source:

New Modified Bank Bank X Question History:

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November21, 2011 12:03:51 PM 19

19. 003 AK3.07 019 Given the following plant conditions:

- Unit 2 is at 12% reactor power during a plant startup.

- The OATC reports that Rod H4 has dropped into the core.

- The crew enters AOP-C.01, Rod Control System Malfunctions.

- While monitoring Tave the OATC reports RCS temperature at 540°F.

- PZR pressure is 2225 psig.

- The Unit Supervisor directs that the reactor be tripped.

Which ONE of the following explains why the reactor is tripped under these conditions?

A. The required DNB parameters are below the limit.

B. The Enthalpy Rise Hot Channel Factor ( EN AH ) value has been exceeded.

C. RCS temperature is below that required to ensure control rods will not bind.

D The Moderator Temperature Coefficient may be outside the initial conditions assumed in accident analysis.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible if the operators confuse the conditions necessary for DNB limit with those provided. With low temperature the plant would be further away from the limit, not higher temperature which would make the plant closer to the limit.

B. Incorrect, Plausible since this a concern when the plant is at normal 100% power, however this limit only applies when the unit is above 50% power, not at 12% as given in the stem.

C. Incorrect, Plausible if candidate gets the temperature drop mentioned in the cooldown procedure, that every 50°F cooldown if the rods are unlatched and inserted that the control rods are to be moved to determine that they are not bound.

In this question, the rods are both latched and withdrawn.

D. Correct, With RCS temperature <541 °F, the MTC may not be within its design value, and also most accident analysis is conducted using NOT and NOP as the starting point for analysis to demonstrate that the plant will stay in a safe condition.

Monday, November21, 2011 12:13:44 PM 19

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 19 Tier: 1 Group 2 K/A: 003 Dropped Control Rod AK3.07 Knowledge of the reasons for the following responses as they apply to the Dropped Control Rod:

Tech-Spec limits for Tave Importance Rating: 3.8 I 3.9 10 CFR Part 55: 41.5, 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate identify the reason for minimum temperature for criticality to allow continued operation.

Technical

Reference:

Tech Spec 3.1.1.4, RCS Temperaure Tech Spec 3.4.4, Pressurizer Proposed references None to be provided:

Learning Objective: 0PL271 AOP-C.01 obj 8 Question Source:

New X Modified Bank Bank Question History: New question developed for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November 21, 2011 12:13:44 PM 20

20. 024 AAI.10 020 Given the following plant conditions:

- Unit 1 tripped from 100% power.

- Crew is performing ES-0.1, Reactor Trip Response.

- Three Control Rods are determined to remain greater than 12 steps out of the core.

- The OATC is directed to perform EA-68-4, Emergency Boration.

- The OATC reports that neither Boric Acid Transfer pump will shift to fast speed.

Which ONE of the following identifies conditions that meet the requirements of EA-68-4 for Emergency Boration?

A. Charging Pumps aligned to the RWST with normal boration flow path aligned and 1-FI-62-139, Boric Acid Flow to Blender, indicating 45 gpm.

By Charging Pumps aligned to the RWST with normal charging flow path aligned and charging flow of 98 gpm.

C. Charging Pumps aligned to the VCT with Alternate Emergency Boration flow path aligned and 1-Fl-62-139, Boric Acid Flow to Blender, indicating 30 gpm.

D. Charging Pumps aligned to the VCT with CCPIT flow path aligned and charging flow of 93 gpm.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate gets confused on where the charging pumps are taking a suction. Since the required flow of borIc acid from the BAT is at least 35 gpm, which goes directly into the suction of the charging pumps at the outlet to the VCT, however with the suction of the charging pumps aligned to the RWST, the required flow for emergency boration is at least 90 gpm.

B. Correct, In accordance with EA-68-4, with the charging pumps aligned to take a suction from the RWST with normal charging flow path (No SI present) the required amount of flow to establish emergency boration is 90 gpm. This will ensure that adequate flow to the RCS is established to re-establish required shutdown margin with 3 stuck rods.

C. Incorrect, Plausible since this is an alternate flow path which can be established for emergency boration, however the minimum flow rate is 35 gpm.

D. Incorrect, Plaulbie since this is a possible flow path, however with charging pump suction ailgned to the VCT, the required flow is 35 gpm boric acid flow not 90 gpm charging flow.

Monday, November21, 2011 12:29:30 PM 20

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 20 Tier: 1 Group 2 K/A: 024 Emergency Boration AA2.04 Ability to determine and interpret the following as they apply to the Emergency Boration:

Availability of RWST Importance Rating: 3.4 / 4.2 IOCFRPart55: 41.7/41.10 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by having the candidate determine, from the different flowpaths presented, if the available flow from the RWST is adequate for emergency boration.

Technical

Reference:

EA-68-4 rev 10 Proposed references None to be provided:

Learning Objective: OPL271ES-0.1 obj 5 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November 21, 2011 12:29:30 PM 21

21. 036 AK2.01 021 Given the following:

- Unit 1 is in Mode 6.

- An irradiated assembly is being transferred from an incore location to the upender.

- The Refueling SRO determines a leak has developed in the Reactor Cavity Seal.

- No assemblies are in the RCCA change fixture.

Which ONE of the following identifies the required action to be taken with the fuel assembly?

A. Place the fuel assembly in the upender and transport to the SEP.

B. Place the fuel assembly in the upender and lower the upender.

C. Place the fuel assembly in the RCCA change fixture.

Dv Place the fuel assembly in any core location.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since this is the direction given in the procedure if there is a fuel assembly in the RCCA change fixture, however as given in the stem the change fixture is empty, so the direction is to place the fuel assembly in any available core location.

B. Incorrect, Plausible since this location would prevent the fuel assembly from being exposed on loss of level, however with the transfer cart on the containment side the operators are prevented from closing the wafer valve which would seperate the refueling cavity from the spent fuel pool.

C. Incorrect, Plausible since this is the normal storage location for a fuel assemble which has been removed from the core and is awaiting transport to the Spent Fuel pool, however for conditions of loss of refueling cavity level the fuel assembly in directed to be placed in any available core location.

D. Correct, In accordance with A OP-M. 04, Refueling Malfunctions, if a fuel assembly is in the manipulator crane and a leak in the Refueling Cavity develops the crew is directed to place the irradiate fuel assemply in any available core location. If not the potential exists that if level in refueling cavity lowers to the level of reactor vessel any fuel assembly not in the core location could be exposed causing excessive radiation levels.

Monday, November21, 2011 1:20:17 PM 21

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 21 Tier: 1 Group 2 K/A: 036 Fuel Handling Accidents AK2.01 Knowledge of the interrelations between the Fuel Handling Incidents and the following:

Fuel Handling Equipment Importance Rating: 2.9 I 3.5 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by having the candidate determine the correct actions necessary to keep irratiated fuel assemblies safe during a refueling accident where a fuel assembly could be exposed thus causing excessive radiation levels.

Technical

Reference:

AOP-M.04, Refueling Malfunctions, rev 7 Proposed references None to be provided:

Learning Objective: OPL271AOP-M.04 obj. 10 Question Source:

New Modified Bank Bank X Question History: SQN bank question with the distractors reorganized due to pyschometric problems.

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November 21, 2011 1:20:17 PM 22

22. 059 AK3.02 022 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- Unit 2 is in Mode 6 for a refueling outage.

- Chemistry reports that sample results for the Turbine Building Sump indicate 1-131 concentration at 2.5 times the ODCM limit value for a liquid release.

- A backup sample indicates the release has been occurring for greater than 60 minutes.

- The SM declares an Unusual Event.

Which ONE of the following identifies (1) the reason for the declaration and (2) how the station sump pump discharge will be aligned?

Av (1) The effluent liquid has exceeded the Effluent Concentration Limits (ECL).

(2) to the LVWT pond with the L valve closed.

B. (1) The effluent liquid has exceeded the Effluent Concentration Limits (ECL).

(2) to the Yard Drainage (Cattail pond).

C. (1) The radiation monitor failed to isolate the release.

(2) to the LVWT pond with the L valve closed.

D. (1) The radiation monitor failed to isolate the release.

(2) to the Yard Drainage (Cattail pond).

DISTRA CTOR ANAL YSIS:

A. Correct, As stated in NP-REP Appendix B basis for liquid effluents under the Unusual Event classification. An unusual event entry for liquid releases is required when sample results indicate an ECL (1-13 1) exceeds 2 times the ODCM limit value for an unmonitored release of liquid radioactivity greater than 60 minutes in duration. Also the discharge would be aligned to LVWT pond rather than to the cattail pond (which is the normal path)

B. Incorrect, Plausible since the first part is correct. Also the if the candidate does not know that the discharge would be aligned to the LVWT pond when activity is present in the sump.

C. Incorrect, Plausible if the candidate confuses this rad monitor with other rad monitors associated with liquid release. The Turbine Building sump rad monitor does not provide any auto closure function. Also the second part is not correct.

D. Incorrect, Plausible if the candidate confuses this rad monitor with other rad monitors associated with liquid release. The Turbine Building sump rad monitor does not provide any auto closure function. Also the second part is correct.

Monday, November 21, 20111:45:22 PM 22

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 22 Tier: 1 Group 2 KIA: 059 Accidental Liquid RadWaste Release AK3.02 Knowledge of the reasons for the following responses as they apply to the Accidental Liquid RadWaste Release:

Implementation of E-plan Importance Rating: 3.2 I 3.5 10 CFR Part 55: 41.10 10CFR5543.b: not applicable KIA Match: This question matches the K/A by having the candidate identify the reason for the emergency plan declaration as it applies to a release of radioactivity through the turbine building sump and how the sump would be aligned for the condition of high activity.

Technical

Reference:

NP-REP Appendix B basis 0-AR-M12-A, E-1, rev 52 Proposed references None to be provided:

Learning Objective:

Question Source:

New X Modified Bank Bank Question History: New question developed for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November 21, 2011 1:45:22 PM 23

23. 069 AA1.01 023 Given the following plant conditions:

- Unit 1 is at 100% power.

- Containment entry is in progress for maintenance.

- The following annunciators are locked in.

- LWR PERS ACCESS OUTER DR LOCK

- LWR PERS ACCESS INNER DR LOCK

- UPRILWR AIR LOCK BREACH

- Containment to Annulus DP has lowered to -0.01 PSID.

- Containment pressure has rapidly dropped from 0.15 psig to 0.0 psig.

In addition to TS 3.0.3, which ONE of the following is a required TS entry AND required action?

A. Enter TS 3.6.1.4 INTERNAL PRESSURE and required to restore Containment to annulus DP within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B EnterTS 3.6.1.1 CONTAINMENT INTEGRITY and required to restore Containment intregrity within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Enter TS 3.6.1.3 CONTAINMENT AIR LOCKS and required to restore both Containment Access doors to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Enter TS 3.6.5.5 DIVIDER BARRIER PERSONNEL ACCESS DOORS AND EQUIPMENT HATCHES and required to close the doors within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

DIS TRA CTOR ANAL YSIS:

A. Incorrect, Plausible if student does not know LCO for Containment pressure with question indicating rapid drop in pressure. Containment pressure still within limits of Tech Spec 3.6.1.4 of-0. I and 0.3 psig.

B. Correct, In accordance with the Tech Spec for Containment Integrity, with both personnel air locks not closed there is a breach in integrity and due to Not meeting requirements of TS 3.6. 1.3 per surveillance requirements of TS 4.6.1.1, Tech Spec action statements apply.

C. Incorrect, Plausible if student does not know LCO for Containment Air Locks. There is no 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action for TS 3.6.1.3 associated with any doors being inoperable.

D. Incorrect, Plausible if student does not realize that this Tech Spec is for the Upper to Lower compartment hatches not the personnel access hatches mentioned in the stem.

Monday, November21, 20111:43:26 PM 23

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 23 Tier: 1 Group 2 K/A: 069 Loss of Containment Integrity AA1 .01 Ability to operate and/or monitor the following as they apply to the Loss of Containment Integrity:

Isolation valves, dampers, and electro-pneumatic devices.

Importance Rating: 3.5! 3.7 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of Containment Integrity and the equipment which if out of normal position, could adversely Containment Integrity and cause entry into the Tech Spec.

Technical

Reference:

0-AR-M12-C page 2,3,4, and 7.

Tech Specs 3.6.1.1, 3.6.1.3.

Proposed references None to be provided:

Learning Objective: OPTSTG200.CNMTSTRUCTURE, obj. 7 & 10 Question Source:

New Modified Bank Bank X Question History: SQN bank used on 2008 NRC exam Comments: Changed D distractor to make more plausible.

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Monday, November 21, 2011 1:43:26 PM 24

24. 074 EA2.03 024 Given the following plant conditions:

- Unit 1 was operating at 100% power.

- Reactor trip occurred due to a SBLOCA.

- FR-C.1, Inadequate Core Cooling, has been entered.

- The Unit Supervisor has directed the depressurization of all intact SGs to 100 psig using the condenser steam dumps.

• All MSIVs are open and the condenser is available.

• The steam dump controller (1-PlC-i -33) is in manual.

• The steam dump control mode selector switch is in the STM PRESS position and steam generator depressurization is underway.

• PZR pressure is 2000 psig.

• As the SG depressurization progresses, the steam flow automatically stops.

Steam flow stopped because of (1) or (2)

(1) (2)

A. Tavg < 540° F and no action PZR pressure < set point and no action has has been taken to defeat the been taken to block Low PZR Pressure SI.

Lo-Lo Tavg interlock B Tavg <540°F and no action Low steamline pressure and no action has has been taken to defeat the been taken to block Low steamline pressure Lo-Lo Tavg interlock SI.

C. Steam header pressure has Low steamline pressure and no action has dropped below the setpoint been taken to block Low steamline pressure on 1-PlC-i -33 SI.

D. Steam header pressure has PZR pressure < set point and no action has dropped below the setpoint been taken to block Low PZR Pressure SI.

on 1-PlC-i -33 Monday, November21, 2011 2:15:30 PM 24

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible if candidate doesnt recognize that as Tavg approaches 540 °F the Tavg Interlock must be defeated or the steam dumps close. This action was not performed. Low PZR pressure SI does not result in a main steam isolation.

B. Correct, As Tavg approaches 540 °F the Tavg Interlock must be defeated or the steam dumps will close. This action was not performed. Also, the Main Steamline Isolation due to exceeding the low steam pressure rate sensitive setpoint is active and could have resulted in an MSIV isolation if the rate of depressurization was excessive and the P- 11 interlock was not blocked.

C. Incorrect, Plausible however with Steam Dump controller in manual, the setpoint is not active. The Main Steamline Isolation due to exceeding the low steam pressure rate sensitive setpoint is active and could have resulted in an MSIV isolation if the rate of depressurization was excessive and the P-Il interlock was not blocked.

D. Incorrect, Plausible however the Steam Dump controller is in manual; therefore, the setpoint is not active. Low PZR pressure SI does not result in a main steam isolation.

Monday, November 21, 2011 2:15:30 PM 25

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 24 Tier: 1 Group 2 KIA: 074 Inadequate Core Cooling EA2.03 Ability to determine or interpret the following as they apply to a Inadequate Core Cooling:

Availability of turbine bypass valves for cooldown.

Importance Rating: 3.8 / 4.1 10 CFR Part 55: 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate determine whether the Turbine Bypass valves are available or would be available to continue the plant cooldown.

Technical

Reference:

FR-Cl 611-63 series drawings 611-1 series drawings Proposed references None to be provided:

Learning Objective: OPL271FR-C.1, Obj 6 Question Source:

New X Modified Bank Bank Question History:

Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12-0 1 Last 2 NRC?: NO Monday, November 21, 2011 2:15:30 PM 26

25. W/E01 EG2.1.7 025 Given the following plant conditions,

- Unit I was operating at 100% power.

- The TD AFW pump is out of service.

- A Pressurizer PORV fails open.

- Reactor Trip and SI have actuated.

- 1 B-B SD board trips and locks out.

- The associated PORV block valve was closed by the OATC.

- Total AFW flow to #1 & #2 SGs 460 gpm.

Which ONE of the following parameters would prevent SI from being terminated?

A. RCS pressure 1850 psig and stable.

B. SG #3 level at 8% and slowly lowering.

C. RCS subcooling 48°F and slowly lowering.

D Pressurizer level is 8% and slowly increasing.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate thinks that RCS pressure needs to be > the SI initiating setpoint to allow SI to be reset. However once the 60 sec timer has timed out SI can be reset no matter the RCS pressure. SI will only actuate again if the Reactor Trip Breakers are cycled. RCS pressure needs only to be stable or increasing.

B. Incorrect, Plausible if the candidate thinks that level in all SGs need to be 10% or being restored to 10%. With the TD AFW pump out of service and lB-B SD board deenergized there is no AFW available to SG #3 & #4. However the requirement is at least 1 SG with 10% level or> 440 gpm AFW flow which is being provided by A MD AFW pump but only to SG #1 & #2.

