ML110140244

Initial Exam 2010-301 Draft RO Written Exam

ML110140244
Person / Time
Site:	Watts Bar
Issue date:	12/10/2010
From:	NRC/RGN-II
To:	Tennessee Valley Authority
References
50-390/10-301
Download: ML110140244 (191)

Text

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

1. 007 EK3.01 001 Given the following plant conditions:

- A reactor trip occurs on Unit 1.

- Off-Site power is lost.

- The crew enters ES-0.1, Reactor Trip Response.

Upon entering ES-0.1, Step 3 directs the operators to monitor for RCS temperature trending to 557°F.

Which ONE of the following identifies the temperature indication the operators will use for monitoring and why?

A. Tcold, to check for natural circulation established.

B Tcold, to ensure adequate RCS heat removal is occurring.

C. Tavg, to check for natural circulation established.

D. Tavg, to ensure adequate RCS heat removal is occurring.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible, because Tcold is the correct indication to use due to no RCPs being in service. Checking for natural circulation is plausible since this is a goal of the procedure, but only towards the end and is not the specific reason at this point in the procedure.

B. Correct, Tcold is the correct indication to use, per ES-O. I because there are no RCPs in service and the reason for the check at this point in the procedure is to ensure there is adequate heat removal from the reactor.

C. Incorrect, Plausible, because Tavg is the parameter to use post trip if the RCPs are in service, but with a loss of offsite power, there is no power to the RCPs.

Checking for natural circulation is plausible since this is a goal of the procedure, but only towards the end and is not the specific reason at this point in the procedure.

D. Incorrect, Plausible, because Tavg is the parameter to use post trip if the RCPs are in service, but with a loss of offsite power, there is no power to the RCPs. Also because the reason for the check at this point in the procedure being to ensure there is adequate heat being removed from the reactor is correct.

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 Question Number: 1 Tier: 1 Group 1 KIA: 007 EK3.01 Reactor Trip Knowledge of the reasons for the following as the apply to a reactor trip:

Actions contained in EOP for reactor trip Importance Rating: 4.0 I 4.6 10 CFR Part 55: 41.5/41.10/45.6 /45.13 10CFR55.43b: Not applicable K/A Match: KA is matched because the question requires the applicant to identify the actions req uried in the EOP as to what parameter to monitor for given conditions and the reason for monitoring the parameter.

Technical

Reference:

ES-0.1, Reactor Trip, Rev 22 Proposed references None to be provided:

Learning Objective: 3-OT-EOP0000

9. Discuss the basis for monitoring RCS temp using T-cold when no RCPs are running as directed by ES-0.1.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History Relocated the correct answer Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

2. 008 AK2.01 002 Consider the following Unit I conditions:

- A safety injection has occurred.

- RCS pressure is 1720 psig and still dropping.

- Pressurizer level initially dropped and is now rising.

- All reactor coolant pumps are in operation.

Which one of the following identifies the leak location?

A Pressurizer safety valve B. Reactor Vessel Head vent line C. Cold Leg Accumulator #1 check valve weld D. Loop 2 Hot Leg temperature instrument well DISTRA CTOR ANAL YSIS:

A. Correct, only a vapor space leak would result in the pressurizer level rising while the pressure continued to drop. Other leaks would result in the level dropping until the pressure stabilized or started to recover.

B. lncorrect with the leak in this location, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the pressurizer level could be rising with a leak on the vessel head vent if SI flow was greater than the leak flow.

C. lncorrect, with the leak in this location, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the pressurizer level could be rising with a leak on the valve weld if SI flow was greater than the leak flow.

D. lncorrect with the leak in this location, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the pressurizer level could be rising with a leak on the vhot leg if SI flow was greater than the leak flow.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 2 Tier: 1 Group 1 K/A: 008 AK2.01 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

Valves Importance Rating: 2.7* / 2.7 IOCFRPart55: 41.7/45.7 IOCFR55.43.b: Not applicable K/A Match: Question requires the knowledge of the interrelations of a pressurizer safety valve leak and conditions in the pressurizer during a vapor space accident Technical

Reference:

WOG E-1 Background document, Rev 2 WOG E-0 Background document, Rev 2 Proposed references None to be provided:

Learning Objective: 3-OT-TAAO13

1. Describe the dynamic behavior of the reactor, from a thermodynamic and hydraulic point of view, following a loss of coolant accident (LOCA) for the following categories:

e. Pressurizer vapor space break.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Point Beach question Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

3. 009 EA2.23 003 Given the following plant conditions:

- Unit 1 was operating at 100% power with the listed equipment out of service and tagged when a LOCA occurred.

- Safety Injection Pump lB-B

- Thermal Barrier Booster Pump IA

- RCS pressure stabilized at 1605 psig.

- Containment pressure has increased to 1.6 psig and slowly rising.

- E-1, Loss of Reactor or Secondary Coolant, is in progress.

Which ONE of the following identifies a condition that would require the RCPs to be removed from service?

A. Thermal Barrier Booster Pump 1 B trips.

B. Containment Pressure increases to 2.1 psig.

C. Safety Injection Pump IA-A fails to start.

D RCS pressure drops to 1480 psig.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the trip of TBBP lB would result in the loss of Thermal Barrier Cooling (an RCP support system) to the RCPs. However, seal injection flow could still be present B. Incorrect, Plausible because the trip of the RCPs is required if Phase B pressure (2.8 psig) is reached due to loss of RCP support systems.

C. Incorreci Plausible because the conditon would result in neither SIP running and the RCP trip critieria does consider status of the SIPs but would require one to be injecting if no CCP was injecting for the RCP trip criteria to be met.

D. Correct, If the RCS pressure dropped to 1480 psig, the RCPs are required to be stopped because the RCP Trip criteria is met (less than 1500 psig and CCP lB-B is injecting.)

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 3 Tier: 1 Group 1 K/A: 009 EA2.23 Small Break LOCA Ability to determine or interpret the following as they apply to a small break LOCA:

RCP operating parameters and limits Importance Rating: 2.8 / 3.3 10 CFR Part 55: 43.5 /45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of RCP operating limits during a small break LOCA.

Technical

Reference:

E-1, Loss of Reactor or Secondary Coolant, Rev 15 Proposed references None to be provided:

Learning Objective: 3-OT-EOPO100

2. Explain the basis for tripping the RCPs in an accident situation given the following conditions:

a. RCS press less than 1500 psig

b. Phase B isolation signal initiated.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN question EOP EOPOIOO 005 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 4.011 EK2.02 004 Unit 1 is operating at 100% power with CSST C out of service:

- A LOCA occurs and the RCS pressure is currently at 300 psig and dropping.

Which ONE of the following identifies the status of the RHR pumps?

A. Both RHR pumps are currently injecting.

B. Only RHR pump I B-B is currently injecting.

C. Only RHR pump IA-A is currently injecting.

D Neither RHR pump is currently injecting.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible if the shutoff head of the RHR pump is incorrectly related to the pressure at which RHR can be placed in service for decay heat removal (350 psig).

B. Incorrect, Plausible because the CSST is the normal feed to shutdown board supplying RHR pump IA-A and with the CSST out of service it can be concluded that the pump is not running leaving only RHR pump lB-B running and capable to inject.

C. Incorrect, Plausible because if the power supplies to the RHR pumps are reversed it can be concluded that the RHR pump 18-B is not running leaving only RHR pump IA-A running and capable to inject.

D. Correct, Neither RHR pump is currently injecting is correct because the RCS pressure is greater than the shutoff head of the RHR pumps but with the RCS pressure continuing to drop the pumps will start to inject when the pressure drops low enough.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 4 Tier: 1 Group 1 K/A: 011 EK2.02 Large Break LOCA Knowledge of the interrelations between the Large Break LOCA and the following:

Pumps Importance Rating: 2.6* I 2.7*

10 CFR Part 55: 41.7/45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the applicant is required to identify the interrelations between Large break LOCA and the RHR pumps.

Technical

Reference:

E-1, Loss of Reactor or Secondary Coolant, Rev 15 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO63A

4. List each subsystem of the ECCS and explain when and how it injects to the core upon initiation of a Safety Injection Signal.

3-OT-EOP01 00

13. Justify the procedure step to shutdown the RHR pumps if the RCS pressure is greater than 150 psig.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

5. O2&i.i 005 Given the following:

- The Unit is at 100% power with letdown valves as shown in the picture below.

- Charging is lost when CCP I B-B trips.

LETD0WNORlF1CE Which ONE of the following identifies how the CVCS letdown valves are closed?

A. Manually close 1-FCV-62-69 and 1-FCV-62-70 which will allow 1-FCV-62-73 to be closed manually.

B. Manually close 1-FCV-62-69 and 1-FCV-62-70 which will allow 1-FCV-62-73 to close automatically.

C Manually close 1-FCV-62-73 which will allow l-FCV-62-69 and 1-FCV-62-70 to be closed manually.

D. Manually close l-FCV-62-73 which will allow 1-FCV-62-69 and 1-FCV-62-70 to close automatically.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because the interlocking scheme is the reverse of this (and could be recalled incorrectly) and both sets of valves do require manual action to close in the conditions in the question.

B. Incorrect, Plausible because the interlocking scheme between the orifice valves and the isolation valves is the reverse of this (and could be recalled incorrectly) and there are condItions that will automatically close 1-FCV-62-69 and 1-FCV-62-70.

C. Correct. 1-PC V-62-73 must be closed first because 1-FCV-62-69 and 1-FCV-62-70 are interlocked such that they will not close if an orifice valve is open. After 1-PC V-62-73 is closed the isolation valves can be manually closed D. Incorrect, Plausible because manually closing 1-FCV-62-73 first is correct and there are conditions that will automatically close 1-FCV-62-69 and 1-FCV-62-70.

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010 Question Number: 5 Tier: 1 Group 1 KIA: 022AA1.O1 Loss of Reactor Coolant Makeup Ability to operate and I or monitor the following as they apply to the Loss of Reactor Coolant Makeup:

CVCS letdown and charging Importance Rating: 3.4 / 3.3 10 CFR Part 55: 41.7/45.5 I 45.6 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of how the letdown valves control switches would be operated to isolate letdown following the Loss of Reactor Coolant Makeup due to a charging pump trip.

Technical

Reference:

AOI-20, Malfunction of Pressurizer Level Control System, Rev 31 1-47W61 1-62-1 R6 Proposed references None to be provided:

Learning Objective: 3-OT-A0I2000

04. Perform required AOl Operator Actions

05. Demonstrate ability/knowledge of AOl, by:

a. Responding to Actions.

b. Responding to Contingencies (RNO).

c. Observe and Interpret Notes and Cautions.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: WBN bank question SYSO62A 007 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 6.025 AG2.1.7 006 Given the following plant conditions:

- The reactor was shutdown 3 weeks ago for a forced outage.

- The RCS has been drained to elevation 719 to support maintenance.

- 2 SIG Hot Leg Manways have been removed.

- No nozzle dams are installed.

- RCS temperature is 140°F.

- A non-recoverable loss of RHR cooling has occurred.

- The operating crew has implemented AOl-i 4, Loss of RHR Cooling.

Which ONE of the following identifies...

(1) the approximate amount of time for core boiling to begin and (2) the feed and bleed method that would be used in accordance with AOl-I 4?

REFERENCE PROVIDED A. (1) 13 minutes (2) Gravity feed to the RCS B(1) 13 minutes (2) Normal Charging to the RCS C. (1) 16 minutes (2) Gravity feed to the RCS D. (1) 16 minutes (1) Normal Charging to the RCS

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because the approximate time to core boil is correct and gravity feed would be the feed and bleed method used if the cold leg manways were removed.

B. Correct In accordance with the chart the approximate time for 21 days after the shutdown is 13 minutes and in accordance with the AOl the method for reed and bleed with hot leg manways removed is by normal charging to the RCS C. Incorrect, Plausible because the approximate time to core boll would be 16 minutes if the RCS level had been above 720.75 and gravity feed would be the feed and bleed method used if the cold leg manways were removed.

D. Incorrect, Plausible because the approximate time to core boll would be 16 minutes if the RCS level had been above 720.75 and the method for feed and bleed with hot leg manways removed is by normal charging to the RCS is correct.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 6 Tier: 1 Group 1 KIA: 025AG2.1.7 Loss of Residual Heat Removal System Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Importance Rating: 4.4 / 4.7 10 CFR Part 55: 41.5 I 43.5 I 45.12 / 45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the ability to evaluate the performance of the plant to determine how long it will take for core boiling to occur based on operating characteristics identified in the question following a loss of RHR and then make an operational judgment of the correct core cooling method to implement if RHR can not be re-established.

Technical

Reference:

AOI-14, Loss of RHR shutdown Cooling, Revision 0036 Proposed references AOI-14, Loss of RHR shutdown Cooling, Rev 0036 to be provided: page 74, Appendix A, Approximate Time To Core Boil Learning Objective: 3-OT-A011400

5. Explain Alternate Cooling Methods Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN questions AOl-i 400.02 001 and A011400.07 005 combined and modified.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 7.029 EA1.08 007 Unit 1 is operating at 55% power with the following conditions:

- 1-Sl-99-10-B, 62 Day Functional Test of SSPS Train B and Reactor Trip Breaker B, is in progress.

- Reactor Trip Breaker (RTB) B is currently open with its bypass breaker BYB closed.

- The Reactor Trip Breaker A (RTA) 1 25v DC control power supply breaker on 125v DC Battery Board 1 trips open.

- A turbine trip occurs but the reactor fails to automatically trip due to failure of SSPS Train A.

Which ONE of the following identifies...

(1) how the reactor will respond to placing the Reactor Trip Switches on 1-M-4 and 1-M-6 to the TRIP position and (2) the indications available on RTA and BYB indicating lights after the reactor is manually tripped?

A. (1) The reactor would trip from actuation of either of the reactor trip switches.

(2) Neither RTA or BYB would have an indicating light lit.

By (1) The reactor would trip from actuation of either of the reactor trip switches.

(2) BYB would have the GREEN indicating light lit but RTA will NOT have an indicating light lit.

C. (1) The reactor would trip from actuation of the reactor trip switch on 1-M-6 only (2) Neither RTA or BYB would have indicating light lit.

D. (1) The reactor would trip from actuation of the reactor trip switch on I -M-6 only.

(2) BYB would have the GREEN indicating light lit but RTA will NOT have an indicating light lit.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the reactor tripping from either switch is correct and the bypass breakers use the same control power supply as the reactor trip breakers. Since BYB is tripped through Train A SSPS its control power could be mistakenly associated with the control power for RTA.

B. Correct The reactor would trip from actuation of either of the 2 reactor trip switches because both switches provide contacts to trip the BYB breaker even though RTA would remain closed due to loss of control power and the Train A SSPS failure.

With its control power breaker tripped RTA would not have any indicating lights lit, but BYB control power circuit is energized and the Green indicating light would be lit.

C. Incorrect, Plausible because the reactor trip switch could be associated with a specific Train because other switches on the board are train specific and the bypass breakers use the same control power supply as the reactor trip breakers.

Since BYB is tripped through Train A SSPS its control power could be mistakenly associated with the control power for RTA.

D. lncorrect Plausible because the reactor trip switch could be associated with a specific Train because other switches on the board are train specific and BYB having the green indicating light lit while RTA has no light lit is correct.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 7 Tier: 1 Group 1 K/A: 029 EA1.08 Anticipated Transient Without Scram (ATWS)

EA1 Ability to operate and monitor the following as they apply to a ATWS:

Reactor trip switch pushbutton Importance Rating: 4.5 I 4.5 10 CFR Part 55: 41.7/45.5/45.6 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge expected result when the manual reactor trip switches (there are no reactor trip push buttons in the control room) are used in response to a failure of the reactor to automatically trip and how they function to trip the reactor along with knowledge of the indicating lights used to monitor status of the reactor trip breakers during off normal conditions.

Technical

Reference:

FR-S.1, Nuclear Power Generation / ATWS, Rev 20 45W600-99-1 R6 Proposed references None to be provided:

Learning Objective: 3-0T-FRS0001

3. List from memory and in order the two Immediate Operator Actions for procedure FRS. 1, Nuclear Power Generation/ ATWS, and discuss the basis for each action.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

8. 038 EK1.03 008 Given the following:

- A steam generator tube rupture has occurred on S/G #4.

- Both manual reactor trip and safety injection signals have been initiated.

- A coincidental loss of support systems has caused the RCPs to be stopped.

- The crew has implemented E-3, Steam Generator Tube Rupture, and is currently cooling down to the target core exit temperature.

Which ONE of the following identifies...

(1) the flow rate through the RCS Loop 4 relative to flow rate through the intact RCS Loops and (2) why a Pressurized Thermal Shock (PTS) RED path could be reached based on Tcold in the ruptured loop?

Li) L)

A. lower than because of the reverse flow in the loop B lower than because of the ECCS flow being injected C. the same as because of the reverse flow in the loop D. the same as because of the ECCS flow being injected

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANALYSIS:

A. Incorrect. Plausible because the RCS loop in loop 4 being lower than the RCS flow in the intact loops is correct and reverse loop flow can cause off normal loop temperatures to occur in different conditions.

B. Correct, Prior to the cooldown, the ruptured steam generator is isolated by closing the MSIV and terminating AFW flow. This will cause a decrease in (possibly cessation of) RCS natural circulation flow in Loop 4. Since cold ECCS is being injected into Loop 4 cold leg with reduced circulation (or no flow) flow temperature will drop severely.

C. Incorrect, Plausible because there are no loop valves to stop the flow and if the steam generator isolation status is not consider it can be concluded that the flow would be the same and reverse loop flow can cause off normal loop temperatures to occur in different conditions.

D. Incorrect, Plausible because there are no loop valves to stop the flow and if the steam generator isolation status is not consider it can be concluded that the flow would be the same and because cool ECCS flow is being injected into a stagnant loop a pressurized Thermal Shock RED path can develop, but FR-P. 1 implementation would be delayed until the ECCS flow was termThated

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 8 Tier: 1 Group 1 K/A: 038 EK1.03 Steam Generator Tube Rupture (SGTR)

Knowledge of the operational implications of the following concepts as they apply to the SGTR:

Natural circulation Importance Rating: 3.9/4.2 IOCFRPart55: 41.8/41.10/45.3 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of an operational implication that can arise during performance of a rapid cooldown in accordance with E-3 while on natural circulation.

Technical

Reference:

E-3, Steam Generator Tube Rupture, Rev. 22 WOG Executive Volume. Rev. 2 Natural Circulation and Stagnant Loops Chapters.

Proposed references None to be provided:

Learning Objective: 3OT-EOP0300

8. Given a set of plant conditions, evaluate the conditions to determine if natural circulation exists and take appropriate action to initiate, restore, or maintain natural circulation.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank X Bank Question History: WBN question EOPO300.05 013 modified (used on 2008 SRO exam)

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

9. O4OAK1.05 009 Plant conditions are as follows:

- A reactor startup is in progress following a five week Refueling Outage.

- MTC is at the maximum value allowed by Technical Specifications without requiring Rod Withdrawal Limits to be established.

- The reactor is critical with:

- NI-I 35 lxi 02% power and stable.

- NI-I 36 lxi Q2% power and stable.

- Subsequently, a steam line break of 3% of rated steam flow occurs.

Assuming no operator actions, which ONE of the following describes the response of the reactor?

A. The reactor will go subcritical.

B. The reactor will remain at the current power level.

C. Reactor power will stabilize at the Point of Adding Heat.

D Reactor power will rise and stabilize at approximately 3% power.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the reactor would go subcritical if the maximum allowable MTC had been a positive number.

B. Incorrect, Plausible because the reactor would at the current power level if the maximum allowable MTC had zero (0 zik/k/°F).

C. Incorrect, Plausible because with the maximum allowable MTC required to be less than 0 zlk!k/°F (rod withdrawal limits are required to be established prior to allowing MTC to reach 0) when the point of adding heat is reached the coefficients will start to provide negative reactivity to offset the power increase but pwoer will rise until the reactivity is balanced.

D. Correct, the maximum MTC allowed by Tech Specs without establishing rod withdrawal limits is -0. 60X10-5 zlk/k/°F. This results in the MTC being negative. A steam line break will cause a positive reactivity insertion due to the cooldown and reactor power will increase to match the demand.

Question Number: 9

• Tier: 1 Group 1 040 AK1 .05

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 KIA: 040 AK1.05 Steam Line Rupture Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture:

Reactivity effects of cooldown Importance Rating: 4.1/4.4 IOCFRPart55: 41.8/41.10/45.3 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to identify the operational implications of the reactivity change due to a cooldown is it applies to a steam line rupture.

Technical

Reference:

Tech Spec 3.1.4, Moderator Temperature Coefficient -

MTC Unit I Cycle 10 Core Operating Limits Report, Rev 3 Proposed references None to be provided:

Learning Objective: 3-OT-T/S0301

2. Determine the bases for the limits placed on reactor core measured parameters (SDM, MTC).

3. Given plant parameters/conditions, correctly determine the compliance with the LCOs or TRs in the Reactivity Control sections of T/S and T/R manuals.

5. Determine the bases for the limits placed on control rod positioning and position monitoring equipment (Rod Insertion Limits, Alignment Limits, and Rod Position Indicating Systems).

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Question modified from Ginna 2008 exam question Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

10. 054 AG2.1.23 010 Given the following plant conditions:

- Unit 1 was operating at 100% power when the lB MFP tripped.

- The operating crew stabilized the unit in accordance with AOl-16, Loss of Normal Feedwater.

- The plant responded as expected and all AOl-16 actions have been completed, with the exception of repairing the lB MFP.

Later in the shift, 6.9kV Unit Board ID trips and locks out.

Which ONE of the following identifies the required action to be taken as a result of the loss of the unit board?

A. If turbine load is greater than 900 MWe, then reduce turbine load to 900 MWe by lowering the EHC Reference Control, in accordance with AOl-37, Turbine Runback Response.

B. If turbine load is 800 MWe or greater, then reduce turbine load to within MFWP capability by lowering the EHC Reference Control, in accordance with AOl-i 6.

C. If turbine load is greater than 900 MWe, then reduce turbine load to 900 MWe with the Valve Position Limiter, in accordance with AOl-37, Turbine Runback Response.

D If turbine load is 800 MWe or greater, then reduce turbine load to within MFWP capability with the Valve Position Limiter in accordance with AOl-i 6.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because there are conditions where the load would be reduced with the EHC Reference control and because the loss of 6.9KV Unit Board ID would also result in a loss of the IC 3# Heater Drain Tank Pump. For the loss of a

1. 3 Heater Drain Tank Pump which results in a runback, AOI-37 is the correct procedure to be used. This AOl directs turbine load to be reduced to 900 MWe if only one pump is available but in this case 2 are available.

B. Incorrect, Plausible because there are conditions where the load would be reduced with the EHC Reference control and the AOI-16 step requiring actions if the turbine load is 800 MWe or greater is correct except the reduction would be with the VPL.

C. Incorrect, Plausible because dropping load with the VPL is correct, and the loss of 6.9KV Unit Board ID would also result in a loss of the IC 3# Heater Drain Tank Pump. For loss of a #3 Heater Drain Tank Pump which results in a runback, AOI-37 is the correct procedure to be used. This AOl directs turbine load to be reduced to 900 MWe if only one pump is available but in this case 2 are available.

D. Correct The plant would be at approximately 80% power or 1000 MWe after the trip of the lB MFP. The loss of 6.9KV Unit Board ID would result in a trip of the Standby MFP, requiring a load drop with the Valve Position limiter (VPL) to ensure the turbine load was within the MFWP capability. After the load reduction with the VPL the crew would monitor SG levels to ensure that levels were returning to normal.

Question Number: 10 Tier: 1 Group 1 KIA: 054AG2.1.23 Loss of Main Feedwater (MEW)

Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Importance Rating: 4.3 I 4.4 10 CFR Part 55: 41.10 I 43.5 I 45.2 I 45.6 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the applicant to know the effects of losing equipment and then apply the knowledge to the procedure that would be applicable and the actions required by the procedure.

Technical

Reference:

AOl-16, Loss of Normal Feedwater, Rev 32 AOl-37, Turbine Runback Response, Rev 14

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010 Proposed references None to be provided:

Learning Objective: 3-OT-A011600

7. Demonstrate ability/knowledge of AOl, by correctly:

a. Recognizing Entry conditions

b. Responding to Action steps

c. Responding to Contingencies (RNO)

d. Responding to Notes and Cautions Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Question modified from WBN Bank question 054 AK3.04 017. Correct answer relocated and wording changed, one distractorchanged, stem reformatted but not significantly modified.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

11. 055 EK3.02 011 Which ONE of the following is a purpose of depressurizing all intact steam generators (S/Gs) to 300 psig during the performance of ECA-0.0, Loss of Shutdown Power?

A. Reduces differential pressure across S/G U-tubes to minimize RCS inventory loss in the event of a tube rupture.

B Reduces differential pressure across RCP seals to minimize leakage and loss of RCS inventory.

C. Maximizes natural circulation flow before reflux cooling begins as the RCS becomes saturated.

D. Maximizes natural circulation flow to allow reactor vessel head to cool since CRDM cooling fans are unavailable.

DIS TRACTOR ANALYSIS:

A. Incorrect, the most likely failure for this event is loss of inventory through failed RCP seals not SGTR. Plausible because a failure potential does exist and limiting the depressurization to 300 psig would limit the differential pressure across the tubes.