C. Incorrect, Plausible since RCS subcooling is one of the criteria to reset SI, however the setpoint is > 40 °F. There is no requirement as to whether it is increasing or decreasing.

D. Correct, For the conditions listed Pressurizer level less than 10%, even though it is increasing, would prevent SI termination. The minimum value is 10%

Monday, November21, 2011 2:24:03 PM 25

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 25 Tier: 1 Group 2 K/A: W/E01 SI Termination EG2.1 .7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Importance Rating: 4.4 I 4.7 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable KIA Match: This question matches the KIA by having the candidate determine if SI can be terminated by interpretting the readings and information given in the stem and comparing that information with the criteria stated in the plant procedures for SI Termination Technical

Reference:

E-0, rev 33 EPM-3, rev 14 Proposed references None to be provided:

Learning Objective: OPL271E-0, obj 3 & 5 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam.

Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Monday, November21, 2011 2:24:03 PM 26

26. W/E13 EK2.2 026 Given the following plant conditions:

- The crew entered FR-H.2, Steam Generator Overpressure, due to an overpressure condition on SG #2.

- SG #2 pressure is 1170 psig.

- SG #2 narrow range level is 82%.

Which ONE of the following describes the appropriate actions, in sequence, to mitigate this event in accordance with FR-H.2?

A. First Isolate AFW flow and then initiate SG Blowdown.

B. First Verify Feedwater Isolation and then initiate SG Blowdown.

C. First Isolate AFW flow and then attempt to dump steam from the affected SG.

D First Verify Feedwater Isolation and then attempt to dump steam from the affected SG.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible however AFW is isolated later in procedure if pressure cannot be brought under control by dumping steam. Additionally, SG blowdown is a later action.

B. Incorrect, Plausible since first action is correct, but SG blowdown is an action that would be performed in FR-H.3. Applicant may confuse these actions since the indicated level in the SG is 72% and think this is the cause of the high pressure.

C. Incorrect, Plausible however AFW is isolated later in procedure if pressure cannot be brought under control by dumping steam. Dumping steam from the affected SO is the correct action.

D. Correct, As directed by FR-H.2 first the crew would verify that a FWI has occurred then dump steam from the affected SG to reduce pressure. This would limit the reliance on SG safeties to control pressure.

Tuesday, November 22, 2011 8:53:28 AM 26

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 26 Tier: 1 Group 2 KIA: W/E13 Steam Generator Overpressure EK2.2 Knowledge of the interrelations between the (Steam Generator Overpressure) and the following:

Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these to the operation of the facility.

Importance Rating: 3.0 I 3.2 10 CFR Part 55: 41.8; 41.10 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the facilities design to remove decay heat following a Reactor Trip and the candidate determines that the normal system functions are not being performed and what action to take to mitigate the accident.

Technical

Reference:

FR-H.2 Proposed references None to be provided:

Learning Objective: OPL271FR-H.2, Obj. 5 & 7 Question Source:

New Modified Bank Bank X Question History: SQN bank question used on 2007 NRC exam (#65)

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 8:53:28 AM 27

27. W/E16EKI.3 027 Given the following plant conditions:

- Unit 1 is responding to a LOCA.

- ES-i .3, Transfer To RHR Containment Sump, was completed and containment recirculation cooling is in progress.

- The STA has identified a yellow path on containment radiation.

Which ONE of the following identifies a high level action performed in FR-Z.3, High Containment Radiation, to minimize the effects of the high containment radiation condition?

A. VERIFY Phase A isolation.

B. VERIFY ABGTS operation.

C VERIFY Containment Ventilation Isolation.

D. NOTIFY chem lab to sample the containment sump.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible since Phase A isolation should have occurred when SI was initiated and verified in service by E-1 procedure. However it is not verified again during FR-Z. 3.

B. Incorrect, Plausible since ABGTS should be in service during a high containment radiation condition, however, it would have been verified in service by E-1 procedure. It is not verified again during FR-Z. 3 C. Correct, The verification of CVI provides assurance that no containment penetrations are open that could provide a release path from the containment atmosphere. This verification does take place during the implementation of FR-Z.3 D. Incorrect, Plausible since the containment sump will be sampled and analyzed as part of post accident sampling procedures as directed by E-1. Post accident sampling is not directed by FR-Z.3.

Tuesday, November 22, 2011 8:56:17AM 27

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 27 Tier: 1 Group 2 KIA: W/E16 High Containment Radiation EK1 .3 Knowledge of the operational implications of the following concepts as they apply to the (High Containment Radiation):

Annuciators and conditions, indicating signals, and remedial actions associated with the (High Containment Radiation)

Importance Rating: 3.0 I 3.3 10 CFR Part 55: 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate identify the high level actions taken to mitigate the effects of high containment radidation during accident conditions.

Technical

Reference:

FR-Z.3 Proposed references None to be provided:

Learning Objective: 0PL271 FR-Z.3, Obj 4 & 5 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 8:56:17 AM 28

28. 003 A1.05 028 Given the following plant conditions:

- Unit 1 is at 30% power.

- #2 RCP trips on overcurrent.

Which ONE of the following identifies how the indicated level in #2 SG initially responds and the effect of RCP trip on reactor status?

SG #2 Level Reactor Status A. Decrease RCP trip generates an automatic reactor trip.

B Decrease Manual reactor trip is required due to RCP tripping.

C. Increase RCP trip generates an automatic reactor trip.

D. Increase Manual reactor trip is required due to RCP tripping.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since the first part is correct and if the candidate does not recognize that the reactor power is below the single loss of flow setpoint, they would assume that the reactor will trip.

B. Correct, The indicated level in #2 SG will initially decrease due to shrink from the loss of heat transfer due to the reduction in flow caused by the RCP tripping. With Reactor power less than P-8 (35% power) the reactor will need to be manually tripped since the plant is below the single loss of flow trip setpoint.

C. In correct, Plausible if the candidate thinks that the initial plant response is for the SG level to increase vs decrease, since the level will rise due to overfeeding of the SG. However the initial plant response will due to the shrink of SG due to loss of flow. Also if the candidate does not recognize that the reactor power is below the single loss of flow setpoint, they would assume that the reactor will trip.

D. Incorrect, Plausible if the candidate thinks that the initial plant response is for the SG level to increase vs decrease, since the level will rise due to overfeeding of the SG. However the initial plant response will due to the shrink of SG due to loss of flow. Also the second part is correct.

Tuesday, November 22, 2011 9:16:23 AM 28

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 28 Tier: 2 Group 1 KIA: 003 Reactor Coolant Pump Al .05 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including:

RCS flow Importance Rating: 3.4 / 3.5 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the effects of running and tripping RCPs and their affects on associated parameters.

Technical

Reference:

AOP-R.04, rev 24 Proposed references None to be provided:

Learning Objective: OPL271R.04, obj. 6 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam.

Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 9:16:23 AM 29

29. 003 A4.05 029 Given the following plant conditions:

- Unit I is at 20% power.

- Excess Letdown had been placed in service while normal letdown was unavailable.

- Operators then return normal letdown to service.

Which ONE of the following identifies the expected response of #1 and #2 RCP seal leakoff flows when Excess Letdown is removed from service?

A. #1 seal leakoff rises.

1. 2 seal leakoff rises.

B #1 seal leakoff rises.

1. 2 seal lea koff drops.

C. #1 seal leakoff drops

1. 2 seal leakoff rises.

D. #1 seal leakoff drops.

1. 2 seal leakoff drops.

DISTRACTOR ANAL YSIS:

A. Incorrect, Pisusible since the first part is correct and if the candidate is confused about the amount of water which would go up through the #2 seal, however the seal leakoff flow from the #2 seal would go down not up.

B. Correct, Excess letdown returns to VCT using the seal return line to the VCT.

When excess letdown is removed from service the backpressure on that line would go down, thus allowing an increase in flow from the #1 seal leakoff to the VCT.

Since more fluid is going to the VCT less would be going up to the #2 seal, thus its flow would go down.

C. Incorrect, Plausible if the candidate does not understand that a reduction in pressure on the seal return line caused by excess letdown being secured would allow #1 seal leakoff to increase not decrease. The second part is also not correct since less flow would go up through the #2 seal thus seal leakoff flow would go down not up.

D. Incorrect, Plausible if the candidate does not understand that a reduction in pressure on the seal return line caused by excess letdown being secured would allow #1 seal leakoff to increase not decrease. Also the second part is correct.

Tuesday, November 22, 2011 9:26:55 AM 29

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 29 Tier: 2 Group 1 K/A: 003 Reactor Coolant Pump A4.05 Ability to manually operate and/or monitor in the control room:

RCP seal leakage detection instrumentation Importance Rating: 3.1 I 3.0 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate predict the response of the control indicators for RCP seal leakoff flow during a plant evolution (placing excess letdown in service)

Technical

Reference:

1 -SO-62-6 Proposed references None to be provided:

Learning Objective: OPT200.CVCS obj. 1 & 6 Question Source:

New Modified Bank Bank X Question History: SQN bank question used on 2008 Audit exam Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 9:26:55 AM 30

30. 004 K6.31 030 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- 1-FCV-62-89, Charging Seal Water Flow Control Valve, is operating at 60% open.

- The RCP seal injection flows indicated in the Main Control Room are:

LOOP 1 = 8.4 gpm LOOP 2 = 8.3 gpm LOOP 3 = 9.8 gpm LOOP 4 = 9.4 gpm

- 1-FCV-62-89 suddenly drifts resulting in the following indications:

LOOP I = 5.4 gpm LOOP 2 = 5.3 gpm LOOP 3 = 6.8 gpm LOOP 4 = 6.4 gpm

- Assume NO adjustments have been made to 1-FCV-62-93, Charging Flow Control Valve.

Which ONE of following identifies both:

(1) the direction I -FCV-62-89 drifted and (2) the required operator action, if any?

A. (1) Further in the Closed direction; (2)Adjust 1-FCV-62-89 OPEN until all RCP seal injection flows are > 6.0 gpm.

B. (1) Further in the Closed direction; (2) No operator action required since seal injection flow to at least one RCP is > 6.0 gpm.

C (1) Further in the Open direction; (2) Adjust 1-FCV-62-89 CLOSED until all RCP seal injection flows are > 6.0 gpm.

D. (I) Further in the Open direction; (2) No operator action required since seal injection to at least one RCP is greater than 6.0 gpm.

Tuesday, November 22, 2011 9:38:17 AM 30

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate thinks that FCV-62-89 falling closed would make RCP seal flow decrease, however since RCP seal injection comes off upstream of this valve by falling closed RCP seal flow would go up not down. Thus taking manual control of valve and Opening would have no affect on seal injection flow. Also the minimum required RCP seal injection flow is 6.0 gpm per pump.

B. Incorrect, Plausible if the candidate thinks that FCV-62-89 failing closed would make RCP seal flow decrease, however since RCP seal injection comes off upstream of this valve, by falling closed RCP seal flow would go up, not down.

Thus taking manual control of valve and Opening would have no affect on seal injection flow. Also the minimum required RCP seal injection flow is 6.0 gpm per pump.

C. Correct, FCV-62-93 and FCV-62-89 are two inline manually controlled flow control valves that are used to regulate flow to the RCP seals. Since flow to the RCPseaIs comes off between these two valves by positioning the down stream valve (FCV-62-89) backpressure on the upstream valve is affected which will thus affect the RCP seal flow. With FCV-62-89 going open, it would decrease the backpressure on the charging pumps and cause a decrease in RCP seal flow. The required operator action is take manual control of 1-FCV-62-89 and close the valve to increae seal flow to a minimum of 6.0 gpm per pump.

D. Incorrect, Plausible since the first part of answer is correct the valve has failed open. However the required procedure action is to increase flow to all RCPs to greater than 6.0 gpm not to at least one RCP.

Tuesday, November 22, 2011 9:38:17 AM 31

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 30 Tier: 2 Group 1 K/A: 004 Chemical and Volume Control (CVCS)

K6.31 Knowledge of the effect of a loss or malfunction on the following CVCS components:

Seal injection system and limits on flow range Importance Rating: 3.1 I 3.5 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate determine the effect a failure of the seal injection flow control valve would have on seal flow and the required action if seal flow is less than the required amount.

Technical

Reference:

1-AR-M5-B, C-3 alarm response, 47W809-1 Proposed references None to be provided:

Learning Objective: OPT200.CVCS, Obj 3 Question Source:

New X Modified Bank Bank Question History: New question for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November22, 2011 9:38:17 AM 32

31. 005K1.11 031 Given the following plant conditions:

- Unit 2 is in Mode 6 with core off-load ready to commence.

- A reactor cavity seal failure has just occurred.

- The crew has entered AOP-M.04, Refueling Malfunctions.

- Cavity makeup from the RWST has been established in accordance with Appendix A, Filling Refueling Cavity From RWST.

Which ONE of the following identifies the alignment directed by AOP-M.04 for the RHR pumps and CCPs if the RWST level drops below 5%?

RHR pumps CCPs A. Place in P-T-L Place in P-T-L B. Place in P-T-L Align suction to VCT C. Align suction to Place in P-T-L RCS Loop 4 hot leg D Align suction to Align suction to VCT RCS Loop 4 hot leg DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since in other procedures such as AOP-R. 02, Shutdown LOCA, if RWST level drops to 8% the operators are directed to stop any pumps taking a suction from the RWST.

B. Incorrect, Plausible if the candidate gets confused on the direction given for the RHR pumps. In Shutdown LOCA procedure, the crew is to stop any pumps taking a suction from RWST, since this is the normal source of makeup to the RCS during refueling, the operator may determine that this would be the correct action to take.

Also aligning the CCPs to the VCT would be correct.

C. Incorrect, Plausible since the first part is correct. The CCP suction is to be aligned to the VCT as a source of makeup during refueling if the RWST is not available.

D. Correct, In accordance with AOP-M.04, Caution Failure to maintain RWST level greater than 5% may cause CCPs and RHR pumps to lose suction. Thus the direction given is realign pump suctions if the minimum level cannot be maintained.

The RHR pumps are realigned to take suction off the Hotleg and the CCPs are realigned to take suction off the VCT.

Tuesday, November 22, 2011 9:54:45 AM 31

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 31 Tier: 2 Group 1 KIA: 005 Residual Heat Removal (RHR) System Ki .11 Knowledge of the physical connections and/or cause-effect relationships between the RHRS and the following systems:

RWST Importance Rating: 3.5 / 3.6 IOCFRPart55: 41.2-41.9 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the interrelationship of the RWST (including minimum level) and RHR system.

Technical

Reference:

AOP-M.04 Proposed references None to be provided:

Learning Objective: OPL271M.04 obj. 7 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 9:54:45 AM 32

32. 006 K1.11 032 Given the following plant conditions:

- Unit I and Unit 2 are at 100% power.

- 2B-B CCS pump is in service.

- 2A-A CCS pump is in A-Auto.

- The 2B-B CCS pump trips.

- 30 seconds later the C-S CCS pump trips.

- The crew enters AOP-M.03, Loss of Component Cooling Water.

Which ONE of the following identifies the effect these failures would have on the A and B Trains of ECCS for each unit?

A. A Train of ECCS on Unit 2 would be unable to perform its design function.

B. B Train of ECCS on Unit 2 ONLY would be unable to perform its design function.

C. Both Trains of ECCS on Unit 2 would be unable to perform their design function.

D Train B of ECCS on Both units would be unable to perform their design function.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate did not recognize that the 2A-A pump would auto start and supply flow to the A train equipment (RHR heat exchanger) thus A train would remain capable of performing its design function.

B. Incorrect, Plausible if the candidate thinks that only B train of CCS flow is lost to Unit 2 only, however the C-S pump provides flow to the B train of ECCS equipment for both units.

C. Incorrect, Plausible if the candidate did not recognize that the 2A-A CCS pump would auto start and provide flow to the A train of ECCS equipment for Unit 2.

D. Correct, During normal plant operation the CCS system is aligned such that the A train of ECCS equipemt is supplied by one of two pumps which have an Auto start feature. However the B train equipment is aligned to be supplied by the C-S pump.

If the C-S pump would trip, then flow through the B train of RHR for each unit would be unavailable until a manual lineup can be made. Both units B train of ECCS would not be able to perform their design function due to the loss of flow throught the RHR heat exchanger.

Tuesday, November 22, 2011 9:59:52 AM 32

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 32 Tier: 2 Group 1 KIA: 006 Emergency Core Cooling System (ECCS)

K1 .11 Knowledge of the physical connections and/or cause-effect relationships between the ECCS and the following systems:

CCWS (CCS)

Importance Rating: 2.8 / 3.2 IOCFRPart55: 41.2-41.9 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candates knowledge of the cause-effect relationship of CCS system as it is aligned to supply the ECCS equipment for both units. The system design at SQN has both units B train of ECCS equipment supplied by the same pump.

Technical

Reference:

AOP-M.03 step 5, 9 Note, rev 12 Proposed references None to be provided:

Learning Objective: OPT200.CCS, Obj 6 Question Source:

New X Modified Bank Bank Question History:

Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 9:59:52 AM 33

33. 006 K6.03 033 Given the following plant conditions:

- A small break LOCA has occurred.

- The crew is in ES-i .2, Post LOCA Cooldown and Depressurization.

- One RCP is operating.

- One CCP is operating.

- RCS Pressure is 1300 psig and stable.

- RCS subcooling is 72°F and stable.

- Two SI Pumps are in service.

Which ONE of the following describes how the plant and subcooling margin respond when one of the running SI pumps trips?

Subcoolinci Margin Plant Response A Decreases RCS pressure decreases in response to reduced ECCS flow.

B. Decreases RCS break flow remains constant while ECCS flow is decreased.

C. Remains the same. Flow from the running SI pump increases, maintaining a balance with break flow.

D. Remains the same. RCS temperature and pressure decrease in response to the reduced ECCS flow.

DIS TRACTOR ANAL YSIS:

A. Correct, The balance of break flow and injection flow maintain RCS pressure at a constant value, thus subcool margin remains relatively constant during this flow balance stage of accident mitigation. If a SI pump trips, the injection flow will decrease thus break flow will be greater than injection flow and RCS pressure will drop until a new equilibrium is established. During this time the subcool margin will be reduced.

B. Incorrect, Plausible since the first part is correct, however if injection flow is reduced RCS pressure will be reduced, thus break flow, although greater than injection flow, will be reduced.

C. Incorrect, Plausible if the candidate thinks that the increase in flow from the running SI pump will make up for the injection flow of both pumps and maintain RCS pressure constant. With RCS pressure constant, Subcool margin would remain constant.

D. Incorrect, Plausible if the candidate thinks that the decrease in injection flow will be matched by a corresponding decrease in RCS temperature, thus keeping the Subcooling margin constant.

Tuesday, November 22, 2011 10:33:01 AM 33

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 33 Tier: 2 Group 1 KIA: 006 Emergency Core Cooling System (ECCS)

K6.03 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:

Safety Injection Pumps Importance Rating: 3.4 / 3.9 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the effects that a loss of a safety injection will on the ability of the ECCS to fulfill its function during an accident.

Technical

Reference:

EPM-3-ES-1 .2 ES-i .2 Proposed references None to be provided:

Learning Objective: 0PL271 ES-i .2, Obj 7 Question Source:

New Modified Bank Bank X Question History: SQN bank question developed from a HB Robinson U2 NRC exam Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 10:33:01 AM 34

34. 007 G2.1.30 034 Given the following plant conditions:

- The OATC on Unit 2 has been directed to lower the level in PRT using 2-SO-68-5, Pressurizer Relief Tank.

- The RCDT pump hand switches are aligned for normal operation.

- The OATC opens 2-FCV-68-310, PRT Drain to RCDT.

Which ONE of the following identifies how the B RCDT pump is stopped?

A. Automatically when the OATC closes 2-FCV-68-31 0.

B. Automatically when the level in the PRT reaches 50%.

C. Manually by the OATC when the level in the PRT is less than 70%.

D Manually by the AUO when directed by the OATC prior to closing 2-FCV-68-31 0.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since the B RCDT pump auto starts when 2-FCV-68-310 is opened from the MCR and the candidate could confuse the pump starts when the valve opens with the pump stop when the vave is closed. However this pump must be stopped manually from the local control panel.

B. Incorrect, Plausible if the candidate gets confused with the auto stop feature of the RCDT pump when level in the RCDT gets to its low llimit. Also 50% level is the lowest level allowed in the PRT.

C. Incorrect, Plausible if the candidate recalls that the normal level in the PRT is 70%

and does not recall that the RCDT pumps are controlled from the local panel 0-L-2.

0. Correct, In accordance with 2-SO-68-5, the B RCDT will start when 2-FCV-68-310 is opened from the control room if the local control switch is aligned for Pull-P-Auto (normal position). After the pump starts it must be locally stopped until the OATC closes 2-FCV-68-310, then the pump is placed back into the Pull-P-Auto position.

Tuesday, November22, 2011 10:47:11 AM 34

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 34 Tier: 2 Group 1 K/A: 007 Pressurizer Relief/Quench Tank G2.1 .30 Ability to locate and operate components, including local controls.

Importance Rating: 4.4 / 4.0 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the controls (in the main control room and local) associated with lowering the level in the Pressurizer Relief Tank.

Technical

Reference:

2-SO-68-5, rev 16 1,2-45N779 -17 Rev 21 1 ,2-45N779 -45 Rev 8 Proposed references None to be provided:

Learning Objective: OPT200.PZRPCS obj. 15 & 22 Question Source:

New X Modified Bank Bank Question History: New question written for the 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 10:47:11 AM 35

35. 008 K4.01 035 Given the following plant conditions:

- lA-A, 2B-B, and C-S CCS pumps are normally aligned and running.

- 1 B-B and 2A-A CCS pumps are aligned in Standby Auto.

- Unit I experiences a loss of both 6.9 kV Start Buses.

- 1 A-A and 1 B-B DIGs load normally on the 6.9 kV Shutdown Boards.

Which ONE of the following describes the response of the CCS system to these conditions?

A. All running CCS pumps will be load shed and will automatically start after the shutdown boards are energized.

B. lA-A and C-S CCS pumps will be load shed and will automatically start after the shutdown boards are energized.

C IA-A CCS pump will be load shed, lA-A and lB-B CCS pumps will automatically start after the shutdown boards are energized.

D. lA-A and C-S CCS pumps will be load shed, IA-A, I B-B and C-S CCS pumps will automatically start after the shutdown boards are energized DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate gets confused about Station Black out vs Loss of power to U-i Shutdown Boards. The C-S pump (one of the running pumps) and the U-2 CCS pumps will not be load shed.

B. Incorrect, Plausible since these two pumps are supplying Train A and Train B CCS flow on U-i. However C-S pump normally arigned to Unit 2. Thus it would not load shed and continue to run.

C. Correct, For this loss of power event, only the IA-A CCS pump will be load shed.

Then when the boards are re-energized, Both the IA-A and 18-B CCS pumps will be automatically started.

D. Incorrect, Plausible since C-S gets a start signal from U-i SI, however C-S wont load shed since it is normally supplied power from U-2.

Tuesday, November 22, 2011 10:57:04 AM 35

QUESTIONS REPORT for SQN JAN 2012 NRC RD Exam Question Number: 35 Tier: 2 Group 1 KIA: 008 Component Cooling Water K4.01 Knowledge of the CCWS design feature(s) and/or interlocks which provide for the following:

Automatic start of standby pump.

Importance Rating: 3.1 / 3.3 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candiates knowledge of the autostart features associated with the CCS pumps.

Technical

Reference:

1 ,2-47W611-70-1 Proposed references None to be provided:

Learning Objective: OPT200.CCS, obj 7 & 11 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 201110:57:04 AM 36

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam

36. 010 K5.01 036 Given the following conditions,

- Unit 2 is in Mode 5.

- The crew is preparing to draw a bubble in PZR per 0-GO-i, Unit Startup From Cold Shutdown to Hot Standby.

- RCS Tave is 140°F and stable.

- Pressurizer is in water solid condition.

- RCS Wide range pressure is 75 psig.

- The PZR bubble is going to be drawn at 150 psig, wide range pressure.

Which ONE of the following identifies when a pressurizer bubble starts forming?

A. 292-295°F B. 333-335°F C. 341-347°F D 353-358° F DISTRACTOR ANAL ISIS:

A. Incorrect, Plausble if candidate gets confused on the pressure at which the bubble is drawn. If they use the current RCS pressure, then 75 psig 15 psi = 60 psia Saturation for 60 psia = 292°F B. Incorrect, Plausible however RCS wide range pressure is 25 psig higher than Pressurizer pressure, thus 150 25 = 125 psig. Then if the candidate subtracts 15 psia vs add 15 psia, they would get 110 psia. Saturation temp for 110 psia = 334°F C. Incorrect, Plausible if candidate does subtract the 25 psi to compensate for the elevation difference of wide range pressure transmitters and PZR, but does not convert to psia. Saturation for 125 psia is between 34 1°F and 347°F.

D. Correct, In accordance with 0-GO-I, Note, RCS wide range pressure is 25 psig higher than Pressurizer pressure thus to draw a bubble at 150 psig wide range, the candidate must subtract the height difference thus 150 25 1-2 psig. Then convert to psia = 140 psia. Saturation temp for 140 psia 334°F Sunday, November 27, 2011 3:19:47 PM 36

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 36 Tier: 2 Group 1 KIA: 010 Pressurizer Pressure Control (PCS)

K5.01 Knowledge of the operational implications of the following concepts as they apply to the PZR PCS:

Determination of condition of fluid in PZR, using steam tables.

Importance Rating: 3.5 / 4.0 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the cadidate determine the condition of fluid (saturation temperature vs pressure) as a bubble is drawn in the PZR.

Technical

Reference:

Steam tables 0-GO-i, Proposed references None to be provided:

Learning Objective: OPTSTG.PZR-PRT obj 16 & 17 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 ILT NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Sunday, November27, 2011 3:19:47 PM 37

37. 012 K2.01 037 Given the following:

- Unit 1 isat 100% power.

- A loss of 125V DC Vital Battery Board I occurs.

Which ONE of the following describes the effect on the associated Reactor Trip Breaker?

A MCB indication is lost and the reactor trip breaker is NOT capable of tripping on a SHUNT trip.

B. MCB indication is lost and the reactor trip breaker trips OPEN due to loss of power to the SHUNT coil.

C. MCB indication remains lit. The reactor trip breaker is NOT capable of tripping on a SHUNT trip.

D. MCB indication remains lit. The reactor trip breaker trips OPEN due to loss of power to the SHUNT coil.

DISTRA CTOR ANAL YSIS:

A. Correct, 125V DC supplies control power to RTB indication and also shunt trip mechanism. Without this power, indication is lost and the shunt trip feature of the RTB is disabled.

B. Incorrect, The MCB indication is lost but shunt trip is required to be energized to actuate. Plausible because the lost of MCB indication is correct and a power loss to the UV coils would cause a trip.

C. Incorrect, MCB indication is lost, because control power for breaker indication is supplied by 125V DC. Plausible if the power supply to the shunt trip is not associated with the power supply to the MCB indication and the second part of the distractor is correct (the RTB is not capable of tripping on a shunt trip).

D. Incorrect, MCB indication is lost, because control power for breaker indication is supplied by 125V DC. Plausible if the power supply to the shunt trip is not associated with the power supply to the MCB indication and a power loss to the UV coils would cause a trip.

Tuesday, November 22, 2011 12:49:02 PM 37

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 37 Tier: 2 Group 1 K/A: 012 Reactor Protection System (RPS)

K2.01 Knowledge of the bus power supplies to the following:

RPS channels, components, and interconnections.

Importance Rating: 3.3 / 3.7 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the power supply to the breaker indications for RPS and the loss of power will effect the RPS signals to the shunt trip coils for the Reactor Trip breakers.

Technical

Reference:

1, 2-45N699-1 R9 AOP-P.02, rev 12 Proposed references None to be provided:

Learning Objective: OPL27IAOP-P.02 obj 11 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: WBN Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 12:49:02 PM 38

38. 012 K6.02 038 Given the following plant conditions:

- Unit I is at 45%.

- Bypass breaker BYB is racked in and closed.

- Electrical Maintenance is working on RTB.

Which ONE of the following identifies the effect that a loss of one of the 1 5v or 48v power supplies inside the Train A SSPS logic cabinet will have on plant operation?

A. Failure of a single I 5v power supply will cause a reactor trip, but failure of a single 48v power supply will NOT.

B. Failure of a single 48v power supply will cause a reactor trip, but failure of a single I 5v power supply will NOT.

C. A reactor trip will NOT occur due to a failure of a single 1 5v or single 48v power supply.

D A reactor trip will occur due to a failure of either a single I 5v or a single 48v power supply.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate thinks that only a loss of 15 V power supply will cause a General Warning and initiate a reactor trip.

B. Incorrect, Plausible if the candidate thinks that only a loss of 48V power supply will cause a General Warning and intiate a reactor trip.

C. Incorrect, Plausible if candidate does not realize that a General Warning is already being generated by the fact that the BYB is closed. Normally a failure of either power supply does not cause a trip only a General Warning signal.

D. Correct, a General Warning is in on the Train B due to the bypass breaker being closed. A failure of either of the redundant power supplies (1 5V or 4V) will generate a addidtional General Warning, which will cause a reactor tr(.

Tuesday, November 22, 2011 12:58:33 PM 38

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 38 Tier: 2 Group 1 KIA: 012 Reactor Protection System (RPS)

K6.02 Knowledge of the effect of a loss or malfunction of the following will have on the RPS:

Redundant channels Importance Rating: 2.9 / 3.1 10 CFR Part 55: 41.7 /45/7 IOCFR55.43.b: Not applicable KIA Match: This question matches the K/A by having the candidate identify how a failure of a redundant channel will affect the RPS when a reactor trip bypass breaker is closed.

Technical

Reference:

Proposed references None to be provided:

Learning Objective: OPT200.RPS obj 6 & 9 Question Source:

New Modified Bank Bank X Question History: SQN bank question written for class 10-02 Audit exam.

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12-0 1 Last 2 NRC?: NO Tuesday, November22, 2011 12:58:33 PM 39

39. 013 K2.01 039 Given the following plant conditions:

- Unit 1 has tripped due a loss of 120V AC Vital Instrument Power Board 1-I.

- The crew has implemented E-O, Reactor Trip or Safety Injection.

- The OATC has justed reported that PZR pressure transmitter 1-PT-68-334 (Channel II) failed LOW.

Which ONE of the following describes the plant response?

(Assume NO operator action)

A. SI master relays on both trains of SSPS would actuate AND both trains of ECCS equipment would start.

B SI master relays on both trains of SSPS would actuate AND only B train ECCS equipment would start.

C. Only the B train SSPS SI master relays would actuate AND both trains of ECCS equipment would start.

D. Only the B train SSPS SI master relays would actuate AND only B train ECCS equipment would start.

DIS TRA CTOR ANAL YSIS:

A. Incorrect, The Master Relays on both trains will have power. Train A from Channel Ill via an auctioneering circuit, however, with the 1-I AC vital Instrument Power Board deenergized (Channel 1), the slave relays that control the Train A equipment will not have a power supply. Plausible if the candidate mistakes the source of the power supply or thinks that the circuit that auctioneers power in the logic cabinet provides power to the slave relays.

B. Correct, Master Relays on both trains will have power. Train A from Channel Ill via an auctioneering circuit, however, with the 1-I AC vital Instrument Power Board deenergized, the slave relays that control the Train A equipment will not have power.

C. Incorrect, Master Relays on both trains will have power. Train A from Channel Ill via the auctioneering circuit, however, Channel I is the only power supply for the slave relays that control the Train A equipment. Plausible if the candidate mistakes the source of the power supply or thinks that the circuit that auctioneers power in the logic cabinet provides power to the slave relays instead of the master relays.

D. Incorrect, Master Relays on both trains will have power. Train A from Channel Ill via an auctioneering circuit, however, Channel I is the only power supply for the slave relays that control the Train A equipment. Plausible if the candidate mistakes the function of the circuit that auctioneers power in the logic cabinet.

Tuesday, November 22, 20111:06:56 PM 39

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 39 Tier: 2 Group 1 K1A: 013 Engineered Safety Features Actuation System (ESFAS)

K2.01 Knowledge of the bus power supplies to the following:

ESFAS/Safeguards equipment control Importance Rating: 3.6 I 3.8 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the power supplies for individual channels and the effect a loss of a power supply will have on the associated ESFAS channel signals.

Technical

Reference:

47W61 1-63-1 Rev 4, AOP-P.03 Revi 9, 0-SO-99-1 Atti, Proposed references None to be provided:

Learning Objective: OPT200.RPS obj. 4 & 6 Question Source:

New Modified Bank Bank X Question History: SQN bank question used on SQN NRC Exam 1/2008, Comments: Bank question RPS-B.9.A 002 with minor format &

wording change Source: BANK Source If Bank: SEQUOYAI-I BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 1/2008 Last 2 NRC?: NO Tuesday, November 22, 20111:06:56 PM 40

40. 013 K5.02 040 Given the following plant conditions:

- A Safety Injection occurred 30 minutes ago on Unit 1 due to a faulted SIG that is now depressurized.

- The crew has transitioned to ES-i .1, SI Termination.

- Reactor trip breaker A failed to open.

- Both Safety Injection Reset pushbuttons have been depressed.

- Both Safety Injection Pump control switches have been placed to STOP and returned to A-AUTO.

Which ONE of the following describes the status of Safety Injection Train A and Safety Injection Pump iA-A?