B. Correct depressurizing the steam generators willcause RCS pressure to be lowered, thus reducing potential for a seal LOCA by reducing the driving force across the RCP seals.

C. Incorrect, steaming is a method to increase natural circ and would occur, however minimizing inventory loss is a greater concern at this point.

Plausible because the action will cause increased natural circulation.

D. Incorrect, steaming is a method to increase natural circ and would occur, however minimizing inventory loss is the greater concern at this point.

Plausible because the action will cause increased natural circulation.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 11 Tier: 1 Group 1 K/A: 055 EK3.02 Loss of Offsite and Onsite Power (Station Blackout)

Knowledge of the reasons for the following responses as the apply to the Station Blackout:

Actions contained in EOP for loss of offsite and onsite power Importance Rating: 4.3 I 4.6 10 CFR Part 55: 41.5/41.10145.6/45.13 IOCFR55.43.b: Not applicable K/A Match: Question requires the applicant to identify the reason for an action taken during the performance of a procedure implemented during a station blackout.

Technical

Reference:

EcA-o:o, Loss of Shutdown Power, Rev 20 WOG Background Document for ECA-0.0, Rev 2 Proposed references None to be provided:

Learning Objective: 3-OT-ECA0000

3. Explain why the intact S/Gs are depressurized to 300 psig during performance of ECA-0.0.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Wolf Creek Bank question Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

12. 057 G2.4.1 012 Given the following:

- Unit I is operating at 14% power.

- The loss of a 120v AC vital Instrument Power Board results in the operators initiating action to decrease reactor power to within the capability of the AFW system.

- During the power decrease the OAC manually trips the reactor using handswitch 1 -RT-1, Reactor Trip, after an automatic trip signal failed.

A loss of which ONE of the following I 20v AC Vital Instrument Power Boards required the reactor power decrease to within AFW capability and how many Immediate Operator Actions (lOAs) would be required by the first Emergency Operating Procedure performed in responding to the event?

BOARD IOAs A. 1-lI 2 B 1-Il 4 C. 1-IV 2 D. 1-IV 4

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the loss of I2OVAC Vital Instrument Power Boards 1-Il is correct and because with the reactor failing to trip automatically, the applicant could conclude the use of FR-S. I which has only 2 Immediate Operator Action steps would be appropriate.

B. Correct, AOI-25.02, Loss of I2OVAC Vital Instrument Power Boards 1-Il or 2-Il, has directions to reduce power to within the AFW system capability and because the reactor was tripped from a MCR reactor trip switch, the crew will enter and perform E-O which contains 4 Immediate Operator Action steps.

C. Incorrect Plausible because while main feedwater control remains available, the AOl for the loss of I2OVAC Vital Instrument Power Boards I-IV identifies the need for manual control and because with the reactor failing to trip automatically, the applicant could conclude the use of FR-S. 1 which has only 2 Immediate Operator Action steps would be appropriate.

D. Incorrect, Plausible because while main feedwater control remains available, the AOl for the loss of I2OVAC Vital Instrument Power Boards I-IV identifies the need for manual control and because the crew entering and performing E-O which contains 4 Immediate Operator Action steps is correct.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 12 Tier: 1 Group 1 K/A: 057 G2.41 Loss of Vital AC Electrical Instrument Bus 2.4.1 Knowledge of EOP entry conditions and immediate action steps.

Importance Rating: 4.6 / 4.8 IOCFRPart55: 41.10/43.5/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of how a Loss of Vital AC Electrical Instrument Bus affects unit operation and knowledge of immediate actions steps in the EOPs.

Technical

Reference:

AOl-25.02, Loss of 120V AC Vital Instrument Power Boards 1-Il or 2-Il, Revision 0030 AOl-25.04, Loss of 120V AC Vital Instrument Power Boards 1-IV or 2-IV, Revision 0027 E-0, Reactor Trip or Safety Injection, Revision 28 FR-S.1, Nuclear Power Generation / ATWS, Rev 20 Proposed references None to be provided:

Learning Objective: 3-OT-A012500

1. [Demonstrate ability to recognize a loss of any 120V AC Vital Power Bd, including effects on equipment and controls (SOER 8 1-02)].

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question Corn rnents:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

13. 058 AA2.02 013 Unit 1 is operating at 100% power when the following annunciator alarms:

Window 18A - 125V VITAL CHGRJBATT II ABNORMAL Which ONE of the following identifies...

(1) where the battery board voltage can be determined, and (2) how long the battery is rated to supply the design basis loads while maintaining a minimum terminal voltage of 105v DC in accordance with the ARI?

A. (1) Locally at the battery board, pjjjy (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. (1) Locally at the battery board, gjy (2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C (1) Both from the Main Control Room and locally at the battery board (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. (1) Both from the Main Control Room and locally at the battery board (2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because there are batteries in the plant that require local determination of voltage ancluding 125v Vital Battery V and the Diesel Generator 125v Batteries and because being able to supply the loads for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in correct.

B. Incorrect, Plausible because there are batteries in the plant that require local determination of voltage (including 125v Vital Battery V and the Diesel Generator 125v Batteries and because the batteries do have a rating of 2320 amp hours over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

C. Correct, The battery board voltage can be determined both from instrumentation on 1-M-1 in the MCR or locally at the battery board. The ARI states in a Note that the 125V DC Vital Battery II is able to supply design basis loads for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while maintaining a minimum terminal voltage of 105V DC.

D. Incorrect, Plausible because the battery board voltage can be determined both from instrumentation on 1-M-1 in the MCR or locally at the battery board and because the batteries do have a rating of 2320 amp hours over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 13 Tier: 1 Group 1 K/A: 058 AA2.02 Loss of DC Power Ability to determine and interpret the following as they apply to the Loss of DC Power:

1 25V dc bus voltage, low/critical low, alarm Importance Rating: 3*3* / 3.6 10 CFR Part 55: 43.5 / 45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the ability to determine the 125V dc bus voltage from plant instrumentation and the knowledge to predict how long the plant safety related batteries have the capacity to maintain the minimum terminal voltage as identified in the alarm response instruction for the alarm due to low voltage.

Technical

Reference:

ARI-15-21, CNTL PWE & Fire PROT, Revision 0024, Window 18-A Proposed references None to be provided:

Learning Objective: 3-OT-SYSO57P

16. Correctly locate control room controls and indications associated with the 125v DC Vital system, including:

a. Alarms

b. Voltmeters

c. Ammeters Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO57P.10 001 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

14. 062 AA2.03 014 With Unit I operating at 100% power the following occurs:

- ERCW supply header IA ruptures in the yard.

- AOl-i 3, Loss of Essential Raw Cooling Water, is implemented to isolate the leak.

When the appropriate section of the AOl is complete, which ONE of the following identifies...

(1) the ERCW supply header that will be supplying water to Auxiliary Building components that are supplied from the 1A Supply Header and (2) how the cooling on the A Train diesel generators (DG5) is affected?

A. (1) ERCW Supply Header2A (2) Only DG IA-A will be supplied from its alternate supply B. (1) ERCW Supply Header 2A (2) Both DG IA-A and 2A-A will be supplied from their alternate supplies.

C. (1) ERCW Supply Header 2B (2) Only DG IA-A will be supplied from its alternate supply Dv (1) ERCW Supply Header 2B (2) Both DG lA-A and 2A-A will be supplied from their alternate supplies

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible because there are valves to crosstie the 1A and 2A ERCW supply headers but these are to allow for strainers maintenance and their use would not isolate the leak. Also, because only one of the four supply headers is affected it could be concluded that only one DG would be affected.

B. lncorrect Plausible because there are valves to crosstie the IA and 2A ERCW supply headers but these are to allow for strainers maintenance and their use would not isolate the leak and because the affect on the DG cooling is correct.

C. Incorrect, Plausible because ERCW supply header 2B supplying the components is correct and with only one of the four supply headers being affected it could be concluded that only one DG would be affected.

D. Correct, AOl-13 will align ERCW supply header 2B to the IA header in the Auxiliary Building to provide a supply of water to the components and because both Train A diesel generators are normally supplled from ERCW supply header IA prior to the header entering the Auxiliary Building, both will be required to be aligned to the alternate ERCW supply.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 14 Tier: 1 Group 1 KIA: 062 AA2.03 Loss of Nuclear Service Water Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:

The valve 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: KA is matched because the questions requires knowledge of ERCW flow paths via valve lineup to restore ERCW flow to affected equipment after a ruptured header is isolated.

Technical

Reference:

AOI-13, Loss of Essential Raw Cooling Water, Rev 38 1 -47W845-1 R56 1 -47W845-2 R76 Proposed references None to be provided:

Learning Objective: 3-OT-A0l1300

8. Demonstrate ability/knowledge of AOl, to correctly:

a. Recognize Entry conditions.

b. Respond to Action steps.

c. Respond to Contingencies (RNO column).

d. Respond to Notes and Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN Bank question AOl-I 300.07 003 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

15. 065 AK3.03 015 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- 1-FCV-32-110, Non-Essential Air to Containment, fails closed causing the air header inside containment to depressurize.

Assuming NO operator action is taken, which ONE of the following Reactor Trip signals will occur first?

A. Overtemperature AT (OTAT)

B High Pressurizer Level C. Low Pressurizer Pressure D. Low Steam Generator Level DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the pressurizer spray valves are also air operated and inside containment. However, they are supplied from the different headers entering containment and the valves fall closed. If the valves were affected and the pressure was dropping, the OTDT setpoint would be dropping. Also if the affect on the letdown valves was reversed additional valves would open; the pressurizer level would drop causing the heaters to trip and the pressure to start dropping.

B. Correct the loss of air will cause letdown to isolate. This causes pressurizer level to start rising. Charging will back down to the minimum, but this only slows down the rate that the pressurizer level is rising. The level will rise to the reactor trip setpoint over time.

C. Incorrect, Plausible because the pressurizer spray valves are also air operated and inside containment. However, they are supplied from the different headers entering containment and the valves fall closed. If they falled open, the pressure would be dropping. Also if the affect on the letdown valves was reversed zdditional valves would open the pressurizer level would drop causing the heaters to trip and the pressure to start dropping.

D. lncorrect Plausible because the steam generators are inside containment several valves affecting the steam generator level are air operated. However none of the valves are inside containment, thus not supplied from the lost air supply.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 15 Tier: 1 Group 1 K/A: 065 AK3.03 Loss of Instrument Air Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air:

Knowing effects on plant operation of isolating certain equipment from instrument air Importance Rating: 2.9 I 3.4 10 CFR Part 55: 41.5,41.10/45.6 /45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the applicant must identify the equipment that loses air from the stem conditions and how the loss of air to the equipment affects plant operation.

Technical

Reference:

1-47W848-i R24 1 -47W848-9 Ri 5 AOl-b, Loss of Control Air, Rev 0039 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO32A

17. Given a loss of non-essential air, explain why a reactor trip will occur Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Question modified from a Ginna Bank question used on their 2008 audit exam.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

16. W/EO4EK1.3 016 Unit us in Mode 3 with RCS stable at 557°F and 2235 psig when the following occurs:

- A rapid drop in pressurizer pressure causes a Safety Injection to occur.

- Pressurizer level is dropping.

- Containment conditions are not changing.

- The following annunciators alarm:

Window 174-B RR-90-1 AREA RAD HI Window 185-B RR-90-12 PARTICULATE RAD HI Which ONE of the following identifies...

(1) the event in progress and (2) the operational implications of the operator checks and actions in support of the initial high level action of the mitigating procedure?

Event Operational Implication A. LOCA outside containment Ensure valves normally open are sequentially cycled closed to determine if the leak can be isolated B LOCA outside containment Ensure status of normally closed valves to prevent subjecting low pressure piping to high pressure conditions.

C. Steam generator tube rupture Ensure overfill of the ruptured steam generator is prevented to reduce aggravating the release of radiation.

D. Steam generator tube rupture Ensure the ruptured steam generator is isolated to prevent loss of RCS subcooling during rapid cooldown.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the conditions describing a LOCA outside containment is correct and identifying and isolating the break is the second of the major actions in the procedure as well as being the first action in the SGTR procedure.

B. Correct, The conditions describe a LOCA outside containment and the first of the major actions in the mitigating procedure is for the operators to verify proper valves alignment to determine if a low pressure system being subjected to high pressure is the cause to loss of coolant (which is the most probable cause of the condition)

C. lncorrect, Plausible because a SGTR would result in a rapid drop in pressurizer pressure, decreasing pressurizer leve) and the normal containment conditions identified in the stem. Also because a SGTR will result in radiation monitor alarms and the annunciators in alarm do not identify which radiation monitor is causing the alarm. Identifying and isolating the ruptured steam generator is the first of the major actions in the SGTR mitigation procedure.

D. Incorrect, Plausible because a SGTR would result in a rapid drop in pressurizer pressure, decreasing pressurizer level and the normal containment conditions identified in the stem. Also because a SGTR will result in radiation monitor alarms and the annunciators in alarm do not identify which radiation monitor is causing the alarm. Cooldown to establish RCS subcooling margin is a major action in the SGTR mitigation procedure and the steam generator must be isolated prior to the cooldown step.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 16 Tier: 1 Group 1 KIA: W/E04 EK1 .3 LOCA Outside Containment Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment)

Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Outside Containment).

Importance Rating: 3.5 / 3.9 IOCFRPart55: 41.8/41.10,45.3 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of annunciators and conditions resulting from a LOCA outside containment and the priority of the remedial actions required to be implemented in response to the conditions.

Technical

Reference:

ECA-i .2 LOCA Outside Containment, Rev 4 E-3 and ECA-i .2 WOG Background Documents, Rev 2 ARI-i 80-i 87, Common Radiation Detectors, Rev 31 ARI-i 73-179, u-i Radiation Detectors, Rev 45 Proposed references None to be provided:

Learning Objective: 3-OT-STG-ECA1

01. Identify and explain the major actions of procedures ECA-i.1 and 1.2.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

17. W/E05 EA1.2 017 Given the following:

- A reactor trip occurred on Unit I and no AFW supply could be established due to equipment failures.

- A transition to FR-H.1 Loss of Secondary Heat Sink, was required.

- The crew is establishing Main Feedwater flow using the Standby Main Feed Pump.

Which one of the following identifies the condensate pumps that would be in service to support the supply of feedwater using the Standby Main Feed Pump in accordance with FR-H.1?

A Hotwell pumps, gjjy B. Hotwell pumps and Cond Demin Booster pumps, gjijy C. Hotwell pumps and Condensate Booster pumps, pjjjy D. Hotwell pumps, Cond Demin Booster pumps, and Condensate Booster pumps.

DISTRA CTOR ANAL YSIS:

A. Correct; During performance of FR-H. I if main feedwater flow is to be established using the Standby Feed Pump to restore the heat sink, the only condensate pumps to be started are the Hotwell Pumps to prevent an over pressure condition. This is identified in a Note preceding the step to establish main feed water flow. The note states If the standby feed pump will be used only the Hotwell Pumps should be started to prevent an overpressure condition.

B. Incorrect; Plausible because the Cond Demin Booster Pumps are used to supply the steam generators if a feedwater pump is not available during performance of FR-H. 1.

C. Incorrect, Plausible because the Condensate Booster Pumps are used to supply the steam generators if a feedwater pump is not available during performance of FR-H. I. also, CBPs are used to support operation lof a TD Main Feed pump is being used.

D. Incorrect; Plausible because the Cond Demin Booster Pumps and the Condensate Booster Pumps are used to supply the steam generators if a feedwater pump is not available during performance of FR-H. I.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 17 Tier: 1 Group 1 KIA: W/E05 EAI .2 Loss of Secondary Heat Sink Ability to operate and I or monitor the following as they apply to the (Loss of Secondary Heat Sink)

Operating behavior characteristics of the facility.

Importance Rating: 3.7 / 4.0 10 CFR Part 55: 41.7 I 45.5 / 45.6 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the knowledge of an operating behavior characteristics of the facility as identified in a Note in FR-H.1 stating If the standby feed pump will be used, only the hotwell pumps should be started to prevent an overpressure condition.

Technical

Reference:

FR-H.1, Loss of Secondary Heat Sink, Rev 17 Proposed references None to be provided:

Learning Objective: 3-OT-FRH0001

17. Given a set of plant conditions, use FR-H.1, H.2, H.3, H.4, & H.5 and the Critical Safety Function Status Trees to correctly diagnose and implement:

Action Steps, RNOs, Foldout Pages, Notes and Cautions.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

18. W/E11 EK2.1 018 Given the following conditions:

- With Unit 1 initially operating at full power a large break LOCA occurred.

- Containment pressure peaked at 7.3 psig and is currently 3.2 psig and slowly dropping.

- Neither RHR pump could be started.

- Both Containment Spray Pumps are running.

- RWST level = 66%.

- Containment sump level = 24%.

- The crew has transitioned to ECA-1 .1, Loss of RHR Sump Recirculation, and is at the table in Step 4 to determine the proper containment spray pump alignment and operation.

Which ONE of the following will result in the proper alignment of the containment spray pumps under existing plant conditions?

A. Reset the Containment Spray Signal, stop both Containment Spray Pumps and close their discharge valves, then place their handswitches in A-AUTO.

B. Reset the Containment Spray Signal, stop both Containment Spray Pumps and close their discharge valves, align suction to the containment sump and place their handswitches in A-AUTO.

C Stop and Pull To Lock one Containment Spray Pump, close its discharge valve. Allow the remaining containment spray pump to continue to run taking suction from the RWST.

D. Continue to run both containment spray pumps until RWST level is less than or equal to 8%, then stop and Pull To Lock both containment spray pumps and close their discharge valves.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect Plausible because this is the normal process for terminating containment spray in the emergency procedures.

B. Incorrect, Plausible because the alignment of the Containment Spray Pumps to the containment sump is directed in the procedure (Steps 5-7)

C. Correct, ECA-1. I Step 4 determines the number of Containment Spray Pump running based on containment pressure. With containment pressure between 2.0 and 13.5 psig, the procedure requires one pump running and directs the other spray pump to be stopped, its control switch placed in Pull-to-Lock and the discharge valve closed.

D. Incorrect, Plausible because a Caution at the beginning of the procedure states IF RWST level drops to 8%, then any ECCS or Containment Spray Pump taking suction from the RWST must be stopped.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 18 Tier: 1 Group 1 K/A: W/E11 EK2.1 Loss of Emergency Coolant Recirculation Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Importance Rating: 3.6 / 3.9 10 CFR Part 55: 41.7/45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge how safety related systems will be manually aligned to perform their required function during a Loss of Emergency Coolant Recirculation Technical

Reference:

ECA-1 .1, Loss of RHR Sump Recirculation, Rev 11 Proposed references None to be provided:

Learning Objective: 3-OT-ECAO1O1 No specific objective identified Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: SQN question with modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

19. 003AG2.4.1 019 Which ONE of the following identifies a condition where entry into E-0, Reactor Trip or Safety Injection, would be required?

A. Shutdown Bank A Group I step counter fails while Unit 1 is operating at 4%

power.

B. Shutdown Bank A Group I and Group 2 step counters fail while Unit 1 is operating at 80% power.

C. Shutdown Bank A control rod D2 drops while Unit 1 is operating at 100%

power.

D Shutdown Bank A control rods D2 and B12 drop while Unit 1 is operating at 30% power.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the condition would have caused the reactor to be tripped and E-0 entered if the unit had been in Mode 3 instead of Mode 2 in accordance with Technical Requirements.

B. Incorrect, Plausible because the condition would require a unit shutdown in accordance with T/S 3.0.3 and would have caused E-0 to be entered if the unit had been in Mode 3 instead of Mode 1 due to the required reactor trip in accordance with Technical Requirements.

C. Incorrect, Plausible because the dropped rod would require action in accordance with A 01-2 to change reactor power. One action required by the operating crew is to reduce power to less than 75%.

D. Correct, A reactor trip and entry into E-0 Reactor Trip and Safety Injection, is required if two or more rods drop in accordance with AOI-2, Malfunction of Rod Control System.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 19 Tier: 1 Group 2 K/A: 003 Dropped Control Rod 2.4.1 Knowledge of EOP entry conditions and immediate action steps.

Importance Rating: 4.6 / 4.8 IOCFRPart55: 41.10/43.5/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of conditions that require EOP entry conditions due to control rod problems (with dropped rods being the corrrect answer.)

Technical

Reference:

AOl-2, Malfunction of Rod Control System, Rev 37 Tech Spec LCO 3.1.8, Rod Position Indication, Amendment 58 Tech Requirements 3.1.7, Position Indication System -

Shutdown, Amendment 58 Proposed references None to be provided:

Learning Objective: 3-OT-A010200

12. Given a set of plant conditions, use the AOl to correctly:

a. Recognize Entry Conditions.

b. Identify Required Actions.

c. Respond toContingencies (RNO).

d. Observe and Interpret Cautions and Notes.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO85 001 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

20. 005 AK1.05 020 Given the following:

- Unit 1 trips from 100% power.

- Shutdown Bank A rod M2 sticks at its fully withdrawn position.

- Control Bank C rod K10 sticks at 70 steps withdrawn.

- All other Shutdown and Control rods are completely inserted.

Which ONE of the following identifies how the shutdown margin calculation (SDM) to verify adequate boron concentration to meet the SDM requirement is affected during the performance of 1-S1-0-10, Shutdown Margin, hand calculations?

A. The stuck rods do not affect the way the SDM is calculated after the trip because after the trip the calculation is done using refueling boron concentration which has enough conservatism to account for the Maximum Stuck Rod Worth.

B. Because the Maximum Stuck Rod Worth is already included in the SDM calculation, the worth of rod M2 does not affect the SDM calculation but the actual worth of rod K10 being stuck at 70 steps withdrawn is required to be determined and included in the calculation.

Cv The Maximum Stuck Rod Worth determined using the applicable NuPOP Table is required to be multiplied by two because there are two rods that are stuck during the performance of the SDM calculation.

D. Because the Maximum Stuck Rod Worth is already included in the SDM calculation, the worth of rod M2 does not affect the SDM calculation but due to rod KI 0 being stuck at 70 steps withdrawn, a 600 pcm reduction is required in the SDM calculation.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorreci Plausible because there is a conservative method for determining the required boron concentration that does used the refueling boron concentration in the calculation but the process cannot be used if actual rods are stuck out of the core B. Incorrect, Plausible not to account for rod M2 because the Technical Specification definition of Shutdown Margin assumes the highest worth rod is stuck out of the core and accounting for the worth of rod K1O would offset the lack of negative reactivity the rod should be providing.

C. Correct During performance of the shutdown margin hand calculation, the procedure requires that the maximum Stuck Rod Worth be multiplied by the number of stuck rods (in this case doubled) during the calculation of the required boron concentration to ensure the required shutdown margin exist.

D. Incorrect, Plausible not to account for rod M2 because the Technical Specification definition of Shutdown Margin assumes the highest worth rod is stuck out of the core and there is a condition where 600 pcm of conservatism is accounted for in the SDM calculation.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 20 Tier: 1 Group 2 K/A: 005 AK1.05 Inoperable/Stuck Control Rod Knowledge of the operational implications of the following concepts as they apply to Inoperable I Stuck Control Rod:

Calculation of minimum shutdown margin Importance Rating: 3.3 / 4.1 IOCFRPart55: 41.8/41.10/45.3 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of how the reactivity associated with stuck control and/or shutdown rods are accounted for in the minimum calculation for shutdown margin.

Technical

Reference:

1-Sl-0-10, Shutdown Margin, Rev. 0022 Proposed references None to be provided:

Learning Objective: 3-CT-S IP1 000

4. Identify Parameters that affect or are used to calculate SDM.

5. Perform, or Evaluate an SDM Calculation for correctness Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

21. 032 AK2.01 021 Which ONE of the following identifies...

(1) the I 20v AC Vital Power supply to Source Range Monitor NI-I 32 and (2) if the power supply failed causing a loss of audible count rate signal inside containment while the plant was in Mode 6, the 1 -M-1 3 switch that would be required to be repositioned in accordance with AOl-4, Nuclear Instrumentation Malfunctions, to restore the containment signal?

Power Suprly 1-M-13 switch A 120v AC Vital Board 1-Il AMPLIFIER SELECT B. 12OvAC Vital Board I-Il CHANNEL SELECTOR C. 12OvAC Vital Board 1-IV AMPLIFIER SELECT D. 120v AC Vital Board I-IV CHANNEL SELECTOR DISTRACTOR ANAL YSIS:

A. Correct; Source Range Monitor N/-132 is supplied from the I2OVAC Vital Power Board 1-Il and if the audible audio count rate inside containment is loss then the AMPLIFIER SELECT switch is required to be repositioned.

B. Incorrect, Plausible because the I2OVAC Vital Power Board 1-Il is correct and the CHANNEL SELECTOR switch is repositioned if the audible audio count rate inside the Main Control Room is lost.

C. Incorrect, I2QVAC Vital Power Board 1-IVis a Train B (as is NI-132) power supply and if the audible audio count rate inside containment is loss then AMPLIFIER SELECT switch is required to be repositioned.

D. Incorrect, I2OVAC Vital Power Board 1-IV is a Train B (as is Nl-132) power supply and the CHANNEL SELECTOR switch is repositioned if the audible audio count rate inside the Main Control Room is lost.