TrainA Pump IA-A A. RESET Running B NOT RESET Running C. RESET Stopped D. NOT RESET Stopped Tuesday, November 22, 20111:25:37 PM 40

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DIS TRACTOR ANAL YSIS:

A. Incorrect, SI train A will not be reset as the A reactor trip breaker will not make up P-4, which is necessary to reset SI train A Because Atrain SI did not reset, IA-A SI pump will restart after its control switch is returned to the A -A uto position.

The first part of this distracter is plausible because the train would reset if the trip breaker had opened. The second part of this distracter is plausible as the SI pump will stop whether Slis reset or not, but will only remain off if its train of SI had reset.

B. Correct, SI train A will not be reset as the A reactor trip breaker will not make up P-4, which is necessary to reset SI train A. Because A train SI did not reset, IA-A SI pump will restart after its control switch is returned to the A-Auto position.

C. Incorrect, SI train A will not be reset as the A reactor trip breaker will not make up P-4, which is necessary to reset SI train A Because A train SI did not reset, IA-A SI pump will restart after its control switch is returned to the A -A uto position.

The first part of this distracter is plausible because the train would reset if the trip breaker had opened. The second part of the distracter is correct.

D. Incorrect, SI train A will not be reset as the A reactor trip breaker will not make up P-4, which is necessary to reset SI train A Because train SI did not reset, IA-A SI pump will restart after its control switch is returned to the A-Auto position.

The first part of this distracter is plausible as it is correct. The second part of this distracter is plausible as the SI pump will stop whether SI is reset or not, but will only remain off if its train of SI had reset.

Tuesday, November 22, 20111:25:37 PM 41

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 36 Tier: 2 Group 1 KIA: 013 K5.02 Engineered Safety Features Actuation System Knowledge of the operational implications of the following concepts as they apply to the ESFAS:

Safety system logic and reliability Importance Rating: 2.9 I 3.3 10 CFR Part 55: 41.5 /45.7 IOCFR55.43.b: Not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the Safety system logic associated with Reactor trip breakers position (P-4 interlock) and its affect on the ability to reset ESFAS following an automatic actuation.

Technical

Reference:

ES-1.1, SI Termination, Revision 0017 1-47W611-63-1 SI System, R12 Proposed references None to be provided:

Learning Objective: OPL271ES-1.1 obj 5 OPT200.RPSobj6 i&6j Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN bank question 013 K5.02 038 used on SQN 1/2009 RETAKE exam Comments: SQN bank question 013 K5.02 038 (originally SQN Exam Bank ES-I .2-B.2A 001)

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 1:25:37 PM 42

41. 022 K2.02 041 Given the following plant conditions:

- Unit 2 is at 100% power.

- 6.9 kv Shutdown Board 2B-B trips on a differential fault.

- All D/Gs functioned as designed.

Which ONE of the following is a result of this condition?

A. Train B RHR pump will be available when the blackout sequencer times out.

B. Feedwater flow is increased due to more demand by the S/G level program.

C Available forced flow from the upper compartment coolers is reduced.

D. Unit runback will occur due to loss of HDT Pump.

DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible if the candidate does not recognize that loads are not available from a board that had a differential fault. Black Out X and Black Out Y relays will not pick up due to dead bus.

B. Incorrect, Plausible, since the MSR HP steam supply valves go closed, however this will cause a decrease in load (a reduction in steam demand) not an increase.

C. Correct, The fans associated with the upper compartment coolers that are fed from the associated reactor vent board are no longer available. With these fans not operating available forced air flow will be reduced.

D. Incorrect, Plausible if the candidate gets confused as to what is lost. The steam supply to the MSRs is lost thus a load reduction will occur, however the plant will not suffer a load runback.

Tuesday, November 22, 2011 1:45:43 PM 41

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 41 Tier: 2 Group 1 KIA: 022 Containment Cooling System K2.02 Knowledge of the bus power supplies to the following:

Chillers Importance Rating: 2.5 / 2.4 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the power supplies for the Containment Cooling Units.

Technical

Reference:

1,2-1 5E500-1 1 ,2-45N724-1 1 ,2-45N755-3 AOP-P.05/06 App T.

Proposed references None to be provided:

Learning Objective: OPT200CNMTCLG& PURGE obj 7 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 20111:45:43 PM 42

42. 025 K5.01 042 Which ONE of the following statements identifies both the maximum containment ice bed temperature and a potential effect of operating above the maximum temperature?

Maximum Temp. Potential Effect A. 20°F Exceeding 12 psig inside Containment during a LOCA B. 20°F Exceeding 12 psig inside Containment during a Steam Line Break 0 27°F Exceeding 12 psig inside Containment during a LOCA D. 27°F Exceeding 12 psig inside Containment during a Steam Line Break DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since 20°F is the maximum temperature by procedure for the optimal range of temperature for normal operation. The candidate may get confused between the optimal temperature and maximum Tech Spec temperature.

Also the second part is correcct.

B. Incorrect, Plausible since 20°F is the maximum temperature by procedure for the optimal range of temperature for normal operation. The candidate may get confused between the optimal temperature and maximum Tech Spec temperature.

Also the second part is not correct, although plausible since steam line breaks also release a lot of thermal energy and would result in high containment temperatures during an accident.

C. Correct, Per Tech Spec 3.6.5.1 the maximum ice bed temperature is 27°F. The entire ice condenser system is designed to limit the inside containment pressure to 12 psig for any LOCA accident.

D. Incorrect, Plausible since the first part is correct, however the design is to limit peak containment pressure during LOCA events. Also the second part is not correct, although plausible since steam line breaks also release a lot of thermal energy and would result in high containment temperatures during an accident.

Tuesday, November 22, 20111:59:09 PM 42

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 42 Tier: 2 Group 1 KIA: 025 Ice Condenser System K5.01 Knowledge of operational implications of the following concepts as they apply to the ice condenser system:

Relationships between pressure and temperature Importance Rating: 3.0 I 3.4 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the relationship of temperature of ice condenser bed and containment pressure during accident conditions.

Technical

Reference:

Tech Spec 3.6.5.1 SO-61-1 Proposed references None to be provided:

Learning Objective: OPT200.ICE, Obj 12 & 16 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last2 NRC?: NO Tuesday, November 22, 2011 1:59:09 PM 43

43. 026 A1.06 043 Given the following plant conditions:

- Unit 2 has experienced a Reactor Trip and SI due to a Steamline Break inside Containment.

- Phase B has actuated.

- The CR0 reports that neither 2A-A nor 2B-B CCS pump is running and they will NOT start.

Which ONE of the listed pumps, is expected to experience bearing failure with in 10 minutes?

A. 2B-B Centrifugal Charging Pump.

B. 2A-A Safety Injection System pump.

C. 2B-B Residual Heat Removal pump.

D 2A-A Containment Spray pump.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since CCPs are cooled by CCS however for SQN it has been determined that the CCPs could operate indefinitely without CCS.

B. Incorrect, Plausible since the SI pumps are cooled by CCS however for SQN it has been determined that the SI pumps could operate indefinitely without CCS.

C. Incorrect, Plausible since the RHR pumps are cooled by CCS however for SQN it has been determined that the RHR pumps could operate indefinitely without CCS.

D. Correct, As stated in a caution in AOP-M.03, Loss of Component Cooling Water, If any containment spray pump is running with NO CCS cooling, spray pump may experience bearing failure after 10 minutes.

Tuesday, November22, 2011 2:10:21 PM 43

QUES11ONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 43 Tier: 2 Group 1 KIA: 026 Containment Spray System Al .06 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including:

Containment spray pump cooling Importance Rating: 2.7 / 3.0 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates ability to recognize the flucuating amps on the running RHR pump and determine the correct course of action to take to mitigate a Loss of RHR event due to decreasing RCS level.

Technical

Reference:

AOP-M.03 Proposed references None to be provided:

Learning Objective: OPL271M.03 obj 6 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 2:10:21 PM 44

44. 039 A3.02 044 Given the following plant conditions:

- A plant cooldown is in progress.

- RCS pressure is 1850 psig.

- RCS temperature is 505°F.

- All required actions have been taken for the cooldown in accordance with plant procedures.

An event occurs:

- RCS pressure is 1700 psig and lowering at 10 psi per second.

- SG pressures are 700 psig and lowering at 25 psi per second.

- Containment pressure is 1.2 psig and rising.

Assuming NO operator action, which ONE of the following describes the ESF actuation status?

A. Safety Injection AND Main Steam Line Isolation have occurred.

B. Safety Injection has occurred; Main Steam Line Isolation has NOT.

C Main Steam Line Isolation has occurred; Safety Injection has NOT.

D. NEITHER Main Steam Isolation NOR Safety Injection have occurred.

DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible since this could occur, however for the given conditions SI is blocked, thus would not occur.

B. Incorrect, Plausible if the candidate gets the actuations confused, the actuations listed are in reverse order of how they would actually occur.

C. Correct, Below P-I I, Low Steam pressure SI is blocked. High negative rate MSLI at 100 psi per 5 seconds is reinstated. Also CNMT pressure is not yet high enough to cause SI (1.54 psig)

D. Incorrect, Plausible however the MSI would occur on rate, since SI is blocked.

Tuesday, November 22, 2011 2:47:37 PM 44

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 44 Tier: 2 Group 1 KIA: 039 Main and Reheat Steam System (MRSS)

A3.02 Ability to monitor automatic operation of the MRSS including:

Isolation of MRSS Importance Rating: 3.1 I 3.5 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the automatic isolation capabilities of the Main Steam System as the plant is cooled down from normal at power conditions.

Technical

Reference:

Tl-28 ATT 9 0-GO-7 Note pg 22 Proposed references None to be provided:

Learning Objective: OPT200.RPS, obj. 6 Question Source:

New X Modified Bank Bank Question History:

Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 4/2007 Last 2 NRC?:

Tuesday, November22, 2011 2:47:37 PM 45

45. 059 A3.04 045 Given the following plant conditions:

- Unit I is at 72% power.

- P-l-33A, Steam Header Pressure to DCS, has been bypassed for Maintenance.

- P-I-33B, Steam Header Pressure to DCS, suddenly ramps to a value that is 50 psi higher than current value.

Which ONE of the following describes the INITIAL effect this failure will have on 1A Main Feed Pump Turbine speed?

A. Increase B. Decrease C. Remains the same in Auto D. Remains the same in Manual DIS TRACTOR ANAL YSIS:

A. Correct, Based on the design of the DCS with one channel bypassed the logic goes to the average of the two remaining signals. With one of those signals going up (but remaining within band) the circuit will average to two signals and Increase the speed of MFP turbine.

B. Incorrect, Plausible if the candidate gets the logic correct to increase the signal but gets confused on whether the increase in steam pressure will require an increase of decrease in MFP speed to compensate. In this case MFP speed must increase to cause feed header pressure to increase over steam header pressure.

C. Incorrect, Plausible since the control circuit could bypass the failed channel and go the single element control and remain in Auto.

D. Incorrect, Plausible since the circuit could think that both channels are failed and transfer feed pump speed control to manual.

Tuesday, November 22, 2011 3:14:44 PM 45

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 45 Tier: 2 Group 1 KIA: 059 Main Feedwater A3.04 AbHity to monitor automatic operation of the MEW, including:

Turbine driven feed pump Importance Rating: 2.5 I 2.6 IOCFRPart55: 41.7 IOCFR55A3.b: not applicable KIA Match: This question matches the KJA by testing the candidates knowledge of the inputs to MFP turbine speed control and how those inputs are processed to control MFP turbine speed.

Technical

Reference:

OPT200.DCS Proposed references None to be provided:

Learning Objective: OPT200.DCS obj 3 & 4 Question Source:

New X Modified Bank Bank Question History: New question written for 2101 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 3:14:44 PM 46

46. 061 A2.07 046 Given the following conditions:

- Unit I was operating at 100% power when a reactor trip occurs.

- The UO reports that neither MD AFW pump will start.

- The UO also reports that the TD Trip/Throttle Valve Position is RED, however there is no indicated AFW flow to any SG.

Which ONE of the following completes the statement below?

The Turbine-Driven AFW Pump normal steam supply will automatically transfer to the alternate steam supply when discharge header pressure is (1) and assuming the TD AFW pump starts, the operators control SG level by (2)

A (1) <100 psig for >60 seconds.

(2) Manually control TD AFP speed to maintain level 10 to 50%

B. (1) <100 psig for >60 seconds.

(2) Monitoring automatic operation of TD LCVs maintaining level at 33%

C. (1) below SG pressure for> 4 seconds.

(2) Manually control TD AFP speed to maintain level 10 to 50%

D. (1) below SG pressure for> 4 seconds.

(2) Monitoring automatic operation of TD LCVs maintaining level at 33%

DISTRA CTOR ANALYSIS:

A. Correct, There two steam supply valves that direct main steam to the TDAFW pump, one off Loop #1 and one off Loop #4. Only one valve is open at a time. If TD AFW pump discharge is less than 100 psig for >60 secs the logic senses a failure of normal supply of main steam to the TD AFW pump to start the pump and will open the alternate steam supply from Loop #4. Also according to EA-3-1 and EA-3-8, when controlling SG levels using the TD AFW pump, the operators are directed to manually control turbine speed to regulate flow.

B. Incorrect, Plausible since first part is correct, however the second part is correct for the MD AFW pump LCVs. The candidate may get the logic for the MD AFW level control valves confused with the TD level control valves. The TD AFW LCVs are manually controlled open or closed to control flow.

C. Incorrect, Plausible since this is control logic for valves associated with AFW, however this logic is for the ERCW valves for the suction of MDAFW pumps, not the steam supply to the TD AFW pump. Also the second part is correct.

D. Incorrect, Plausible since this is control logic for valves associated with AFW, however this logic is for the ERCW valves for the suction of MDAFW pumps, not the steam supply to the TD AFW pump. Also the second part is not correct. this logic is for the MD AFW pumps not the TD.

Tuesday, November 22, 2011 3:27:56 PM 46

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 46 Tier: 2 Group 1 KIA: 061 Auxiliary I Emergency Feedwater System (AFW)

A2.07 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Air or MOV failure Importance Rating: 3.4 I 3.5 10 CFR Part 55: 41.5 10CFR5543.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the how the steam supply valves for the TD AFW pump are designed to provide alternate steam supply is case of a system failure to ensure TD AFW pump remains available for heat removal.

Technical

Reference:

O-47W611-1-1 ES-0.1 EA-3-1 EA-3-8 Proposed references None to be provided:

Learning Objective: OPT200.AFW, Obj 8.c, 14.d Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAII Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 3:27:57 PM 47

47. 062 K3.01 047 Given the following plant conditions:

- Unit 2 is in Mode 3 preparing for startup.

- Start Bus Selector switches are in Manual

- C CSST is inadvertently deenergized.

2B Diesel failsto start.

- RCS pressure is 2100 psig.

Which ONE of the following identifies the pressurizer heater groups that will automatically energize to control RCS pressure?

A Backup heater groups 2A-A, 2B-B, and 2C B. Backup heater groups 2A-A and control heater group 2D C. All backup heaters groups and the control heater group 2D D. Backup heaters groups 2A-A, 2B-B, and control heater group 2D DISTRACTOR ANALYSIS:

A. Correct, For the conditions in the stem, B shutdown board is not affected and 2A heater group is the only group that will come back on after blackout.

B. Incorrect, Plausible if the candidate does not recognize that B Shutdown Bd. did not lose power, thus this would not have an effect.

C. Incorrect, Plausible if the candidate does not recognize which PZR heater groups will have lost power, also control heater 2D will not come back on after blackout.

D. Incorrect, Plausible however this assumes B shutdown board lost power and 2D fed off A Shutdown Bd. and not B.

Tuesday, November 22, 2011 3:52:58 PM 47

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 47 Tier: 2 Group 1 KIA: 062 AC Electrical Distribution K3.O1 Knowledge of the effect that a loss or malfunction of the AC distribution system will have on the following:

Major system loads Importance Rating: 3.3 I 3.4 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of how a malfunction of one the offsite supply transformers will affect the electrical distribution system and its affect on major equipment (PZR heaters)

Technical

Reference:

2-47W611-68-3 Proposed references None to be provided:

Learning Objective: OPT200.PZRPCS, Obj. 6 OPT200.AC, Obj 6 OPT200.DG, Obj 4.a, & 8 Question Source:

New Modified Bank Bank X Question History: SQN bank question with distractors rearranged due to psychometric problems Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 3:52:58 PM 48

48. 063 K3.02 048 Given the following plant conditions:

- Unit 2 was operating at 85% power.

- Annunciator125V DC VITAL CHGR III FAIL/VITAL BAT III DISCHARGE is in alarm.

- Battery Board Ill voltage indicates 0 volts.

Assuming NO operator action, which ONE of the following describes the impact on Unit 2?

A. Control Air Compressors C and D will unload due to a loss of control power.

B. An increase in Reactor Power to 100% due to the S/G #3 Atmospheric Relief Valve failing open.

C All MSIVs fail closed due to loss of power to solenoid valves.