Question Number: 21 Tier: 1 Group 2 K/A: 032 AK2.O1

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 K/A: 032 AK2.01 Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the following:

Power supplies, including proper switch positions Importance Rating: 2.7*! 3.1 10 CFR Part 55: 41.7/ 45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the power supply to a source range monitor and how source range instrumentation switch positions affect the operation of the monitor functions when the power supply is lost.

Technical

Reference:

1-47W611-99-2, R12 45N706-2 R21 45N1652-4 R5 AOI-25.02, Loss of 120V AC Vital Instrument Power Boards 1-Il or 2-Il, Rev. 0030 AOl-4, Nuclear Instrumentation Malfunctions, Revision 0029 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO92A

31. Describe the distribution of Instrument and Control Power in the Nuclear Instrumentation System, including the effects of a loss of one or both supplies under various plant conditions.

3-OT-A010400

3. Describe Operator Actions for an SRM failure.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank questions SYSO92A.10 046 and SYSO92A.09 002 combined and modified.

Comments:

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010

22. 033 AA2.03 022 Given the following:

- Unit 1 is at 30% power when Intermediate Range Monitor NI-i 35 fails due to one blown fuse.

Which ONE of the following NI-I 35 drawer indications identifies a condition where the IR high flux reactor trip signal could NOT be manually bypassed using the Level Trip Switch and list the Tech Spec LCD(s) that would be currently applicable due to the blown fuse?

Note:

LCO 3.3.1, Reactor Trip System (RTS) Instrumentation LCO 3.3.3, Post Accident Monitoring (PAM) Instrumentation A Control Power light is DARK; LCD 3.3.3, only B. Control Power light is DARK; LCD 3.3.1 and LCD 3.3.3 C. Instrument Power light is DARK; LCD 3.3.3, only D. Instrument Power light is DARK; LCD 3.3.1 and LCO 3.3.3

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Correct, The CONTROL POWER ON light being DARK would indicate the blown fuse would prevent the blocking of the trip signal from the drawer and while Tech Spec 3.3.1 is not applicable at the current power levei Tech Spec 3.3.3 would require a shutdown after 30 days.

B. Incorrect, The CONTROL POWER ON light being DARK would indicate the blown fuse would prevent the blocking of the trip signal from the drawer but while Tech Spec 3.3.3 is applicable, 3.3.1 is not applicable. Plausible because the light being DARK is correct and Tech Specs 3.3.1 would be applicable at lower power levels.

C. lncorrect The INSTRUMENT POWER ON light being DARK would not indicate the blown fuse would prevent the blocking of the trip signal from the drawer and while Tech Spec 3.3.3 would require a shutdown after 30 days. Plausible because there is a fuse that if blown would cause the light to be DARK and Tech Spec 3.3.3 being applicable is correct.

D. Incorrect, The INSTRUMENT POWER ON light being DARK would not indicate the blown fuse would prevent the blocking of the trip signal from the drawer and while Tech Spec 3.3.3 is applicable, 3.3.1 is not applicable. Plausible because there is a fuse that if blown would cause the light to be DARK and Tech Specs 3.3.1 would be applicable at lower power levels.

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 Question Number: 22 Tier: 1 Group 2 KIA: 033 AA2.03 Loss of Intermediate Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

Indication of blown fuse Importance Rating: 2.8 / 3.1 10 CFR Part 55: 43.5/ 45.13 IOCFR55.43.b: 2 K/A Match: Applicant is required to determine the indication of a blown control power fuse on an Intermediate Range Monitor.

Technical

Reference:

AO1-4, Nuclear Instrument Malfunctions, Rev 0029 1-SI-92-132, 3lDay Channel Operational Test and Full Power Alignment of Source, and Intermediate Range Neutron Flux Channel II, Rev 19 Technical Specifications, 3.3.1, Technical Specifications, 3.3.3, Amendment 72 Proposed references None to be provided:

Learning Objective: 3-OT-A010400

4. Explain indications, Auto actions, and Operator actions for an Intermediate Range Monitor (IRM) failure Question Source:

New Modified Bank X Bank Question History: SQN bank question 033AA203 084 used on SRO exam modified for RD at WBN Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

23. 036 G 2.4.46 023 Unit I is in a refueling outage with core reload in progress, when the following occurs:

0905 - The Main Control Room is notified of an alarm sounding in the Incore Instrument Room.

0907 - An AUO and RADPRO are dispatched to the room.

0912 - Annunciator window 128-A, SEP LEVEL Hl/LO alarms.

0915 - Annunciator window 174-A, 1-RR-90-1 AREA RAD HI alarms.

0915 - CR0 reports:

1-RM-90-59, UPPER CONTAINMENT, count rate rising.

All other Rad monitors with input to window 174-A are normal.

- Fuel Handling SRO notified of the radiation concern and fuel movement stopped.

Which ONE of the following events would cause the above conditions and how should 1-ISV-78-600, Fuel Transfer Tube Isolation, be aligned for the conditions?

A. Leakage on the Reactor Cavity Seal; 1-ISV-78-600 would remain open and SEP makeup initiated to maintain cavity level.

B Leakage on the Reactor Cavity Seal; 1-ISV-78-600 would be closed to protect the spent fuel pit level.

C. RCS leak on an Incore Detector 5 path transfer device; 1-lSV-78-600 would remain open and SEP makeup initiated to maintain cavity level.

D. RCS leak on an Incore Detector 5 path transfer device; 1-ISV-78-600 would be closed to protect the spent fuel pit level.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because leakage on the reactor cavity seal is correct and initiating make-up to the spent fuel pit is required but 1-IS V-78-600 would not be left open.

B. Correct, A high level condition in the Containment Pit Sump causes an alarm in the Incore Instrument Room. The sump high level would be caused by the reactor vessel cavity seal leaking. This would result in the level in the cavity and the spent fuel pit dropping. Lower level in the cavity would allow radiation to increase in upper containment. Conditions represent entry conditions for A 01-29 which will direct the closing of 1-IS V-78-600 and the spent fuel pit level returned to normal via the initiation of make-up.

C. Incorrect, Plausible because leakage on a 5-path transfer device would cause the loss of inventory and would cause the area rad monitor in the Incore Instrument Room to alarm. However, this radiation monitor also inputs to the MCR area radiation monitor alarm and the question identifies the monitor is reading normal.

Also because initiating make-up to the spent fuel pit is required which would supply water to the reactor cavity if 1-IS V-78-600 was left open.

D. Incorrect, Plausible because leakage on a 5-path transfer device would cause the loss of inventory and would cause the area radiation monitor in the Incore Instrument Room to alarm. However, this radiation monitor also inputs to the MCR area radiation monitor alarm and the question identifies the monitor is reading normal. Also, because closing 1-IS V-78-600 is correct.

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010 Question Number: 23 Tier: 1 Group 2 K/A: 036 G 2.4.46 Fuel Handling Incidents Ability to verify that the alarms are consistent with the plant conditions.

Importance Rating: 4.2 I 4.2 10 CFR Part 55: 41.10/43.5/45.3/45.12 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the alibilty to use information from resultingconditions to determine that a fuel handling incident (cavity seal failure) is occuring.

Technical

Reference:

AOI-29, Dropped orDamaged Fuel or Refueling Cavity Seal Failure, Revision 20 Proposed references None to be provided:

Learning Objective: 3-OT-A012900

6. Demonstrate ability/knowledge of AOl, to correctly:

a. Recognize Entry conditions.

b. Respond to Action steps.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

24. 037 AA1.04 024 Given the following:

- Unit I is operating at 100% power when a small SGTL is determined to exist but the leak rate has not been determined.

- AOl-33, Steam Generator Tube Leak, is entered and Appendix A, Steam Generator Tube Leak Monitoring, has been implemented

- 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the initiation of AOI-33, Appendix A, Data Sheet 1, the reactor trips due to all turbine throttle valves going closed.

- Annunciator window 175-E, VAC PMP EXH 1-RM-119 INSTR MALE alarms.

Which ONE of the following identifies...

(1) the required frequency for recording information on Data Sheet 1 of the AOl and (2) the condition resulting from the reactor trip that would cause the instrument malfunction alarm?

A (1) every 15 minutes.

(2) Rise in condenser backpressure B. (1) everyl5 minutes.

(2) Generator PCBs opening C. (1) every30 minutes.

(2) Rise in condenser backpressure D. (1) every3o minutes.

(2) Generator PCBs opening

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANAL YSIS:

A. Correct, The data that is entered in the RM-90-119 column on Data Sheet 1 is a calculated leak rate taken from The CHEM7 screen on ICS and is required to be logged every 15 minutes if the leak rate is not known and in accordance with the ARI for annunciator window 175-E, VAC PMP EXH 1-RM-1 19 INSTR MALF a rise in condenser back pressure following a reactor trip will cause the alarm.

B. lncorrecj, Plausible because the data entered in the RM-90-1 19 column on Data Sheet I being required to be logged every 15 minutes if the leak rate is not known is correct and switching in the switchyard is a condition stated in the AR! as having the potential to cause annunciator window 175-B, VAC PMP EXH 1-RM-1 19 RAD H! to alarm.

C. lncorrecl, Plausible because 30 minutes is a time identified during performance of Data Sheet 1 and 30 is also a number used in the data sheet relative to leakage rates and a rise in condenser backpressure due to a reactor trip is a condition identified in the ARI as causing annunciator window I 75-E, VAC PMP EXH 1-RM-1 19 INSTR MALF to alarm.

D. lncorrect Plausible because 30 minutes is a time identified during performance of Data Sheet I and 30 is also a number used in the data sheet relative to leakage rates and switching in the switchyard is a condition stated in the ARI as having the potential to cause annunciator window 175-B, VAC PMP EXH 1-RM- 119 RAD HI to alarm.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 24 Tier: 1 Group 2 K/A: 037AA1.04 Steam Generator (SIG) Tube Leak Ability to operate and I or monitor the following as they apply to the Steam Generator Tube Leak:

Condensate air ejector exhaust radiation monitor and failure indicator Importance Rating: 3.6 I 3.9 10 CFR Part 55: 41.7 / 45.5 / 45.6 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the ability to monitor and trend the condenser vacuum pump exhaust radiation monitor and recognize a condition that would cause a failure of the monitor.

Technical

Reference:

AOl-33, Steam Generator Tube Leak, Revision 32 ARI-1 73-1 79, U-i Radiation Detectors, Rev. 0045 windows 175B and 175-E 105 screen, Chem7 Proposed references None to be provided:

Learning Objective: 3-OT-A013300

9. Identify the methods available to quantify a SIG Tube Leak as given in AC 1-33.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: SQN question 037 AA1 .04 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

25. 061 AK2.01 025 Unit 1 is operating at 100% power when the following occurs:

- Annunciator Window 174-B, 1-RR-90-1 AREA RAD HI alarms.

Which ONE of the following identifies...

(1) the location(s) where a rate meter is located that will indicate the radiation level being sensed by the Area Radiation Monitor (ARM) causing the alarm and (2) what type alarm(s) is/are available locally at the ARM?

A. (1) In the Main Control Room, only; (2) An audible alarm, only B. (1) In the Main Control Room, only; (2) Both an audible alarm and an indicating light C. (1) Both in the Main Control Room and locally at the monitor; (2) An audible alarm, only D (1) Both in the Main Control Room and locally at the monitor; (2) Both an audible alarm and an indicating light

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because area radiation monitor O-RM-90-135, Main Control Room area radiation monitor detector is behind the control boards and this monitor is not equipped with a local rate meter. However this monitor does not input to this alarm. Also, because there are plant alarms that are audible only.

B. lncorrec1, Plausible because area radiation monitor O-RM-90-135, Main Control Room area radiation monitor detector is behind the control boards and this monitor is not equipped with a local rate meter. However this monitor does not input to this alarm. Also, because the monitors inputting to this alarm having both visual and audible alarms locally is correct.

C. Incorrect, Plausible because the monitors having rate meters both in the Main Control Room and locally at the monitor is correct. Also, because there are plant alarms that are audible only.

D. Correct All of the Area Radiation Monitors that cause this alarm have rate meters on O-M-12 in the Main Control Room and also locally at the monitor. Each of the monitors is also equipped with both audible and visual indication of HIGH radiation conditions locally at the detector location.

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010 Question Number: 25 Tier: 1 Group 2 KIA: 061 AK2.01 Area Radiation Monitoring (ARM) System Alarms Knowledge of the interrelations between the Area Radiation Monitoring (ARM) System Alarms and the following:

Detectors at each ARM system location Importance Rating: 2.5*! 2.6*

10 CFR Part 55: 41.7/45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the interrelations of Area Radiation Monitoring (ARM) System Alarms and the detectors at each ARM system location.

Technical

Reference:

ARI-1 73-179, u-i Radiation Detectors, Rev. 0045 Window 174-B 1-47W610-90-2 R51 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO9OA

04. Identify MCR panel where Area Radiation monitors read out.

06. Identify 10 areas that used Area Radiation Monitors (ARMs)

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

26. 076 2.O2 026 Unit I is operating at 100% power when the following occurs:

- Chem-lab reports the activity level in the RCS is now 0.315 pCi/gram Dose Equivalent lodine-131 and the activity level has risen from a steady state value of 0.085 iCi/gram on the previous sample taken 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago.

- The operating crew is in the process of establishing the letdown flow rate required by AOl-28, High Activity in the Reactor Coolant.

Which ONE of the following identifies the status of Tech Spec LCO 3.4.16, RCS Specific Activity, and how the operation of the CVCS Mixed Beds will be affected due to the letdown flow rate established?

A. LCO 3.4.16 entry currently required; The second Mixed Bed will be placed in service.

B. LCD 3.4.16 entry NOT currently required; The second Mixed Bed will be placed in service.

Cv LCD 3.4.16 entry currently required; Only one of the CVCS Mixed Beds will be in service.

D. LCD 3.4.16 entry NOT currently required; Only one of the CVCS Mixed Beds will be in service.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANALYSIS:

A. Incorrect Plausible because Tech Spec LCO 3.4.16 entry being currently required because the Dose Equivalent lodine-131 activity is greater than 0.265 pmCiigram limit and the capacity of the CVCS Cation Bed is only 75 gpm which could be mistaken for the capacity of a Mixed Bed.

B. IncorrecI, Plausible because the Tech Spec LCO 3.4.16 entry value could mistakenly be thought to be 2lpmCiYgram which is the value in the Tech Spec that requires a shutdown/cooldown to be initiated and the capacity of the CVCS Cation Bed is only 75 gpm which could be mistaken for the capacity of a Mixed Bed.

C. Correct, Tech Spec LCO 3.4.16 entry is currently required because the Dose Equivalent lodine-131 activity is greater than 0.265 pmCiigram limit and the letdown flow would be raised to 120 gpm which is the rated flow of each of the CVCS Mixed Beds.

D. Incorrect, Plausible because the Tech Spec LCO 3.4.16 entry value could mistakenly be thought to be 2lpmCi/gram which is the value in the Tech Spec that requires a shutdown/cooldown to be initiated and only one CVCS Mixed Bed being in service is correct because the letdown flow would be raised to 120 gpm which is the rated flow of each of the CVCS Mixed Beds.

Question Number: 26 Tier: 1 Group 2 KIA: 076 AA2.02 High Reactor Coolant Activity Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity:

Corrective actions required for high fission product activity in RCS Importance Rating: 2.8 / 3.4 10 CFR Part 55: 43.5 / 45.13 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of the corrective actions required due to increasing RCS activity levels including the Tech spec applicability, letdown flow rate control, and CVCS mixed bed operation in accordance with the Abnormal oeprating Instruction.

Technical

Reference:

AOl-28, High Activity in Reactor Coolant System,

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Technical

Reference:

AOl-28, High Activity in Reactor Coolant System, Revision 21 Tech Spec LCO 3.4.16,RCS Specific Activity, Amendment 41,55 Proposed references None to be provided:

Learning Objective: 3-OT-A012800

4. Explain how activity is reduced if activated corrosion/erosion products are the reason for Hi activity.

7. Given a set of plant conditions, use the AOl to correctly:

b. Identify Required Actions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question 3-OT-AOl-2800.07 009 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

27. W!E02 EK3.1 027 Unit I was operating at 100% power when the following occurred:

- Operators initiated a Reactor Trip and Safety Injection due to flow past a S/G #2 safety valve.

- Actions were taken in accordance with the emergency instructions and the crew is now performing ES-i .1, SI Termination.

- After the first CCP has been stopped, the BIT isolated and normal Charging established, a procedure steps reads:

IF any SIG Faulted, THEN DO NOT CONTINUE this Instruction UNTIL Faulted SIG depressurization stops.

Which ONE of the following identifies the basis for delaying the perfomance of subsequent steps?

A. Because the RCS cooldown may result in requiring the use FR-P.i, Pressurized Thermal Shock, which has less restrictive SI termination criteria.

B. Because if the SIG pressure stablizes greater than 0 psig, it is indicative of a SIG tube rupture requiring use of ECA-3.1, SGTR and LOCA Subcooled Recovery.

C Because the RCS inventory balance and adequacy of ECCS injection flow cannot be accurately determined while the faulted S/G continues to depressurize.

D. Because the final S/G pressure is used to determine the maximum RCS pressure allowed to prevent exceeding the 1600 psid AP limit across the SIG tubes.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the subsequent steps are removing ECCS pumps and with the SG blowdown cooling the RCS, an FR-P. 1 entry could be required where the SI termination criteria does change.

B. Incorrect, Plausible because the if a SG were ruptured the SG would not completely depressurize and if detected the ECA would be the correct procedrue.

C. Correct, The status of the unit is such that charging flow has been established and the operator is trying to control charging flow to maintain pressurizer level. If the faulted SG is continuing to depressurize the RCS inventory balance and adequacy cannot be accurately determined. So the procedure is not continued while the SG blowdown continues because the subsequent steps will evaluate and remove (if conditions allow) the ECCS pumps while ensuring that the operator has control of the RCS conditions.

D. Incorrect, Plausible because there is a 1600 psid limit discussed in a different application and with the SG continuing to blowdown the delta pressure would be rising.

Question Number: 27 Tier: 1 Group 2 KIA: W/E02 EK3.1 SI Termination Knowledge of the reasons for the following responses as they apply to the (SI Termination)

Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

Importance Rating: 3.3/ 3.6 10 CFRPart 55: 41.5/41.10, 45.6, 45.13 IOCFR55.43b: Not applicable K/A Match: KA is matched because the question contains knowledge of the reasons for operating limitations and operating characteristics related to RCS pressure and temperature of while performing SI termination.

Technical

Reference:

ES-li, SI Termination, Rev. 0016 WOG Backgound Documents, Rev 2

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 to be provided:

Learning Objective: 3-OT-EOP01 00

8. Given a set of plant conditions, use E-1, ES-i .1, ES-i .2, ES-i .3, and ES-i .4 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes, and Cautions.

11. Explain the basis for waiting for a faulted S/G to complete depressurization before checking RCS press stable or increasing following the establishment of normal charging and prior to stopping any running SI pumps.

16. Discuss the requirement to check RCS subcooling greater than 65°F prior to RCS depressurization.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RD NRC Exam as Submiffed 7/2/2010

28. 003 K5.04 028 Given the following plant conditions:

- The Unit is at 40% power.

- RCP #1 trips.

Assuming no operator action, which ONE of the following identifies the immediate effect the RCP trip will have on indicated steam generator #1 pressure and level?

SG #1 Pressure SGs #1 Level A. Increase Increase B. Increase Decrease C. Decrease Increase D Decrease Decrease DISTRA CTOR ANAL YSIS:

A. lncorrect, Plausible because the steam flow from the steam generator stops and typically when steam flow is stopped in a flow path, the steam pressure rises and because if the steam flow dropping is related to steam flow feed flow mismatched then too much feedwater will cause level to rise. However the immediate effect is for the steam generator level to drop due to void collapse.

B. lncorrect, Plausthie because the steam flow from the steam generator stops and typically when steam flow is stopped in a flow path, the steam pressure rises and because the SG#1 level dropping is correct.

C. Incorrect, Plausible because the steam generator pressure dropping is correct and if the steam flow dropping is related to steam flow feed flow mismatched then too much feedwater will cause level to rise. However the immediate effect is for the steam generator level to drop due to void collapse.

D. Correct, Without operator action, if the RCP #1 trips with the unit at 40% power the lack of heat input to the steam generator will cause less heat to be transferred by SG #1 and more heat transferred by the other three steam generators. Overall steam pressure will drop with steam flow from SG#1 going to zero. This results in less heat being transferred to the secondary causing steam pressure to drop The lack of heat transfer and the collapse of the voids in SG #1 causes the drop in pressure and also results in the indicated level in the steam generator dropping which leads to overfeeding if no action is taken.

Question Number: 28

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Tier: 2 Group 1 K/A: 003 K5.04 Reactor Coolant Pump System (RCPS)

Knowledge of the operational implications of the following concepts as they apply to the RCPS:

Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow, and feed flow Importance Rating: 3.2 I 3.5 IOCFRPart55: 41.5/45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of how secondary parameters (steam pressure and steam generator level) are affected when an RCP trips (is shutdown) during power operations.

Technical

Reference:

FSAR Ch 15, section 15.2.5.1 Westinghouse Transient Accident Analysis AOl-24, RCP Malfunctions During Pump Operation, Revision 0029 Proposed references None to be provided:

Learning Objective: 3-OT-TAAO15

1. Discuss loss of coolant flow theory

a. Derive the time dependent behavior of Th

b. Define DNR and DNBR 3-OT-A012400

12. Describe basic operator actions to shutdown an RCP.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question AOI-2400 001 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

29. 003 1(6.14 029 Given the following plant conditions:

- RCS heat-up in progress.

- The Shift Manager has directed the crew to start the first RCP.

Which ONE of the following sets of parameters identifies conditions that will allow start of the RCP in accordance with SOl-68.02, Reactor Coolant Pumps?

RCS Pressure (psig) Seal mi. Flow (cjpm) VCT Pressure (psig)

A. 310 9 35 330 12 19 C. 350 11 13 D. 370 7 25 DISTRACTOR ANAL YSIS:

A. lncorrect, RCS pressure is below the minimum limit in SOI-68.02 but plausible because the number is higher than the 200 psid number required for the RCP seal differential pressure.

B. Correct, All of the identified parameters are within the required values for the pump start.

C. IncorrecL, VCT pressure is below the limit in SOI-68.02 but plausible because 13 is a number associated with the VCT.

D. Incorrect, Seal injection flow is below the allowable limit in SOl-68.02 but plausible because there are times when performing GO-I that the seal flow is only required to be 6 gpm.

Question Number: 29 Tier: 2 Group 1 K/A: 003 K6.14 Reactor Coolant Pump Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:

Starting requirements.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Importance Rating: 2.6 I 2.9 IOCFRPart55: 41.7/45/5 IOCFR55.43.b: Not applicable KIA Match: Question presents applicant with parameters for starting RCPs, some of which are outside of allowable bands (malfunctions), and requires applicant to determine which set of parameters would allow start of the RCP based on knowledge of starting requirements.

Technical

Reference:

SOl-68.02, Reactor Coolant Pumps, Rev. 0033 GO-i, Unit Startup From Cold Shutdown To Hot Standby, Revision 0066 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68B

5. Describe the RCPs Seal Injection System, including:

a. FlowpathlComponents

b. Flowrate

c. Purpose 3-OT-GOOl 00

7. Explain, as described in GO-i, the operating precautions for the Reactor Coolant Pumps (RCPs).

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Question used on WBN 2006 exam (4 exams ago) with several of the parameter values in choices changed and the correct answer relocated.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

30. 004 K6.36 030 c Given the following:

- Unit is operating at 100%.

- Pressure transmitter 1-PT-62-81 that controls 1-PCV-62-81, Letdown Pressure Control Valve, fails high.

Which ONE of the choices completes the statement below?

1-PCV-62-81 will go full causing A. open; the letdown relief valve to the PRT to open B open; flashing to occur in the letdown line C. closed; the letdown relief valve to the PRT to open D. closed; flashing to occur in the letdown line DISTRACTOR ANALYSIS:

A. Incorrect Plausible if the location of the relief is mistaken because the pressure control valve will go to the open position which would cause pressure to increase downstream of the pressure control valve in the letdown line.

B. Correct, the transmitter sensing pressure upstream of the valve and maintains pressure at setpoint. If the pressure sensed is high the valve will open to pass more flow to drop the pressure. When transmitter falls high the valve will come full open which drops the pressure resulting in flashing in the letdown line.

C. Incorrect, Plausible because the valve would go to the closed position if the transmitter had been sensing pressure downstream of the valve instead of upstream which would stop flow cause pressure to increase and open the relief valve.

D. Incorrect, Plausible because the valve would go to the closed position if the transmitter had been sensing pressure downstream of the valve instead of upstream which would stop flow causing pressure to decrease and cause flashing downstream of the valve.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 30 Tier: 2 Group 1 K/A: 004 K6.36 Chemical and Volume Control System Knowledge of the effect of a loss or malfunction on the following CVCS components:

Letdown pressure control to prevent RCS coolant from flashing to steam in letdown piping Importance Rating: 2.9 I 3.1 IOCFRPart55: 41.7/45.7 IOCFR55A3.b: Not applicable K/A Match: K/A is matched because the question requires the applicant to understand the operation of the letdown pressure control valve PCV-62-81 and be able to identify the effect of a malfunction of the transmitter controlling the valve.