D. A high pressurizer level Reactor Trip because FCV-62-89, Charging Flow Control, fails open.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since a loss of a Vital Battery Board will affect the compressors, however it is VBB-l that will affect Compressors C & D, not VBB-llI.

B. Incorrect, Plausible since a loss of Vital Battery Board will affect operation of the SG atmospheric relief valves for Unit 2, however the SG atmospheric relief valves will fail closed.

C. Correct, In accordance with AOP-P.02, a loss of VBB-III will cause all 4 Unit 2 MSIVs to close due to loss of power to the solenoids associated with control air which is used to hold the MSIVs open.

D. Incorrect, Plausible since a loss of VBB-III (DC power) will cause FCV-62-89 to fail, however the valve fails closed not open.

Tuesday, November 22, 2011 4:18:12 PM 48

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 48 Tier: 2 Group 1 KIA: 063 DC Electrical Distribution K3.02 Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following:

Components using DC control power.

Importance Rating: 3.5 I 3.7 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the equipment that has DC control power (Vital Battery Board) and how a failure of Vital Battey Board Ill will affect the unit.

Technical

Reference:

AOP-P.02 Notes Proposed references None to be provided:

Learning Objective: OPL271AOP-P.02, Obj 11 Question Source:

New Modified Bank Bank X Question History: SQN bank question with reason for correct answer being changed.

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 4:18:12 PM 49

49. 064 G2.4.8 049 Given the following plant conditions:

- Unit I at 100% power.

- Panel M-1 B Annunciator Window 6900V SD BD IA-A FAILURE OR BUS UNDERVOLTAGE/OVERVOLTAGE alarms.

- The crew enters AOP-P.05, Loss Of Unit 1 Shutdown Boards.

- A reactor trip occurs on high PZR pressure.

Which ONE of the following identifies the allowed usage of AOP-P.05 after the Emergency Operating Procedure network is entered following the reactor trip?

Continued performance of AOP-P.05 is A. allowed after the crew enters ES-0.1, Reactor Trip Response because ES-0.1 is NOT an accident mitigation EOP.

B allowed after the crew enters ES-0.1, Reactor Trip Response because restoring power could have an impact on meeting the goals of the EOP.

C. NOT allowed until the crew exits ES-0.1, Reactor Trip Response because the procedure reader must remain dedicated to the EOP in effect until the EOPs are exited.

D. NOT allowed until the crew exits ES-0.1, Reactor Trip Response because actions taken in AOP-P.05 could degrade the performance of the EOP.

Tuesday, November22, 2011 4:34:32 PM 49

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DIS TRACTOR ANAL YSIS:

A. Incorrect, As identified in EPM-4, selected AOPs, such as loss of vital power can be implemented concurrently with the EOPs. While ES-O. us not an accident mitigation procedure, that is not reason for parallel implementation, it is because the loss of powerAOPs can have a significant impact on the ability of the EOP to achieve it goals. Plausible because the parallel implementation is correct but the reason for parallel implementation is not correct.

B. Correct, As identified in EPM-4, selected AOPs, such as loss of vital power can be implemented concurrently with the EOPs because the loss of power AOPs can have a significant impact on the ability of the EOP to achieve it goals.

C. Incorrect, EPM-4 provides that EOPs have priority over AOPs, and normally a dedicated procedure reader is utilized in the EOP network however, selected AOPs are allowed to be pet-formed concurrently with EOPs. Plausible if candidate fails to recognize that AOP-P.05 is allowed to be used while performing ES-O. 1.

D. Incorrect, EPM-4 provides that EOPs have priority over AOPs, however, selected AOPs are allowed to be performed concurrently with EOPs. Plausible if candidate falls to recognize that A OP-P. 05 is allowed to be used while performing ES-O. 1 and knows that the AOP actions can not be taken if the action would degrade the EOP performance.

Tuesday, November 22, 2011 4:34:32 PM 50

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 49 Tier: 2 Group 1 KIA: 064 Emergency Diesel Generator G2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs Importance Rating: 3.8 / 4.5 10 CFR Part 55: 41.10 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by having the candidate determine how an AOP to restore power to a SD board that a diesel failed to re-energize can be used in conjunction with the EOP network procedures.

Technical

Reference:

EPM-4 rev 20 AOP-P.05 rev 17 Proposed references None to be provided:

Learning Objective: 0PL271 EPM-4, Obj. 1 Question Source:

New Modified Bank X Bank Question History: original question 063 G2.4.8 from 2009 NRC exam Comments: Changed unit affected, shutdown board affected, and procedure affected Source: BANK MOD Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 4:34:33 PM 51

50. 064 A1.05 050 Given the following plant conditions:

- Unit I isat 100% power

- Diesel generator lA-A is being manually loaded from the control room per 0-SO-82-1, Diesel Generator 1 A-A, following a period of operation under light load.

- Outside temperature is 32° F.

- The outside AUO reports that temperature in EDG building is 75°F and lowering.

Which ONE of the following is the expected response to the lowering temperature?

A. As temperature drops below 72°F, the DG IA-A Electrical Board Room Exhaust fan will automatically cycle OFF.

B. As temperature drops below 70.3°F, 3 of the intake dampers would go closed.

C. If temperature reaches 68°F, the DG Bldg Corridor Air Intake Damper would go closed.

D If temperature reaches 68°F, the exhaust fan selected to P-Auto must be cycled OFF manually.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since the Electrical Board Room Exhaust fans are on occasion turned off during EDG operation, however they are turned off the prevent smoke from being pumped into the EDG room they do not have an automatic shutoff on lowering temperature in the EDG building.

B. Incorrect, Plausible if the candidate gets confused on the temperature limit for the ED/G to be operable with 3 dampers closed. This is an upper temperature limit for operability but does not cause the dampers to open or closed.

C. Incorrect, Plausible since this damper is directed by procedure to be closed, however it is closed due to conditions of high levels of exhaust fumes accumulating is DG building corridor not because the temperature inside the DG room is getting too cold.

D. Correct, Normally the exhaust fan which is selected to P-Auto will cycle on and off between 90°F (on) and 68°F (off), to ensure the diesel room temperature stays in its design value of 68°F to 90°F, however when the diesel is running the automatic shutoff is disabled, thus the fan must be manually turned off Tuesday, November 22, 2011 4:42:45 PM 50

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 50 Tier: 2 Group 1 K/A: 064 Emergency Diesel Generator Al .05 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including:

ED/G room temperature Importance Rating: 2.5 / 2.5 IOCFRPart55: 41.5 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates ability to predict the EDG ventilation system response to decreasing outside air temperature.

Technical

Reference:

0-SO-82-l, rev 36 Proposed references None to be provided:

Learning Objective: OPT200.DG obj. 2. e Question Source:

New X Modified Bank Bank Question History: Question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYA}{

Date: 12-01 Last 2 NRC?: NO Tuesday, November 22, 2011 4:42:45 PM 51

51. 073 A4.01 051 Given the following plant conditions:

- Unit I isatlOO%RTP.

- Alarm 1-RA-120B/121B STM GEN BLDN LIQ SAMP MON INSTR MALFUNC (M-1 2A-B6) annunciates.

Which ONE of the following describes the cause of the alarm and the mitigating actions that the crew would implement?

Cause Mitigating Action A. Loss of power to the Verify automatic termination of radiation monitor SIG blowdown release.

B Loss of power to the Place the standby monitor in service radiation monitor C. High flow through the Verify automatic termination of radiation monitor SIG blowdown release.

D. High flow through the Place the standby monitor in service radiation monitor DISTRACTOR ANALYSIS:

A: Incorrect, Plausible since the first part is correct, however the loss of power to the monitor only causes the malfunction alarm to actuate and does not cause SG blowdown to auto isolate.

B. Correct, The loss of power to 1-RA-120/121 will cause the instrument failure alarm to be generated. In accordance with AR-M12-A, B-6 this failure does not cause an auto isolation. The operators are directed to place the standby monitor in service.

C. Incorrect, Plausible since a flow condition can cause this alarm, however it is low flow, not high flow which causes alarm. Also the second part is not correct, the malfunction alarm does not cause SG blowdown to auto isolate.

D. Incorrect, Plausible since a flow condition can cause this alarm, however it is low flow, not high flow which causes alarm. Also second part is correct.

Tuesday, November 22, 2011 4:52:2 1 PM 51

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 51 Tier: 2 Group 1 KIA: 073 Process Radiation Monitoring(PRM)

A4.01 Ability to manually operate and/or monitor in the control room:

Effluent release Importance Rating: 3.9 / 3.9 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates ability to monitor the progress of a potential release through the SG bbwdown system and determine based on alarm indications if a potential release has been automatically terminated or if manual action is required.

Technical

Reference:

0-AR-M12-A, B6 Proposed references None to be provided:

Learning Objective: OPT200.RM, obj. 3.m Question Source:

New Modified Bank Bank X Question History: SQN bank question that has been updated and correct answer changed for 1201 NRC exam.

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 4:52:21 PM 52

52. 076 K4.03 052 Given the following plant conditions:

- Unit 1 was at 100% power when a manual SI is initiated.

Which ONE of the following describes the automatic response of the ERCW system?

1-FCV-67-146, CCS Hx 1AI and 1A2 ERCW Return to Header B _(1)__ and 0-FCV-67-152, CCS HXs OBI and 0B2 Discharge to B Discharge Header _(2).

A. (1) CLOSES (2) REPOSITIONS to 35% position B. (1) remainsASlS (2) remains CLOSED Cv (1) remainsASiS (2) REPOSITIONS to 35% position D. (1) OPENS fully (2) remains CLOSED DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since 1-FCV-67-146, CLOSES and O-FCV-67-152 OPENS to 35% position the misconception exists- thinking opening the outlet allows more cooling for additional SI component heat loads.

B. Incorrect, Plausible is candidate thinks that maintaining status quo, 1-FCV-67-146 and O-FCV-67-152 remaining in their normal conditons, unchanged, considering that SI loads contribute to Component Coo/mg additional heat load only.

C. Correct, Per 1,2-47W611-67-5; 1-FCV-67-146, remains THROTTLED and O-FCV-67-152 OPENS to 35% position. This arrangement assures adequate cooling water under both normal and emenrgency conditions.

D. Incorrect, 1-FCV-67-146, OPENS fully; O-FCV-67-152, remaining CLOSED, is plausible based on potential confusion with the cross-tie valve FCV-67-151 which does remain closed to prevent unbalance and maintain header supply independence; which would imply that 1-FCV-67-146 must open to supply the IA Hdr.

Tuesday, November 22, 2011 4:58:59 PM 52

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 52 Tier: 2 Group 1 KIA: 076 Service Water K4.03 Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following:

Automatic opening features assciated with SWS isolation valves to CCW heat exchangers.

Importance Rating: 2.9 / 3.4 IOCFRPart55: 41.7 IOCFR55.43.b: N/A K/A Match: This question matches the K/A by having the operator determine the automatic opening/repositioning of ERCW (Service Water) valves to the CCS heat exchangers during an emergency event.

Technical

Reference:

1 ,2-47W611-67-5; 1 ,2-45N779-31, 38 Proposed references None to be provided:

Learning Objective: OPT200.ERCW obj. 1.b & 8.c Question Source:

New Modified Bank Bank X Question History: SQN bank question originally 076 A3.02 used on 2007 audit exam Comments: stem wording modified to make more operational Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?:

Tuesday, November 22, 2011 4:58:59 PM 53

53. 078 G2.4.11 053 Given the following plant conditions:

- Unit I is in MODE 3 following a shutdown.

- A loss of Auxiliary Air occurred.

- Auxiliary Feedwater was aligned as directed by AOP-M.02, Loss Of Control Air to support current plant operations.

- EA-3-4, Local Alignment of TD AFW LCV Backup Air Supply has been implemented.

Which ONE of the following completes the statement below?

The operators will use (1) to restore Auxiliary Feedwater valve operation to normal and bottle pressure must be at least _(2)____ to meet the OPERABILITY requirements following the restoration of the plant air systems.

A (1) EA-3-4 (2) 800 psig B. (1) EA-3-4 (2) 1500 psig C. (1) AOP-M.02 (2) 800 psig D. (1) AOP-M.02 (2) 1500 psig DISTRA CTOR ANAL YSIS:

A. Correct, EA-3-4, Local Alignment of TD AFW LCV Backup Air Supply Section 4.3 is the procedure that will be used to return the AFW system to normal. Also the minimum design criteria of 800 psig bottle pressure is needed to ensure adequate valve operation.

B. Incorrect, Plausible since first part is correct, however 1500 psig is the pressure that a WO is written to have the bottles changed out, but the minimum design pressure is 800 psig.

C. Incorrect, Plausible since AOP-M.02 is the overall controlling procedure, however it will direct to operators to use EA-3.4 to perform the realignment. Also the second part is correct.

D. Incorrect, Plausible since AOP-M.02 is the overall controlling procedure, however it will direct to operators to use EA-3.4 to perform the realignment. Also the pressure is normal pressure to write a WO to get bottles changed out not minimum design pressure.

Tuesday, November22, 2011 5:10:40 PM 53

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 53 Tier: 2 Group 1 K/A: 078 Instrument Air System G2.4.1 1 Knowledge of abnormal condition procedures Importance Rating: 4.0 I 4.2 IOCFRPart55: 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidate knowledge of the abnormal procedures associated with loss of Instrument Air to the TD AFW pump flow control valves, and the requirements for air pressure to continuel operation of the those values.

Technical

Reference:

AOP-M.02, EA 3-4, TIS LCO 3.7.1.2 Proposed references None to be provided:

Learning Objective: OPL27IAOP-M.02, Obj. 6 & 8 Question Source:

New Modified Bank Bank X Question History: SQN bank question (original question #AOP-M.0282)

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 5:10:40 PM 54

54. 103 A2.03 054 Given the following plant conditions:

- Unit us at 98% power.

- The Channel III Containment HIGH-HIGH PRESS, pressure switch is out of service for surveillance testing, and its bistable is in BYPASS.

- The associated Channel III HIGH PRESS, bistable has been TRIPPED as required by Tech. Specs.

- A loss of ONE 161 kv offsite power line occurs.

- The resulting voltage spike causes the Channel II Containment HIGH PRESS, pressure switch to fail HIGH.

Which ONE of the following describes both, (1) the Containment Isolation system response, and (2) the procedure the crew would implement for this event?

Containment Isolation response Procedure A. Phase B AOP-P.05, Loss of Unit 1 Shudown Boards B. Phase B E-O, Reactor Trip or Safety Injection C. Phase A AOP-P.05, Loss of Unit 1 Shutdown Boards D Phase A E-O, Reactor Trip or Safety Injection DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate gets confused at to which bystables would become tripped. When channel 2 falls high, it makes up the logic for SI only. Also if the candidate does not recognize that the EDG5 would pick up the board they would assume going to AOP-P.05 would be correct.

B. Incorrect, Plausible if the candidate gets confused at to which bystables would become tripped. When channel 2 falls high, it makes up the logic for SI only. Also the second part is correct. The unit would trip and the crew would implement E-O.

C. Incorrect, Plausible since the first part is correct. However if the candidate does not recognize that the EDGs would pick up the board they would assume going to AOP-P.05 would be correct.

D. Correct, With a High-High Press switch in BYPASS, the logic for Phase B and containment spray becomes 2/3; with the High Press switch in Tripped, the logic for SI becomes 1/2. When channel 2 falls high, it makes up the logic for SI only.

EDGs will pickup shutdown boards therefore E-O would be implemented instead of ECA-O. 0.

Tuesday, November 22, 2011 5:21:48 PM 54

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 54 Tier: 2 Group 1 K/A: 103 Containment System A2.03 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:

Phase A and Phase B Isolation Importance Rating: 3.5 I 3.8 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the logic which initiated CNMT Phase A and Phase B. Also tests the candidates knowledge of the procedure that would be implemented for the conditions in the stem.

Technical

Reference:

47W61 1-88-1 Tl-28Att9 Proposed references None to be provided:

Learning Objective: 0PT200.CNTMTSTRUCTURE, Obj 6 Question Source:

New Modified Bank Bank X Question History: SQN bank question (used on 2008 Audit exam)

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 5:21:49 PM 55

55. 103 A2.05 055 Given the following plant conditions:

- Unit 2 is operating at 100% power.

- Alarm LS-68-53A/B REAC COOL PMP 2 OIL RESERVOIR LEVEL HI-LOW (M5-B) has just been received.

- An emergency containment entry is being prepared to allow maintenance to add oil to the #2 RCP motor lower bearing from the remote fill line.

In accordance with 0-Pl-OPS-000-01 1.0, Containment Access Control, which ONE of the following identifies both:

(1) the chemical hazard that is evaluated prior to making the containment entry and (2) the method used to control personnel exposure if the chemical is greater than the limit or allowed stay time?