Technical

Reference:

1-47W61 1-62-3 R9 1-47W809-1 R59 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO62A

7. Explain the function and operation of the letdown pressure control valve PCV-62-81.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

31. 005 K5.01 031 Given the following:

- Unit I is being cooled down for a refueling outage.

- Due to the computer program being unavailable, manual performance of 1-SI-68-44, RCS Temperature/Pressure Limits and Pressurizer Temperature Limits, is required.

- RHR Train A has been placed in service in accordance with SOl-74.01, Residual Heat Removal System.

Which ONE of the following identifies the temperature that will be recorded to ensure compliance with the RCS limits if the RCPs are required to be stopped and the primary operational concern if the limits are violated?

At. RHR temperature on the Train A RHR heat exchanger inlet.

Making an existing vessel beitline crack more susceptable to a brittle failure.

B. RHR temperature on the Train A RHR heat exchanger inlet.

Creating a higher potential to develop a new crack in the reactor vessel head.

C. RHR temperature on the Train A RHR heat exchanger outlet.

Making an existing vessel beltline crack more susceptable to a brittle failure.

D. RHR temperature on the Train A RHR heat exchanger outlet.

Creating a higher potential to develop a new crack in the reactor vessel head.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. Correct, 1-SI-68-44 requires the in-service RHR train heat exchanger inlet temperature to be recorded every 15 minutes and the primary concern due to violating the temperature limits is an existing vessel beitline crack undergoing a brittle failure.

B. Incorrect, Plausible because using the RHR heat exchanger inlet temperature is correct and because with the loss of the RCPs bypass flow through the reactor vessel head is lost and the industry has recently experienced several issues/concerns with reactor vessel head cracking.

C. Incorrect, Plausible because the temperature at the RHR heat exchanger outlet is a measure of the water entering the RCS cold legs and because the primary concern due to violating the temperature limits is that an existing vessel beitline crack could undergoing a brittle failure is correct.

D. Incorrect, Plausible because the temperature at the RHR heat exchanger outlet is a measure of the water entering the RCS cold legs and because with the loss of the RCPs bypass flow through the reactor vessel head is lost and the industry has recently experienced several issues/concerns with reactor vessel head cracking.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 31 Tier: 2 Group 1 KIA: 005 K5.01 Residual Heat Removal System (RHRS)

Knowledge of the operational implications of the following concepts as they apply the RHRS:

Nil ductility transition temperature (brittle fracture)

Importance Rating: 2.6 /2.9 IOCFRPart55: 41.5/45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the parameters used to monitor RCS cooldown after RHR is in service if the RCPs are not running and the operational implications of not maintaining the RCS within limits established.

Technical

Reference:

N3-68-4001, Reactor Coolant System, Rev. 0028 1-SI-68-44, RCS Temperature/Pressure Limits and Pressurizer Temperature Limits, Rev. 0008 Control Panel photograph Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68A

18. Given a set of RCS temperature and pressure conditions and the Technical Specifications, determine if the pressure/temperature limitations of the Technical Specifications are met.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question FRP0001 .08 003 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

32. 006 K2.01 032 Given the following conditions:

- Unit us in Mode 4 with Train A RHR in service.

Which ONE of the choices below completes the following statement?

The RHR pump IA-A motor supply breaker is located on (1) and following a loss of offsite power the RHR pump (2) be sequenced back on after a time delay.

Li) L)

A. 480V Shutdown Board lA-IA will B. 480V Shutdown Board IA-IA will NOT C. 6.9kV Shutdown Board IA-A will D 6.9kV Shutdown Board lA-A will NOT DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the other components of the RHR system are supplied from this board (as well the board supplying other pump motors) and the RHR pump does restart automatically under conditions different than as stated in the stem.

B. Incorrect, Plausible because the other components of the RHR system are supplied from this board (as well the board supplying other pump motors) and the RHR pump not restarting automatically is correct.

C. Incorrect, Plausible because the power supply board is correct and the RHR pump does restart automatically under conditions different than as stated in the stem.

D. Correct, the board that supplies the power to the RHR pump motor is 6.9 kV Shutdown Board IA-A and during a blackout the pump will be stripped from the board and will not automatically restart. For the RHR pump to automatically restart, a Safety Injection Signal is required to be present.

08/2010 Watts Bar RD NRC Exam as Submiffed 7/2/2010 Question Number: 32 Tier: 2 Group 1 K/A: 006 K2.01 Emergency Core Cooling System (ECCS)

Knowledge of bus power supplies to the following:

ECCS pumps Importance Rating: 3.6 / 3.9 IOCFRPart55: 41.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the applicant to have knowledge of the bus power supply to an ECCS pump.

Technical

Reference:

SDl-74.01, Residual Heat Removal System, Rev 0055 1-47W61 1-74-2 R5 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO74A

07. Describe the RHR pumps, including power supply, logic, and capacity

09. State the conditions that will result in an auto start of the RHR pumps in accordance with the RHR System Description.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank questions SYSO63A.15 003 and SYSO74A.07 003 modified by combining and establishing stem conditions.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

33. 006 K6.13 033 Given the following:

0900 - A Large Break LOCA occurs on Unit 1.

0910 - ECCS flows are:

1-Fl-63-170, BIT FLOW, is 530 gpm.

1-Fl-63-151, SI PMPA FLOW, is 520 gpm.

1-Fl-63-20, SI PMP B FLOW, is 510 gpm.

1 -Fl-63-91 B, RHR TO CL 2 & 3 WR FLOW, is 2700 gpm.

1-Fl-63-92B, RHR TO CL 1 & 4 WR FLOW, is 3250 gpm.

- RWST Level is 57% on all 4 level indicators.

- Containment Sump level is 50% on all 4 level indicators.

If RHR pump lA-A trips at 0910, which ONE of the following identifies the ECCS flow indicator(s) that will indicate a lower flow?

A. 1-Fl-63-91B, only B. l-Fl-63-92B, only C 1-Fl-63-91B and l-Fl-63-92B, only D. All 5 of the ECCS flow indicators

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because if the pump tripped after the RHR crosstie valves were closed during the alignment process for cold leg recirculation mode of operation, the IA-A RHR pump tripping would affect result in only 1-Fl-63-91B indicating a lower flow.

B. Incorreci Plausible because if the pump tripped after the RHR crosstie valves were closed during the alignment process for cold leg recirculation mode of operation, the IA-A RHR pump tripping would affect result in only one of the RHR flow indicators and which pump supplies which cold legs could be reversed.

C. Correct, both 1-FI-63-91B and 1-FI-63-92B will indicate a lower flow because the RHR heat exchanger outlets are cross- tied to allow either RHR pump to inject into all four cold legs (as required by Tech Specs.) The other three indicators would not drop as they would be taking suction from the RWST at this time and independent of the RHR pumps.

D. Incorrect, Plausible because if the pump tripped later in the event with ECCS in cold leg recirculation mode of operation, an RHR pump tripping would affect all of the flow indicators.

Simulator data with malfunction THQI (Hot leg break) at 100% after 10 minutes.

Before pump trip After pump trip I-Fl-63-170 - 530 gpm 530 gpm I-Fl-63-151 - 520 gpm 520 gpm 1-FI-63-20 - 510 gpm 510 gpm 1-FI-63-91B - 2700 gpm 1000 gpm 1-Fl-63-92B - 3250 gpm 2400 gpm RWST level - 57%

Cnmt Sump level - 50%

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 33 Tier: 2 Group 1 KIA: 006 K6.13 Emergency Core Cooling System (ECCS)

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

Pumps Importance Rating: 2.6 I 2.9 IOCFRPart55: 41.7/45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the applicant to determine how the loss of an ECCS pump will affect the flow rate indicated on various ECCS flow indicators.

Technical

Reference:

SOl-63.01, Safety Injection System, Rev. 0044 ES-i .3, Transfer To Containment Sump, Rev. 17 1-47W811-1 R52 i-47W810-1 R18 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO63A

24. Given a set of plant conditions, determine the correct response of the Emergency Core Cooling System.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

34. 007 A1.01 034 Which ONE of the following completes the statement below relative to the operation of the Reactor Coolant Drain Tank (RCDT) pumps while the operator is lowering the level in the Pressurizer Relief Tank (PRT) in accordance with SOl-68.01, Reactor Coolant System Pressurizer Relief Tank Operations?

When 1-FCV-68-310, PRT Drain to RCDT is opened, the (1) will auto-start and if the RCDT level drops to the low level setpoint while the PRT is being drained, the pump will (2)

A. RCDT Pump A continue to run B RCDT Pump B continue to run C. RCDT Pump A auto-stop D. RCDT Pump B auto-stop There are two installed RCDT pumps, A (50 gpm) and B (150 gpm). RCDT pump A does not have the auto-start feature when 1-FCV-68-310 is opened. RCDT pump B does auto start when the PRT drain valve 1-FCV-68-310 opens.

RCDT pump A auto-stops at the RCDT low level setpoint.

RCDT pump B auto-stops at the RCDT low level setpoint IF the PRT drain valve is closed.

DISTRA CTOR ANAL YSIS:

A. lncorrect, Incorrect regarding auto-start, correct regarding auto-stop.

B. CorrecI, Correct regarding both auto-start and auto-stop.

C. Incorrect, Incorrect regarding auto-start, incorrect regarding auto-stop.

D. Incorrect, Correct regarding auto-start, incorrect regarding auto-stop.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 34 Tier: 2 Group 1 K/A: 007A1.01 Pressurizer Relief/Quench Tank Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including:

Maintaining quench tank water level within limits.

Importance Rating: 2.9 I 3.1 IOCFRPart55: 41.5/45.5 IOCFR55.43.b: Not applicable KIA Match: Question requires applicant to recall automatic control logic for valves and pumps used to lower quench tank (PRT) water level.

Technical

Reference:

SOl-68.01, Reactor Coolant System Pressurizer Relief Tank Operations, Rev. 0046 1-47W611-68-1 RIO l-57W61 1-77-1 R8 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68C

22. Explain the operation of major system components.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Question used on the WBN 2006 exam. Question changed to a fill in the blank. Correct answer relocated.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

35. 008 A2.05 035 Given the following:

- Unit I is operating at 100% power with a known 3 gpm leak on CCS pump IA-A discharge:

- Unit I CCS surge tank level is indicating 68% on both the A and the B sides of the tank.

- The control air header in the Auxiliary Building rapidily depressurized due to a large leak and the header is isolated.

Which ONE of the following identifies the impact of the loss of air to 1-LCV-70-63, UI SURGE TANK MAKEUP LCV and action to control/mitigate the event?

The surge tank level would...

A. rise resulting in CCS water being spilled out the surge tank vent valve until the LCV is closed or isolated.

B. rise resulting in the automatic closing of the surge tank vent to prevent loss of water from the system until the LCV is closed or isolated.

C. continue to drop on the A side of the tank with level on the B side remaining at 68% until a gas bottle is installed to allow opening of the valve.

D continue to drop to less than 68% on both sides of the tank until a gas bottle is installed to allow opening of the valve.

DISTRA CTOR ANALYSIS:

A. IncorrecI, Plausible because the level would be high if the valve had failed open and would continue to rise until water did come out the vent on the top of the tank.

B. Incorrect, Plausible because the level would be rising if the valve had failed open due to loss of air and the Surge tank vent valve does get an automatic close signal but it is from radiation detected in the system, not due to tank level.

C. Incorrect, Plausible because the valve will be prevented from opening by the loss of air and if the level had been below the top of the divider, the leak would be causing the level to be dropping on only the A side of the tank.

D. Correc1, the valve will prevented from opening by the loss of air and with the Surge Tank level at 68%, the level will be dropping on both sides of the tank because the level is above the top of the baffle (divider) in the tank (5 7%).

Question Number: 35

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Tier: 2 Group 1 K/A: 008 A2.05 Component Cooling Water System (CCWS)

Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Effect of loss of instrument and control air on the position of the COW valves that are air operated Importance Rating: 33* / 3*5 10 CFR Part 55: 41.5 I 43.5 / 45.3 I 45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the applicant to identify the impact of the loss of air on a valve in the COWS system and use procedures to control and mitigate the consequences of the malfunction.

Technical

Reference:

ARI-241 -253, CCS, Rev 9 windows 249-A and 253-A Proposed references None to be provided:

Learning Objective: 21. Given a loss of instrument air/control power, determine the effect on the following valves:

a. Surge tank make up valve.

b. Surge tank vent valve.

c. Letdown Heat Exchanger temp. control valve.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

36. 008 A4.10 036 Given the following:

- Unit is in Mode 5 with RCS temperature at 140°F following completion of a refueling outage.

- CCS Heat Exchanger C is isolated and drained for tube sheet repair.

- CCS pump C-S is tagged for impeller replacement.

If RCS heatup is initiated, which ONE of the following identifies...