A (1) Formaldehyde (2) Containment Purge B. (1) Formaldehyde (2) Containment Air Return Fans C. (1) Hydrogen (2) Containment Purge D. (1) Hydrogen (2) Containment Air Return Fans Tuesday, November 22, 2011 5:44:59 PM 55

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DIS TRACTOR ANAL ISIS:

A. Correct, As directed by 0PI-OPS-000-01 1.0, Containment Access Control, formaldehyde concentrations are measured and evaluated at Sequoyah prior to personnel entry in accordance with 0-Tl-OPS-000-001.0, Containment Formaldehyde Stay Time Calculation. If the formaldehyde concentration is greater than the limit, the preformer is directed to initiate a containment purge of the associated compartment in accordance with 0-SO-30-3.

B. Incorrect, Plausible since formaldehyde is sampled prior to entry into containment, however even though the containment air return fans are located in the compartment they would not reduce the formaldehyde concentration within the containment building.

C. Incorrect, Plausible since hydrogen is added to the RCS and would be present if a leak existed and because has explosive nature at higher concentrations the candidate may select Hydrogen as the chemical hazard of concen. However formaldehyde is the hazard of concern in 0-Pl-OPS-000-01 1.0. Also if hydrogen was present a containment purge would be reasonable to remove/reduce the concentration.

D. Incorrect, Plausible since hydrogen is added to the RCS and would be present if a leak existed and because has explosive nature at higher concentrations the candidate may select Hydrogen as the chemical hazard of concen. However formaldehyde is the hazard of concern in 0-Pl-OPS-000-01 1.0. Also the containment purge fans are installed partly to ensure the hydrogen is mixed and does not form large pockets of concentrated gas, thus it would be reasonable to use fans to disperse the hydrogen gas.

Tuesday, November 22, 2011 5:44:59 PM 56

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 55 Tier: 2 Group 1 KIA: 103 Containment System A2.05 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 Rating: 2.9 / 3.9 10 CFR Part 55: 41.5 IOCFR55A3.b: not applicable K/A Match: This question matches the K/A by having the candidate determine the chemical hazard that required to be sampled for prior to making an emergency entry into upper containment and based on the quidance of procedure identify the action necessary to take to reduce the concentration to acceptable levels.

Technical

Reference:

0-Pl-OPS-000-01 1.0 rev 1 0-Tl-OPS-000-001 .0 rev 6 Proposed references None to be provided:

Learning Objective: OPTSTG200 .CNTMTSTRUCTU RE obj 6 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 5:44:59 PM 57

56. 002 K6.06 056 Given the following plant conditions:

- Unit I is shutdown.

- All RCS temperatures are approximately 300°F.

- Pressure is 600 psig.

- Cold overpressure protection is armed.

- Both PORVs are closed.

- Loop 3 cold leg temperature instrument (68-60) fails LOW.

Which ONE of the following identifies the plant response to this instrument failure?

Reference Provided Av PORV-334 will open until pressure decreases to 20 psig below the minimum lift setpoint.

B. Only PORV-334 will open and remain open.

C. Both PORV-334 and PORV-340 will open and remain open.

D. Neither PORV will open.

DIS TRACTOR ANALYSIS:

A. Correct, Three of the four Cold Overpressure Protection system inputs are made up. COPS is Armed such that an input from RCS temperature detector would cause the PORV-334 only (since Loop 3 or 4 temperature feeds PORV-334) to open and actuate as designed to reduce pressure.

B. Incorrect, Plausible if candidate thinks that the failed input would cause the PORV to open and remain open. There are other singe instrument failures which would affect the PORV in that way, however RCS temperature is not one of them.

C. Incorrect, Plausible if candidate thinks that both PORVs will get an open signal, however the two PORVs have different inputs with Loop I & 2 input to PORV-340 and Loops 3 & 4 input to PORV-334, so the temperature detector for loop 3 only inputs to PORV-334.

D. Incorrect, Plausible since three of the four inputs to PORV-334 for COPS are made up (Temp <350°F; RCS COPS ARM switch selected to ARM; and PORV-334 switch selected to P-Auto.), however it takes a combination of temperature and pressure to get the signal for COPS to actuate.

Tuesday, November 22, 2011 5:56:37 PM 56

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 56 Tier: 2 Group 2 KIA: 002 Reactor Coolant System (RCS)

K6.06 Knowledge of the effect of a loss or malfunction on the following RCS components:

sensors and detectors Importance Rating: 2.5 I 2.8 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate determine the effects of a failed RCS temperature instrument on the Cold Overpressure Protection System control and logic.

Technical

Reference:

1-AR-M5.C, Windows C-5, 0-6, D-5, D-6 Proposed references Sequoyah Unit 1 LTQPS Selected Setpoints graph pg to be provided: 11 of PTLR Learning Objective: OPT200.PZR-PRT, Obj. 4 Question Source:

New Modified Bank Bank X Question History: SQN question from PZR PRESS-B.12.C Comments: Stem revised to different temperature instrument and higher starting pressure, also distractors revised to make more credible.

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 5:56:37 PM 57

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam

57. 015 K2.01 057 Which ONE of the following identifies all nuclear instruments that are powered from 120V AC Vital Instrument Power Board 1-I?

A. N-31 and N-41 gjjy B. N-35 and N-41 çpjy C. N-31 and N-35fly Dv N-31, N-35, and N-41 DIS TRA CTOR ANAL YSIS:

A. Incorrect, Intermediate Range Instrument N-35 is also powered from I2OVAC Vital Instrument Power Board 1-I. Plausible because the two nuclear instruments listed are powered from the 120V AC Vital Instrument Power Board 1-I.

B. Incorrect, Source Range instrument N-31 is also powered from I2OVAC Vital Instrument Power Board 1-I. Plausible because the two nuclear instruments listed are powered from the 12OVAC Vital Instrument Power Board 1-I.

C. Incorrect, Power Range instrument N-41 is also powered from I2OVAC Vital Instrument Power Board 1-I. Plausible because the two nuclear instruments listed are powered from the I2OVAC Vital Instrument Power Board 1-I.

D. Correct, All three of the nuclear instruments listed (N-31, N-35, and N-41) are the nuclear instruments powered from I2OVAC Vital Instrument Power Board 1-I.

Sunday, November 27, 2011 3:31:42 PM 57

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 57 Tier: 2 Group 2 KIA: 015 Nuclear Instrumentation System K2.O1 Knowledge of the bus power supplies to the following:

N IS channels, components, and interconnections.

Importance Rating: 2.7 I 2.8 IOCFRPart55: 41.7 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the power supply to the Nuclear Instruments.

Technical

Reference:

AOP-I.01, Nuclear Instrument Malfunction, Rev. 9 Proposed references None to be provided:

Learning Objective: OPT200.NIS obj. 6 Question Source:

New Modified Bank Bank X Question History: SQN bank question used on Sept 09 Retake exam Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Sunday, November 27, 2011 3:31:42 PM 58

58. 016 K5.01 058 Which ONE of the following identifies the primary function of the Isolation Amplifier associated with the Pressurizer Pressure Transmitters?

A. Protects the Pressurizer Pressure Control circuit.

B Protects the Solid State Protection System circuit.

C. Amplifies the pressure output signal between containment and the instrument rack.

D. Amplifies the pressure output signal between the instrument rack and the main control room.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since protection and control signals are isolated from one another, however Reactor Protection is what is being protected not Control signal B. Correct, Isolation Amplifiers protect the SSPS from signal perturbations that could arise in the control portion of the circuit.

C. Incorrect, Plausible if the candidate thinks that the signal output is boosted prior to being sent to instrument rack to protect the pressure transmitters.

D. Incorrect, Plausible if the candidate thinks that a signal boost is needed since there is a long run of cable between Containment and the Instrument racks.

Tuesday, November 22, 2011 6:24:12 PM 58

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 59 Tier: 2 Group 2 K/A: 016 Non-Nuclear Instrumentation System K5.01 Knowledge of the operational implications of the following concepts as they apply to the NNIS:

Separation of control and protection circuits.

Importance Rating: 2.7*/2.8 10 CFR Part 55: 41.5 IOCFR55.43.b: Not applicable KIA Match: This question matches the K/A because it asks the candidate to recall the purpose of the Isolation amplifiers in this portion of NNIS.

Technical

Reference:

OPT200.EAGLE21, pg 20 Proposed references None to be provided:

Learning Objective: OPT200.RPS Obj 6 Question Source:

New Modified Bank Bank X Question History: modified from bank question PZR Press-B.2 006 Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 2/2010 Last 2 NRC?: NO Tuesday, November 22, 2011 6:24:12 PM 59

59. 033 A1.01 059 Given the following plant conditions:

- Both Units are operating at 100% power.

- Alarm, LS-78-3 SPENT FUEL PIT LEVEL HIGH-LOW, (M6-D, D3) has just been received.

- The Aux Bldg AUO reports a leak on the discharge of the in-service Spent Fuel Pit Cooling Pump.

Assuming NO operator action, which ONE of the choices correctly completes the following statement?

Spent Fuel Pit level lowers until the running pump A. begins to cavitate at the NPSH limit.

B. trips on low spent fuel pit level interlock.

C becomes air-bound when the suction line uncovers.

D. becomes air-bound when the anti-siphon holes uncover.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since pump cavitation is likely to occur, this will not completely stop the outflow of water from the discharge of the pump.

B. Incorrect, Plausible if the candidate gets the SFP pump confused with other pumps which trip on low level in the tank, however for the Spent Fuel Pool there is no interlock on low pool level.

C. Correct, The system design is such that when the suction strainer becomes uncovered the Spent Fuel Pool cooling pump will lose its suction and prevent the pump from completely draining the Spent Fuel Pool, and possibly uncovering the fuel bundles in the Spent Fuel racks.

D. Incorrect, Plausible since some of the piping for the Spent Fuel Pool, does have siphon break holes drilled in them, however this in on the discharge of the pumps, not the suction, thus they would not affect operation of the Spent Fuel Pool pumps for low level conditions.

Tuesday, November 22, 2011 6:35:47 PM 59

QUESTIONS REPORT for SQN JAN 2012 NRC RD Exam Question Number: 59 Tier: 2 Group 2 K/A: 033 Spent Fuel Pool Cooling (SFP) System Al .01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the spent fuel pool cooling system controls, including:

Spent Fuel Pool level.

Importance Rating: 2.7! 3.3 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates ability to predict how the Spent Fuel Pool Cooling system will respond to lowering level in the Spent Fuel Pool without operator action.

Technical

Reference:

47W855-l Proposed references None to be provided:

Learning Objective: OPT200.SFPC, Obj 8 Question Source:

New Modified Bank Bank X Question History: SQN bank question, Modified from a Salem I RD exam on 2002.

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November22, 2011 6:35:47 PM 60

60. 034 A4.01 060 Given the following plant conditions,

- Unit 1 is in Mode 6.

- Refueling activities are in progress.

- A fuel assembly has been slightly damaged during removal from the core.

- AOP-M.04, Refueling Malfunctions, is in progress.

Which ONE of the following would provide the FIRST indications in the control room of a developing leak from the damaged fuel assembly?

A. Annunciator 1-RA-90-106A CNTMT BLDG LWR COMPT AIR MON HI RAD (1-M12-A, A4) alarms and slowly rising counts on 1-RM-90-59A, Reactor Building Area Rad Monitor.

B. Annunciator1-RA-272A UPPR IN CNTMT HI RAD (1-M30, B1)and slowly rising counts on 1-RM-90-130A and 131A Containment Purge Exhaust Monitors.

C Annunciators CONTAINMENT VENTILATION ISOLATION TRAIN A or TRAIN B (1-M6-C, C-5 & C-6 respectively) and rapidly rising counts on 1-RM-I12AIB/C Upper Containment Air Monitor D. Annunciator 1-RA-90-59A RX BLDG AREA RAD MON HIGH RAD, and slowly rising counts on 1-RM-90-106, Containment Lower Compartment Radmonitor.

Tuesday, November 22, 2011 6:44:23 PM 60

QUESTIONS REPORT for SQN JAN 2012 NRC RD Exam DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since 1-RM-90-106 is a gaseous monitor and is used during normal plant operation to detect small RCS leaks in the lower containment (where the RCS components are located), however during periods of refueling the first monitor that will detect gaseous activity is 1-RM-90-1 12 for upper containment.

Also the area monitor 1-RM-90-59 may show some increase, however the activity released from a damaged fuel assembly is radioactive gases which is more readily detected by the gaseous monitors.

B. Incorrect, Plausible since 1-RA-272 is an upper containment rad monitor, and 1-RM-90-130 & 131 would see the gaseous activity releases from the fuel assembly, however the 1-RA-272 is a post accident monitor with an alarm setpoint of IOOR/hr. This alarm is used to detect large amounts fuel damage during an inadeuate core cooling event and would not alarm for the given conditions.

C. Correct, With a damaged fuel assembly inside containment, the first rad monitors that would detect the failure would be monitors that sample for radioacitive gases which may be present in the gas gap of the fuel rods. Since a containment purge is established as part of the refueling lineup the purge air rad monitors (1-RM-90-130

& 131) would sense the leak quickly and initiate a Containment Vent Isolation, also the Upper Containment air monitor 1-RM-90-1 12, would also sense the radioactive gases that were released and show an upward trend.

D. Incorrect, Plausible since 1-RM-9O-59 is an area monitor that would see an increase in area activity and 1-RM-90-106 may see some increase in counts from the gaseous release of the damaged fuel assembly, however for the area monitor go into alarm the fuel assembly would have to be brought near the surface of the water in the refueling canal, not just leak some of its gases, thus this would not be the first indication of the event in the control room.

Tuesday, November 22, 2011 6:44:23 PM 61

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 60 Tier: 2 Group 2 KIA: 034 Fuel Handling Equipment A4.01 Ability to manually operate and or monitor in the control room:

Radiation levels Importance Rating: 3.3 I 3.7 10 CFR Part 55: 41.7 10CFR55.43b: not applicable KIA Match: This question matches the K/A by having the candidate determine the radition monitors in the control room that would first tell them there was some damage to a fuel assembly inside containment during fuel handling operation.

Technical

Reference:

AOP-M.04, Refueling Malfunctions 1-XA-55-6C, C-5 & C-6, Proposed references None to be provided:

Learning Objective: OPL271AOP-M.04 obj 2 & 3 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 6:44:23 PM 62

61. 035 A3.01 061 Given the following plant conditions:

- Unit 2 is at 50% RTP

- Turbine Impulse PT-i -73 fails low.

Assuming NO operator action, which ONE of the following identifies the effect, if any, on the Distributed Controls System (DCS)?

A Automatic SG Level Control will be maintained.

B. SG Level Control will switch to Manual on all 4 SGs.

C. SG Level will decrease flow initially and level will stabilize at 33%.

D. SG Level will increase flow resulting in Feed Water Isolation at 81 % level.

DISTRA CTOR ANAL YSIS:

A. Correct, When the Distributed Controls System (DCS) was installed, part of the modification was to add 2 additional Turbine Impulse Pressure Transmiters. The DCS compares the three pressure transmitters and will automatically bypass any signal that it deems outside the preselected range. It then takes the average of the other two signals for the control function which for Turbine Impulse Pressure is to establish the ramp from 33% at no load to 44% at full load.

B. Incorrect, Plausible since some of the inputs to DCS have only two signals and DCS will transfer control from Auto to Manual if it deems the inputs as questionable, thus if the candidate gets confused as to which failed inputs will cause the DCS to switch to manual they would select this answer.

C. Incorrect, Plausible since this is the way the system would respond prior to the installation of DCS. With the Turbine Impulse Pressure failed low the automatic level control would assume the no load value for level control and would allow SG level to decrease to 33% and then control at that value.

D. Incorrect, Plausible if the candidate thinks that the failed input would cause the DCS to lose its reference and continue to raise SG level until either operator action or automatic FWI and Hi SG Level trip would occur.

Tuesday, November 22, 2011 6:53:42 PM 61

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 61 Tier: 2 Group 2 K/A: 035 Steam Generator System (SIGS)

A3.01 Ability to monitor automatic operation of the SIG including:

SIG water level control Importance Rating: 4.0 I 3.9 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate determine how an input to Distrubitive Controls System (DOS) will affect the system during operation when the controls are in automatic.

Technical

Reference:

AOP-L08, Turbine Impulse Pressure Instrument Malfunction, Rev 10 Proposed references None to be provided:

Learning Objective: OPT200.MFW obj. 12 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 6:53:42 PM 62

62. 071 G2.2.40 062 Which ONE of the following choices completes the statement below?

In accordance with Technical Specifications, the Waste Gas Decay Tank (WGDT) oxygen limit (in percent volume) is (1) when the hydrogen concentration is> (2) (in percent volume).

A. 2% 2%

B 2% 4%

C. 4% 2%

D. 4% 4%

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate gets confused on the limits for Oxygen and Hydorgen. The 02 limit is corrrect at 2%, but the H2 limit is not 2%; it is 4%.

B. Correct, In accordance with Tech Spec LCO 3.11.2.5, the limit of Oxygen is 2% by volume whenever Hydrogen concentration is greater than 4% by volume. Concern is flammability of 02 and H2 combined.

C. Incorrect, Plausible if the candidate gets confused on the limits for Oxygen and Hydorgen The 02 limit is not 4%; it is 2%. H2 limit is not 2%; it is 4%.

D. Incorrect, Plausible if the candidate gets confused on the limits for Oxygen and Hydorgen. The 02 limit is not 4%; it is 2%, but the H2 limit is correct at 4%.

Tuesday, November22, 2011 7:00:13 PM 62

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 62 Tier: 2 Group 2 KIA: 071 Waste Gas Disposal System G 2.2. 40 Ability to apply Technical Specifications for a system.

Importance Rating: 2.7 / 3.5 IOCFRPart55: 41.10/43.5/45.12 IOCFR55.43.b: Not applicable KIA Match: This question matches the K/A by having the candidate apply the requirements of Oygen and Hydrogen concentrations that are allowed, in waste gas decay tanks, by Tech Spec 3.11.2.5 Explosive Gas Mixture Technical

Reference:

Technical Specification 3.11.2.5 Amend 301 Proposed references None to be provided:

Learning Objective: OPT200.TS-APP obj 3 Question Source:

New Modified Bank Bank X Question History: WBN 07 exam question 071 Al .02 O0lwith formatting changes. Question used on Feb 2010 Audit Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 7:00:13 PM 63

63. 072 K4.02 063 Given the following plant conditions:

- Both Units are at 100% power

- 0-RM-90-103A, Spent Fuel Pit Radiation Monitor, fails HIGH Which ONE of the following indentifies the expected plant response?

A. ONLY Train B ABGTS cleanup fan starts; ONLY Train B Aux. Building General Supply, Exhaust and Fuel Handling exhaust fans trip.

B ONLY Train B ABGTS cleanup fan starts; ALL Aux. Building General Supply, Exhaust and Fuel Handling exhaust fans trip.

C. BOTH Train A and Train B ABGTS cleanup fans start; ONLY Train B Aux. Building General Supply, Exhaust and Fuel handling exhaust fans trip.

D. BOTH Train A and Train B ABGTS cleanup fans start; ALL Aux. Building General Supply, Exhaust, and Fuel handling exhaust fans trip.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since the first part of answer is correct and if the candidate thinks that the entire Aux Bldg vent system is train specific then only the train B components will actuate. However due to relaying both Aux Building General Supply, exhaust and fuel handling exhaust fans will trip.

B. Correct, The Fuel Pool Rad monitors are train specific. 0-RM-90-103A will cause a Train B Aux Building Gas Treatment System cleanup fan to start. And due to non-train specific relays both Aux Building General Supply, exhaust and fuel handling exhaust fans trip. It would take both monitors (RM-90-102 and 103 to have both trains actuate)

C. Incorrect, Plausible if the candidate does not know that RM-90-102 would actuate Train A and RM-90-103 will actuate Train B of the Aux Bldg Gas Treatment (ABGTS). Also second part is not correct.

D. Incorrect, Plausible if the candidate doesnt know that RM-90-102 and 103 are train specific which is unlike other automatic isolations. Also the second part is correct.

Tuesday, November 22, 2011 7:13:45 PM 63

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 63 Tier: 2 Group 2 K/A: 072 Area Radiation Monitoring(ARM) system K4.02 Knowledge of ARM system design feature(s) and/or interlock(s) which provide for the following:

Fuel Building Isolation Importance Rating: 3.2 / 3.4 10 CFR Part 55: 41.7 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate identify the interlock(s) associated with Fuel Pool Area Radiation monitors as they apply to Fuel Handling Area Isolation.

Technical

Reference:

O-AR-M12b, Windows B-3, B-5 45N630-4, 45N779-5 Proposed references None to be provided:

Learning Objective: OPT200.RM obj. 4 & 6 Question Source:

New Modified Bank Bank X Question History: Question used on 2008 Audit exam.

Comments: SQN bank question modified such that RM-90-103A is rad monitor in stem vs RM-90-102.

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November22, 2011 7:13:45 PM 64

65. 086 A2.04 065 Given the following plant conditions,

- Both Units are operating at 100% power.

- Several fire alarms are received from the Unit I Auxiliary Building (charging pump area and cross zone detectors are in alarm).

- An AUO is sent to investigate.

- A Unit 2 CR0 reports that neither the electric fire pump nor the diesel fire pump is running.

- Prior to a report from the Aux Bldg AUO, the IA-A charging pump trips and LCV-62-1 32, VCT Outlet To CCP, indicates CLOSED.

- The crew has implemented AOP-N.01, Plant Fires.

Which ONE of the following identifies the expected plant response to the above conditions and the required procedure actions?

A. (1) The fire pumps would not be required to start for these conditions.

(2) Go to AOP-M.09, Loss of Charging.

B. (1) The fire pumps would not be required to start for these conditions.

(2) Go to AOP-N.08, Appendix R Fire Safe Shutdown.

C. (1) Manually start the Electric Fire pump from 0-M-29.

(2) Go to AOP-M.09, Loss of Charging.

D (I) Manually start the Electric Fire pump from 0-M-29.

(2) Go to AOP-N.08, Appendix R Fire Safe Shutdown.

Tuesday, November22, 2011 7:47:16 PM 65

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam DISTRACTOR ANALYSIS:

A. Incorrect, Plausible if the candidate did not recognize that start criteria for the electric pump was present. Also plausible if candidate choses to go to AOP-M.09 to restore normal charging, which would normally be done. However as stated in the purpose of AOP-M.09, this procedure does not apply if CCP trip due to Appendix R fire (AOP-N.08)

B. Incorrect, Plausible if the candidate did not recognize that start criteria for the electric pump was present. Also plausible since the second part of the answer is correct.

C. Incorrect, Plausible if candidate recognizes that the electric fire pump should have started but did not, thus requiring the diesel fire pump to be started. However for the condlitons in the stem the correct procedure would be AOP-N.08, Appendix R fire due to closing of the CCP suction valve.

D. Correct, Given the conditions that not only the local fire detectors are in alarm but also the cross zone detectors the local deluge valve should have opened, and thus causing the electric fire pump to start. The direction in AOP-N.O1, is to ensure at least one fire pump is running, with the preferred pump being the electric fire pump, if it does not start then start the diesel driven pump. The guidance in AOP-N.O1, step 9 states that if any of the listed items occurs then the crew will either transition to A OP-C. 04 or A OP-N. 08. The suction valves of the CCPs going closed is one of the criteria, and with the fire in Aux Bldg, the crew would go to AOP-N.08.

Tuesday, November 22, 2011 7:47:16 PM 66

QUESTIONS REPORT forSQN JAN 2012 NRC RD Exam Question Number: 65 Tier: 2 Group 2 KIA: 086 Fire Protection System (FPS)

A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure to actuate the FPS when required, resulting in fire damage Importance Rating: 3.3 I 3.9 10 CFR Part 55: 41.5 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by having the candidate determine the potential effects of a failure of the FPS to properly actuate for a valid signal and which procedure would be used to mitigate the consequences of the failures caused by the fire.

Technical

Reference:

AOP-N.01, Plant Fires, rev 28 AOP-N.08, Appendix R Fire Safe Shutdown, rev 14 AOP-M.09, Loss of Charging, rev 2 Proposed references None to be provided:

Learning Objective: OPL27IAOP-N.01 obj 5 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 7:47:16 PM 67

66. G2.1.17 066 Given the following plant conditions:

- Unit I has automatically tripped from 100% and SI has actuated.

- The crew is performing actions in E-0, Reactor Trip or Safety Injection.

- You are the At the Controls operator.

- The following conditions exist:

- RCS temperature is 540°F and lowering

- RCS pressure is 1400 psig and stable

- Containment pressure is 3 psig and rising.

- The Unit Supervisor has reached step 8, CHECK RCP trip criteria,

- The Unit Supervisor states OATC, Check RCS pressure less than 1250 psig AND At least one CCP or SI pump RUNNING.

In accordance with OPDP-1, Conduct of Operations, which ONE of the following is the correct method to communicate the requested information to the Unit Supervisor?

A. RCS pressure is NOT less than 1250 psig and at least one CCP or SI pump is running.

B. Yes, RCS pressure is NOT less than 1250 psig and at least one CCP or SI pump is running.

C No, (Name), RCS pressure is 1400 psig and stable and all CCPs and SI pumps are running.

D. Yes, (Name), RCS pressure is NOT less thanl25O psig, it is 1400 psig and all CCP and SI pumps are running.

Tuesday, November 22, 2011 8:01:24 PM 66

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate thinks that they need to repeat back the question as stated however by the example provided mt OPDP-1, the parameter requested with a value units and a trend are needed to meet the expectations of the OPDP-1, however this response does not answer the question with a yes or no.

B. Incorrect, Plausible if the candidate thinks that they need to repeat back the question as stated however by the example provided mt OPDP-1, the parameter requested with a value units and a trend are needed to meet the expectations of the OPDP-1.

C. Correct, In accordance with OPDP-1, when a crew member is to communicate important plant information the communications of indicator readings should be provided in the format of PARAMETER-VALUE-UNITS-TREND (with rate when appropriate. Thus the communication would answer the question (yes or no) then provide the parameter value units trend to complete the response correctly.

D. Incorrect, Plausible if the candidate thinks that they need to repeat back the question as stated however by the example provided mt OPDP-1, the parameter requested with a value units and a trend are needed to meet the expectations of the OPDP-1.

Tuesday, November 22, 2011 8:01 :24 PM 67

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 66 Tier: 3 Group KIA: G 2.1.17 Ability to make accurate, clear, and concise verbal reports.

Importance Rating: 3.9 I 4.0 10 CFR Part 55: 41.10 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by testing the candidates ability to determine when the use of 3 way communication are required to ensure accurate information is relayed to others.

Technical

Reference:

OPDP-1, rev 20 Proposed references None to be provided:

Learning Objective: 0PL271 OPSM-GMT obj A Question Source:

New Modified Bank Bank X Question History: Bank question used on 2009 Palo Verde NRC exam.

Updated stem and distractors to make plant specific.

Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 8:01:24 PM 68

67. G 2.1.32 067 Given the following plant conditions:

- Unit I is in Mode 5 with water solid operation.

- RHR Train A is operating in shutdown cooling mode.

- RCS is at 160°F.

- VCT temperature is 115°F.

- #2 RCP was the only RCP running and was inadvertently stopped 10 minutes ago after a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> run.

Which ONE of the following identifies the requirements that must be met prior to attempting a restart of the #2 RCP?

A. Must wait 30 minutes from the time the pump was stopped and can be restarted with current plant conditions.

B. Must wait 30 minutes from the time the pump was stopped but a steam bubble must be established in the Pressurizer.

C. No time restriction on restarting the pump and can be restarted with current plant conditions.

D No time restriction on restarting the pump but a steam bubble must be established in the Pressurizer.

Tuesday, November 22, 2011 8:06:53 PM 67

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DISTRACTOR ANALYSIS:

A. Incorrect, There is no time restriction on restarting the pump and a steam bubble is needed in the pressurizer. Plausible because the 30 minutes could be applied since the time of the pump stop instead of the time of the start and an RCP can be started in solid water operation with different conditions.

B. Incorrect, 1-SO-68-2 Precaution F and 0-GO- 1 Precaution J. 10 identify that when the RCS temperature is greater than charging and seal injection temperature and the RCPs have been idle for> 5 minutes, a steam bubble is needed in the pressurizer prior to starting an RCP to minimize the pressure transient when the cold water injected by the charging pump is circulated in the RCS. Plausible because the 30 minutes could be applied since the time of the pump stop instead of the time of the start and the need for a steam bubble is correct.

C. Incorrect, There is no wait required to restart the pump, however a steam bubble is needed in the pressurizer in accordance with the precautions in 1-SO-68-2 and 0-GO-I. Plausible because there is no required time wait prior to starting the pump and an RCP can be started in solid water operation with different conditions.

D. Correct, I-SO-68-2 Precaution F and 0-GO- I Precaution J. 10 identify that when the RCS temperature is greater than charging and seal injection temperature and the RCPs have been idle for> 5 minutes, a steam bubble is needed in the pressurizer prior to starting an RCP to minimize the pressure transient when the cold water injected by the charging pump is circulated in the RCS. I-SO-68-2 Precaution D identifies that thirty minutes must be allowed between RCP starts.

The pump has been in seivice for an hour, therefore, the 30 minute time between starts is met.

Tuesday, November 22, 2011 8:06:53 PM 68

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 67 Tier: 3 Group KIA: G 2.1.32 Ability to explain and apply system limits and precautions Importance Rating: 3.8 I 4.0 IOCFRPart55: 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the precautions and limitations that apply to starting RCPs during shutdown conditions.

Technical

Reference:

l-SO-68-2 precaution D & F 0-GO-i, precaution J Proposed references None to be provided:

Learning Objective: OPT200.RCP obj. 10 Question Source:

New Modified Bank Bank X Question History: SQN bank question used on 1/2009 NRC exam Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYA}I Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 8:06:53 PM 69

68. G 2.1.41 068 Given the foNowing plant conditions:

- Unit 1 core re-load is in progress with sixty assemblies loaded in the core.

- The movement of a Source Bearing Fuel assembly is in progress.

- Source Range Detector Nl-31 indicating 9 cps.

- Source Range Detector Nl-32 failed to bottom of scale.

- Both SRM High Flux at Shutdown switches are in BLOCK position

- Annunciator SOURCE RANGE HIGH SHUTDOWN FLUX ALARM BLOCK is in alarm.

- Source Range counts are audible in containment.

- Annunciator LS-78-3 SPENT FUEL PIT LEVEL HIGH-LOW is in alarm.

- Spent Fuel Pit level is at elevation 725 11.

- Spent Fuel Pit Boron concentration is 2180 ppm.

In accordance with Tech Specs, why must core alterations be suspended?

A. Spent Fuel Pit level is below the minimum.

B Only one Source Range Nuclear Monitor is in service.

C. Spent Fuel Pit boron concentration is below the minimum.

D. Source Range High Flux Level at Shutdown alarm must be in service.

DISTRA CTOR ANAL YSIS:

A. Incorrect, The level at the identified elevation is slightly above the high level alarm setpoint so the level is not below the Tech Spec minimum. Plausible because if the level were low the same alarm would be in and there is a Tech Spec requirement to have 23 feet of water above the fuel or immediately suspend fuel movement.

B. CORRECT, Tech Spec require both source range monitors to be in service and if not fuel movement is must be suspended immediately.

C. Incorrect, The boron concentration is above the Tech Spec minimum requirement of 2000ppm not below the Tech Spec minimum. Plausible because if the boron concentration was lower than the minimum there is a Tech Spec requirement to immediately suspend fuel movement.

D. Incorrect, The High flux at Shutdown Alarm is not required to be in service with the identified conditions. Plausible because the alarm is normally in service but may be blocked. O-RT-NUC-000-002. 0, Core Configuration, and GO-9, Refueling Operations, address blocking the alarm during movement of Source Bearing Fuel Assembly.

Tuesday, November22, 2011 8:11:40 PM 68

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 68 Tier: 3 Group KIA: G 2.1.41 Knowledge of the refueling process Importance Rating: 2.8 / 3.7 10 CFR Part 55: 41.2/41.10 IOCFR55.43.b: not applicable KIA Match:

Technical

Reference:

Tech Specs Section 9, Refueling Ammendment 322 0-RT-NUC-000-002.0, Core Reconfiguration, Rev 20 GO-9, Refueling Operations, Rev 34 Proposed references None to be provided:

Learning Objective: OPL271 GO-9 obj. 2 Question Source:

New Modified Bank Bank X Question History: Question used on 2009 Retake exam Comments: SQN question REFUELING-B.1.C 001 was original question Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 8:11:40 PM 69

69. G2.2.12 069 Given the following plant conditions:

- Unit 1 is at 100% power.

- It is 0630 and you have just relieved the Unit I CRC.

In accordance with 0-Sl-OPS-000-002.0, Shift Log, which ONE of the following completes the statement below?

The (1) is responsible for ensuring that all Outside MCR appendices are complete and the applicable day shift appendices are required to be completed by (2)

A. (1) Assistant Unit Operator (2) 1830 B. (1) Unit Operator/RO (2) 1830 C. (1) Assistant Unit Operator (2) 1000 Dv (1) Unit Operator/RO (2) 1000 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since normally the AUOs take the readings for the appendices Outside MCR and the candidate could get confused as to who is responsible.

Also based on a note in the procedure the appendices are verified complete by the end of the shift but the data is required between 0800 and 1000 on day shift.

B. Incorrect, Plausible since the first part is correct, and based on a note in the procedure the appendices are verified complete by the end of the shift but the data is required between 0800 and 1000 on day shift.

C. Incorrect, Plausible since normally the AUOs take the readings for the appendices Outside MCR and the candidate could get confused as to who is responsible.

Also the second part is correct to ensure the 25% extension time on suiveillances is not exceeded.

D. Correct, In accordance with 0-Sl-OPS-000-002. 0 shiftly logs states that although normally it is the AUO who completes Outside MCR appendices, it is ultimately the Unit Operator/RO responsibility to ensure the appendices are completed no matter who takes the readings. Also to ensure the 25% extension time for Tech Spec required surveillances are not exceeded the normal time for completion of the data on the appendices is between 0800 and 1000.

Tuesday, November22, 2011 8:15:47 PM 69

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 69 Tier: 3 Group KIA: G2.2.12 Knowledge of surveillance procedures.

Importance Rating: 3.7 /4.1 IOCFRPart55: 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the requirements contained in the daily surveillance procedure that the site uses to ensure that the Tech Spec required surveillances are being met.

Technical

Reference:

0-SI-OPS-000-002.0 rev 99 Proposed references None to be provided:

Learning Objective: OPL271LOGKEEPING obj. B.2 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12-01 Last 2 NRC?: NO Tuesday, November 22, 2011 8:15:47 PM 70

70. G 2.2.38 070 Given the following,

- Unit 1 is operating at 100% power.

- Control Rod H-8 drops into the core.

- Unit 1 remains at power and no automatic protective or operator actions have occurred.

- The OATC reports that during the initial stages of the event Pressurizer pressure had dropped to 2195 psig before recovering to normal.

Which ONE of the following completes the statements below?

During this event, the Tech Spec DNB limit _(1)____ exceeded.

Tech Spec 3.1.3 Moveable Control Assemblies, requires Reactor Power to be a maximum of _(2)__ within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, unless the rod is realigned.

A. (1)was (2) 85%

Be (1)was (2) 75%

C. (1)wasNOT (2) 85%

D. (1)was NOT (2) 75%

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since the first part is correct. The DNB tech spec limit is 2205 psia. Also the Tech Spec for Moveable Control Assemblies requires that reactor power be reduced to < 75% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and reactor trip setpoints must be reduced to < 85%. Since both requirements are in the same paragraph in Tech Spec the candidate made get confused as to which power level applies.

B. Correct, lAW Tech Spec 3.2.5, DNB Parameters, PZR pressure must remain

> 2220 psia while in Mode 1. Therefore with pressure at 2195 psig the DNB LCO would be in affect. Also Tech Spec 3.1.3.1 requires that reactor power be reduced to < 75% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Incorrect, Plausible if the candidate does not recognize the fact that the DNB parameter for pressure was exceeded and setpoint for reactor power is 75%

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. <85% applies to the reactor trip setpoints not the maximum power level.

D. Incorrect, Plausible if the candidate does not recognize the fact that the DNB parameter for pressure was exceeded and the power level for maximum power level is < 75%.

Tuesday, November22, 2011 8:20:57 PM 70

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 70 Tier: 3 Group K/A: G 2.2.38 Knowledge of conditions and limitations in the facility license.

Importance Rating: 3.6 / 4.5 10 CFR Part 55: 41.7/ 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate recall the Tech Spec limits for DNB parameters and the Tech Spec limit on reactor power for a dropped rod. This demonstrates that they have knowledge of these Tech Spec items as stated in Tech Specs and Tech Specs are part of the facility license.

Technical

Reference:

Tech Spec 3.2.5 Ammendment 138 Tech Spec 3.1.3.1 Moveable Control Assemblies Ammendment 215.

Proposed references None to be provided:

Learning Objective: OPT200.TS-APP obj. 3, 4 Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam.

Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYA}I Date: 12/0 1 Last 2 NRC?: NO Tuesday, November22, 2011 8:20:57 PM 71

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam

71. G 2.3.5 071 Given the following:

- The CRC is performing a source check on radiation monitor 0-RM-90-1 03, Spent Fuel Pit Radiation Monitor.

Which ONE of the following completes the following statement?

When the radiation monitor control switch is positioned to the CHECK SOURCE position the (1) and the High Rad relay (2) be automatically blocked.

A indicator will deflect upscale will NOT B. indicator will deflect upscale will C. operate light will illuminate will NOT D. operate light will illuminate will DISTRA CTOR ANAL YSIS:

A. CORRECT, If the O-RM-90-103 is source checked the indicator will deflect upscale if the check is successful. The high rad relay will not be automatically blocked by placing the control to the CHECK SOURCE position. To prevent the potential for an automatic actuation, the monitor would have to be manually blocked.

B. Incorrect, If the O-RM-90-103 is source checked the indicator will deflect upscale if the check is successful but the high rad relay being automatically blocked by placing the control to the CHECK SOURCE position is not correct. Plausible because the indicator will deflect upscale during the source check and the instrument can be manually blocked.

C. Incorrect, Placing the control on O-RM-90-103 to the CHECK SOURCE position does not cause the operate light to illuminate and the high rad relay not being automatically blocked by placing the control to the CHECK SOURCE position is correct. Plausible because there is an operate light and the instrument not being blocked is correct.

D. Incorrect, Placing the control on O-RM-90-103 to the CHECK SOURCE position does not cause the operate light to illuminate and the high rad relay being automatically blocked by placing the control to the CHECK SOURCE position is not correct. Plausible because there is an operate light and the instrument can be manually blocked.

Sunday, November27, 2011 3:46:42 PM 71

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 71 Tier: 3 Group K/A: G2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance Rating: 2.9 I 2.9 10 CFR Part 55: 41.11, 41.12 IOCFR55A3.b: not applicable KIA Match:

Technical

Reference:

RCI-01, Radiation Protection Program, Rev. 67 RCI-05.303, Calibration, Response Check and Operation of the Thermo Electron Small Article Monitor (SAM-i 1), Rev. 001 Proposed references None to be provided:

Learning Objective:

Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Sunday, November 27, 2011 3:46:42 PM 72

72. G23.13 072 Given the following plant conditions:

- A release of the Monitor Tank is in progress through the Liquid Radwaste System using the normal effluent flow path.

- O-RA-90-122A WDS LIQ EFF MON HIGH RAD (M12-B, C-i) annunciator alarms.

Which ONE of the following is required to continue the release if O-RM-90-122 is declared inoperable?

A. Release cannot be restarted until O-RM-90-122 is repaired and restored to operable status.

B. Notify chemistry to perform grab samples during the release and analyze for principal gamma emitters. Additional samples are to be documented on the post release permit.

C. No further action is required since the tank has been sampled and a valid permit has been generated to comply with the ODCM requirements to document the liquid radioactivity release.

D Release may resume only if two addtional samples are drawn and independently analyzed, release rate calculations are independently verified, and lineup independently verified.

Tuesday, November22, 2011 9:01 :52 PM 72

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam DISTRACTOR ANALYSIS:

A. Incorrect, The 0-AR-M12-B alarm response procedure requires the Chemistry Shift supervisor to comply with the ODCM requirements. The radiation monitor is not required to be operable as long as 2 independent samples of tank contents are analyzed. 2 independent release rate calculations, and 2 independent discharge valve alignment, are verified per ODCM 1.1.1 action 30. Plausible if candidate knows action for High Rad annunciator is to stop the release but does not know that the release can continue after meeting ODCM verification requirements.

B. Incorrect, The 0-AR-M12-B alarm response procedure requires the Chemistry Shift supervisor to comply with the ODCM requirements. The radiation monitor is not required to be operable as long as 2 independent samples of tank contents are analyzed. 2 independent release rate calculations, and 2 independent discharge valve alignment, are verified per ODCM 1.1.1 action 30. Plausible if student believes compensatory sampling is adequate for continuing the release. Many plant process monitors, such as SG blowdown radiation monitor 0-RM-90-120/121, allow for continuing the relese for inoperable radiation monitors if compensatory sampling at a specified frequency is performed.

C. Incorrect, The 0-AR-M12-B alarm response procedure requires the Chemistry Shift supervisor to comply with the ODCM requirements. The radiation monitor is not required to be operable as long as 2 independent samples of tank contents are analyzed. 2 independent release rate calculations, and 2 independent discharge valve alignment, are verified per ODCM 1.1.1 action 30. Plausible if student believes the existing release permit is adequate to document the release based upon the chemistry samples priop to the release.

D. Correct, The 0-AR-M12-B alarm response procedure requires the Chemistry Shift supervisor to comply with the ODCM requirements. The Radiation Monitor is not required to be operable as long as 2 independent samples of tank contents are analyzed, 2 independent release rate calculations, and 2 independent discharge valve alignment, are verified per ODCM 1.1.1 action 30.

Tuesday, November 22, 2011 9:01:52 PM 73

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 72 Tier: 3 Group K/A: G 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance Rating: 3.4 I 3.8 10 CFR Part 55: 41.12 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate demonstate knowledge of the radiological procedural requirements for starting, stopping and monitoring radioactive liquid releases.

Technical

Reference:

0-AR-M12-B C-i rev 29 ODCM Sect 1.1.1 action 30 rev 55 Proposed references None to be provided:

Learning Objective: 0PT200.ODCM obj. B.e Question Source:

New X Modified Bank Bank Question History: New question written for 1201 NRC exam Comments:

Source: NEW Source If Bank:

Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 9:01:52 PM 74

73. G 2.4.4 073 Given the following plant conditions:

- Unit 1 is cooling down for refueling.

- RCS pressure is 340 psig.

- Tavg is 175°F.

- IA-A DIG is tagged for maintenance.

- A loss of Off-Site power occurs.

- 1 B-B D/G falls to start.

Which ONE of the following identifies the plant procedure that the crew is required to enter to mitigate the event?

A. ECA-0.0, Loss of All AC B AOP-P.01, Loss of Offsite Power C. E-0, Reactor Trip or Safety Injection D. AOP-P.05, Loss of Unit 1 Shutdown Boards DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since this is a procedure which can be entered directly if the crew recognizes the station blackout, and this procedure is normally entered for Station Blackout conditions when at power, however ECA-0. 0 is only applicable in Modes 1-3 not Mode 4.

B. Correct, A OP-P. 01 is the applicable procedure for loss of Off-Site power conditions in Mode 4. It contains all the requirements to maintain the plant in a stable condition and restore power to the SD boards.

C. Incorrect, Plausible since E-0 is normally entered for emergency conditions that arise during normal plant operation, however the EOP network is not written to mitigate accidents that are originated in MODE 4, thus this is not the correct procedure to enter.

D. Incorrect, Plausible since AOP-P.05 does deal with loss of power to the SD boards, however a note at the beginning of AOP-P.05 states that ifa loss of off site power has occurred, AOP-P.01 takes precedence over AOP-P.05.

Tuesday, November 22, 2011 9:07:04 PM 73

QUESTIONS REPORT forSQN JAN 2012 NRC RO Exam Question Number: 73 Tier: 3 Group KIA: G 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

Importance Rating: 4.5 / 4.7 10 CFR Part 55: 41.10 IOCFR55.43.b: not applicable K/A Match: This question matches the K/A by determining if the candidate can recognize the correct procedure to enter for the given plant conditions.

Technical

Reference:

AOP-P.01, rev 25 AOP-P.05, rev 17 Proposed references None to be provided:

Learning Objective: OPL271AOP-P.O1 obj 2 Question Source:

New Modified Bank Bank X Question History: SQN bank question with distractors C & D changed to increase the plausibility.

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/01 Last 2 NRC?: NO Tuesday, November 22, 2011 9:07:04 PM 74

74. G2.4.32 074 Given the following plant conditions:

- Unit 1 is operating at 45%.

- Due to scheduled maintenance in the 3KVA annunciator inverter cabinet, Channel B annunciators are inoperable.

- Annuciator ANNUNCIATOR DIAGNOSTIC ERROR (M6-D, E-4) is half lit.

- All channel B annunciator windows are dark.

- Two hours after shift change the OATC notices all lit channel A annunciator windows change to the dark condition.

Which ONE of the following identifies the required operating crew action to be performed for the given conditions?

A. Trip the Reactor, and GO TO E-0, Reactor Trip or Safety Injection.

B. Trip the Turbine, and GO TO AOP-S.06, Turbine Trip Below P-9 (50% Power).

C Go to AOP-P.08, Loss of Control Room Annuciators, and station personnel for increased monitoring.

D. Go to 0-SO-55-1, Annuciator System, to reboot the ADDS terminal to restore indication.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible since this is a possible event which could lead to an Emergency Plan activation, thus the candidate may determine that a Reactor Trip would be required for a loss of all annuciators.

B. Incorrect, Plausible if the candidate thought that by losing the turbine control annunciators that a Turbine Trip would be required, thus entry into AOP-S.06 would be required.

C. Correct, Based on the indications presented in the stem, there has been a loss of all Unit I Control Room annunciators. This would require the crew to enter AOP-P.08 for a loss of all annunciators. One of the requirements of AOP-P.8 is to station personnel in the control room to monitor parameters in lieu of having alarms available to warn the operators about potential system operational problems or trends.

D. Incorrect, Plausible since O-SO-55-I, is the procedure that is used to diagnose problems associated with the Annuciator System. However, going to the normal procedure would be directed byAOP after the crew entered the AOP. It is not an expectation to go to a normal procedure first before entering an AOP when conditions warrent. Also based on the indications, rebooting the ADDS computer would not likely correct the problem.

Tuesday, November22, 2011 9:13:35 PM 74

QUESTIONS REPORT for SQN JAN 2012 NRC RO Exam Question Number: 74 Tier: 3 Group K/A: G 2.4.32 Knowledge of the operator response to loss of all annunciators.

Importance Rating: 3.6 / 4.0 10 CFR Part 55: 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by having the candidate identify that a loss of all annuciators has occurred based on the indications presented and recall the proper actions for a loss of all control annunciators.

Technical

Reference:

AOP-P.08, Loss of Control Room Annunciators, rev 3 Proposed references None to be provided:

Learning Objective: OPL271AOP-P.08, Obj 2 Question Source:

New Modified Bank Bank X Question History: 1/2008 Audit exam, original question SQN bank question AOP-P.08-B.2 002 Comments: Stem and choices have been changed from Unit 2 to Unit 1 and distractors altered for plausibility Source: BANK MOD Source If Bank: SEQUOYAH BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 9:13:35 PM 75

75. G 2.4.34 075 Given the following plant conditions:

- A fire in the Control Building has forced an evacuation of the Main Control Room due to excessive smoke.

- AOP-C.04, Shutdown from Auxiliary Control Room, has just been entered.

In accordance with AOP-C.04, which ONE of the following identifies both the location the CR0 is directed to go to and what action the CR0 is directed to initially perform?

Location Action A. Aux Control Room Direct AUOs to manually stabilize plant until Controls are transferred to Aux Control Room B. Aux Control Room Ensure personnel are dispatched to perform applicable checklists and appendices using Appendix Z, Task Assigment Sheet.

C AOP-C.04 Cabinet Ensure personnel dispatched to perform applicable checklists and appendices using Appendix Z, Task Assignment Sheet.

D. AOP-C.04 Cabinet Direct AUOs to manually stabilize plant until Controls are transferred to Aux Control Room DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible since that will be the location that the Unit will ultimately be controlled from, however initialily the CR0 is to go to the A0P-C.04 cabinet. Also second part is not correct. It is logical to have the AUOs attempt to locally control equipment until control can be established at the Aux Control Board, however they have specific tasks to perform as identified by the Appendix Z assignment sheet.

B. Incorrect, Plausible since that will be the location that the Unit will ultimately be controlled from, however initiallly the CR0 is to go to the A0P-C.04 cabinet. Also the second part is correct.

C. Correct, In accordance with A0P-C.04, the CR0 is directed to go to the AOP-C.04 cabinet (6.9 Ky Shutdown Board Rm A) and ensure that the Appendix Z Task assignment sheet is completed.

D. Incorrect, Plausible since the first part is correct, however the AUOs have specific tasks to performed as outlined in Appendix Z.

Tuesday, November22, 2011 9:17:48 PM 75

QUESTIONS REPORT for SQN JAN 2012 NRC RD Exam Question Number: 75 Tier: 3 Group K/A: G2.4.34 Knowledge of RD tasks performed outside the main control room during an emergency and the resultant operational effects.

Importance Rating: 4.2 1 4.1 10 CFR Part 55: 41.10 IOCFR55.43.b: not applicable KIA Match: This question matches the K/A by testing the candidates knowledge of the one of the tasks that the CRC is to perform outside the Main Control Room during implementation of AOP-C.04.

Technical

Reference:

AOP-C.04 section 2.1 step 1 Proposed references None to be provided:

Learning Objective: OPL271AOP-C.04 obj. 4 Question Source:

New Modified Bank Bank X Question History: SQN bank question (used on 2009 Audit exam)

Comments:

Source: BANK Source If Bank: SEQUOYAH BANK Cognitive Level: LOWER Difficulty:

Job Position: RO Plant: SEQUOYAH Date: 12/0 1 Last 2 NRC?: NO Tuesday, November 22, 2011 9:17:48 PM 76