(1) the lowest RCS temperature at which the heat exchanger is required to be restored to operable status before the temperature is exceeded and (2) the CCS pump(s) required to be aligned to the heat exchanger to meet Tech Spec operability requirements if the CCS pump C-S remains unavailable when the Train is required?

A 200°F CCS pump lB-B, PDJY B. 200°F Either CCS pump 1 B-B or 2B-B C. 350°F CCS pump I B-B, gjy D. 350° F Either CCS pump 1 B-B or 2B-B

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Correct, In accordance with Tech Spec LCO 3.7. 7, two trains of CCS are required to be operable in Modes 1, 2, 3 and 4 Mode 4. Mode 4 is greater than 200°F (Mode 5 is equal to or less than 200°F) so 200°F cannot be exceeded prior to restoration of the second heat exchanger. If the CCS pump C-S is not restored only the CCS pump lB-B could be aligned to meet Tech Spec requirements because the CCS pump 2B-B will not start on a Unit I Safety injection and is not tested to Surveillance Requirements.

B. Incorrect, Plausible because 200°F is the correct temperature and either CCS pump lB-B or 2B-B can physically be aligned to supply CCS Train B.

C. Incorrect, Plausible because 350°F is the temperature at which MODE 31s entered and because CCS pump lB-B being the only pump that can be aligned to supply CCS Train B that meets Tech Spec requirement is correct.

D. Incorrect, Plausible because 350°F is the temperature at which MODE 3 is entered and because either CCS pump lB-B or 2B-B can physically be aligned to supply CCS Train B.

Question Number: 36 Tier: 2 Group 1 K/A: 008 A4.10 Component Cooling Water System (CCWS)

Ability to manually operate and/or monitor in the control room:

Conditions that require the operation of two CCW coolers Importance Rating: 3.1* / 3.1 10 CFR Part 55: 41.7 / 45.5 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowing the temperature at which conditions require two CCS heat exchangers during a plant heatup.

Technical

Reference:

Tech Spec 3.7.8, Essential Raw Cooling Water (ERCW System Amendment 69 Tech Spec 3.7.7, Component Cooling System (CCS)

Proposed references None to be provided:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Learning Objective: 3-OT-SYSO7OA

16. Regarding Technical Specifications and Technical Requirements for this system:

a. Identify the conditions and required actions with completion time of one hour or less.

b. Explain the Limiting Conditions for Operation, Applicability, and Bases.

c. Given a status/set of plant conditions, apply the appropriate Technical Specifications and Technical Requirements.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

37. oio&.oi 037 Given the following:

- Unit I is operating at 100% power.

- 1 -FCV-68-333, BLOCK VLV FOR PORV 340A, stroke time testing in accordance withl-Sl-68-901-A, Valve Full Stroke Exercising During Plant Operation: Reactor Coolant A-Train, is in progress.

- 1 -FCV-68-333 has been stroked closed and the procedure now directs the valve to be reopened.

Which ONE of the following identifies...

(1) the Block valve required stroke time acceptance criteria that must be met in accordance with 1-S1-68-901-A and (2) a condition that has the potential to cause the Pressurizer Relief Tank (PRT) pressure and temperature to rise when the Block valve was being open?

A. (1) Valve opening time (2) Pressure between the PORV 340A and the block valve increased rapidly.

B. (1) Valve opening time (2) PORV 340A handswitch remained in AUTO while the block valve was opening.

Cv (1) Valve closing time (2) Pressure between the PORV 340A and the block valve increased rapidly.

D. (1) Valve closing time (2) PORV 340A handswitch remained in AUTO while the block valve was opening.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect Plausible because the stroke time in the open direction is measured and recorded in the Surveillance Instruction and the condition causing the PRT pressure and temperature to increase rapidly is correct.

B. Incorrect, Plausible because the stroke time in the open direction is measured and while the PORV handswitch is left in automatic during the performance of the Sl the positioning of the valve is due to the pressure surge on the seat and not due to an automatic control function.

C. Correct The direction of PORV Block valve travel with a required acceptance time during the quarterly testing in accordance with 1-SI-901-A is travel in the closed direction and the PRT pressure and temperature could increase rapidly if the PORV opened due to the rapid increase in pressure between the PORV and the block valve causing the PORV to open and stick as identified in the SI..

D. Incorrect, Plausible because the stroke time in the closed direction is the time with acceptance criteria and while the PORV handswitch is left in automatic during the performance of the SI, the positioning of the valve is due to the pressure surge on the seat and not due to an automatic control function.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 37 Tier: 2 Group 1 K/A: 010 A3.D1 Pressurizer Pressure Control System (PZR PCS)

Ability to monitor automatic operation of the PZR PCS, including:

PRT temperature and pressure during PORV testing Importance Rating: 3.0! 3.2 10 CFR Part 55: 41.7 /45.5 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of an operation (inadvertent PORV opening) that has the potential to affect PRT pressure and temperature while the pressurizer control system is operating in automatic during the testing of the PORV block valves.

Technical

Reference:

1-SI-68-901-A, Valve Full Stroke Exercising During Plant Operation: Reactor Coolant A-Train, Rev. 0009 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO68C

11. Describe the indication an operator has that a PORV is open or leaking through.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

38. 012 K1.05 038 Given the following plant information:

- The unit was at 20% RTP when inadvertent opening of the steam dump valves resulted in an automatic Reactor Trip and Safety Injection due to pressurizer pressure dropping.

- Subsequent to the trip and SI the operator placed the steam dump valve control in manual and gained control of the valves.

- Pressurizer pressure has recovered to 1910 psig and continues to rise.

- The crew is currently performing ES-I .1, SI Termination.

Which ONE of the following identifies the actions and conditions required to re-enable automatic safety injection actuation?

A. Cycle the reactor breakers, then reset SI signal and can be done at the current pressurizer pressure.

B. Cycle the reactor breakers, then reset SI signal but can NOT be done until pressurizer pressure rises above the P-l 1 setpoint.

Cv Reset SI signal, then cycle reactor trip breakers and can be done at the current pressurizer pressure.

D. Reset SI signal, then cycle reactor trip breakers but can NOT be done until pressurizer pressure rises above the P-I I setpoint.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because the actions required are listed but are reversed in order and being able to perform the action at the current pressurizer pressure is correct.

B. Incorrect, Plausible because the actions required are listed but are reversed in order and the P-Il pressure setpoint when reached does input the SSPS system.

C. Correcl To re-enable automatic Safety Injection capability the SI must be reset and the then the Reactor Trip Breakers must be cycled and the action can be taken at the current pressurizer pressure.

D. Incorrect, Plausible because the action to reset the SI signal then cycle the reactor trip breaker is correct and the P-Il pressure setpoint when reached does input the SSPS system.

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010 Question Number: 38 Tier: 2 Group I K/A: 012 K1.05 Reactor Protection System Knowledge of the physical connections and/or cause effect relationships between the RPS and the following systems:

K1.05 ESFAS Importance Rating: 3.8*! 39 10 CFR Part 55: 41.2 to 41.9 / 45.7 to 45.8 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the knowledge of how the RPS interacts with the ESFAS as to how to re-enable automatic Safety Injection capability following an actuation of a Safety Injection.

(Physical connection and cause/effect relationship)

Technical

Reference:

ES-i.1, SI Termination, Revision 0016 i-47W61 1-63-1 R13; Proposed references None to be provided:

Learning Objective: 3-OT-SYSO99A

6. Briefly describe the inputs to the SSPS.

3-OT-EOPO1 00

8. Given a set of plant conditions, use E-1, ES-1.1, ES-I .2, ES-i .3, and ES-i .4 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes, and Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO3A.25 007 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

39. 013 G2.4.6 039 Given the following:

- Unit 1 was operating at 100% when a small steam leak inside containment occurred.

- Due to containment pressure rising, the reactor was tripped.

- After the reactor was manually tripped, no AFW flow could be established resulting in the implementation of FR-H.1, Loss of Secondary Heat Sink.

- After entering FR-H.1, a Safety Injection occurred due to containment pressure which is now 1.7 psig and stable.

- The Crew has placed all four SI Block Switches on 1-M-4 to BLOCK and the Safety Injection signal has been reset using the SI RESET pushbuttons on 1-M-6.

- The crew is currently attempting to restore main feedwater.

Which ONE of the following identifies the mitigation strategy required to allow the Feedwater Isolation (FWI) signal to be reset?

Av Auto SI must be blocked using IMI-99.040, Auto SI Block, then the reactor trip breakers must be cycled before the FWI signal can be reset.

B. The FWI signal can be reset without blocking Auto SI using IMI-99.040, Auto SI Block, or cycling the reactor trip breakers.

C. Auto SI must be blocked using IMI-99.040, Auto SI Block, but the reactor trip breakers do NOT have to be cycled before the FWI signal can be reset.

D. The FWI signal can be reset without blocking Auto SI using IMI-99.040, Auto SI Block, but the reactor trip breakers must be cycled before the FWI signal can be reset.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. Correct, Because the containment pressure is greater than the SI setpoint (1.5 psig), instrument maintenance must manually block the AUTO SI in the SSPS cabinets to prevent the SI signal from coming back in as soon at the reactor trip breakers are cycled and the reactor trip breakers must be cycled to clear the FWI signal seal in circuit.

B. Incorrect, Plausible because the FWI signal could be cleared if the containment pressure had been below the setpoint after the 1-M-4 SI block switches are placed to block and the 1-M-6 SI RESET push buttons are pushed the other 2 automatic SI signals are blocked and there are conditions where a FWI signal can be cleared without cycling the reactor trip breakers. (see FR-H. 1 Note below)

C. Incorrect Plausible because with the containment pressure greater than the SI setpoint (1.5 psig), instrument maintenance must manually block the AUTO SI in the SSPS cabinets to prevent a standing SI signal from being present awaiting the reactor breakers to be cycled and there are conditions where a FWI signal can be cleared without cycling the reactor trip breakers. (see FR-H. I Note below)

D. Incorrect, Plausible because the FWI signal could be cleared if the containment pressure had been below the setpoint after the 1-M-4 SI block switches are placed to block and the 1-M-6 SI RESET push buttons are pushed the other 2 automatic SI signals are blocked and the cycling of the reactor trip breakers are required is correct FR-H.1, Loss of Secondary Heat Sink, Rev 17 page 10 NOTE . Cycling reactor trip breakers to allow MEW Isolation reset is only required if SI or HI-HI S/G level has occurred.

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010 Question Number: 39 Tier: 2 Group 1 K/A: 013 G2.4.6 013 Engineered Safety Features Actuation System (ESFAS)

Knowledge of EOP mitigation strategies.

Importance Rating: 3.7 I 4.7 IOCFRPart55: 41.10/43.5/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requries knowledge of EOP mitigation strategies associated with resetting the ESFAS to allow recovery actions to be initiated.

Technical

Reference:

AOl-38, Main Steam or Feedwater Line Leak, Revision 8 FR-H.1, Loss of Secondary Heat Sink, Rev 17 1 -47W6 11-3-2 R22 1-47W611-63-1 R13 Proposed references None to be provided:

Learning Objective: 3-OT-FRH0001

17. Given a set of plant conditions, use FR-H.1, H.2, H.3, H.4, & H.5 and the Critical Safety Function Status Trees to correctly diagnose and implement:

Action Steps, RNOs, Foldout Pages, Notes and Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN bank question SYSOO3A.13 008 modified Comments:

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010

40. 022 G2.2.36 040 Given the following:

- Unit 1 is in Mode 3.

Which ONE of the following identifies Containment Cooling equipment that when removed from service and tagged for maintenance, places the Unit in a condition that requires the coolers be returned to service within a specified time unless the unit is placed in Mode 5 in accordance with the Technical Requirements Manual?

A a Lower Compartment Cooler, gpjy B. a Lower Compartment Cooler or a Control Rod Drive Motor Cooler C. a Lower Compartment Cooler or a Reactor Coolant Pump Motor Cooler D. a Lower Compartment Cooler or an Upper Compartment Cooler DIS TRACTOR ANAL YSIS:

A. Correct, The Lower Compartment Coolers are required by TR 3.6.3 if the Unit is in Modes 1-4 and if not returned within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed completion time the plant is required to be placed in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

B. Incorrect, Plausible because the Control Rod Drive Motor Coolers are coolers in lower containment and are required to be placed in during the heatup of the unit C. Incorrect, Plausible because the Lower Compartment Coolers being out of service for maintenance is correct and the Reactor Coolant Pump Motor Coolers are coolers in lower containment required to be placed in during the heatup of the unit.

D. Incorrect, Plausible because the Lower Compartment Coolers being out of service for maintenance is correct and the Upper Compartment Coolers are coolers in upper containment required to be placed in during the heatup of the unit as containment heats up.

Question Number: 40 Tier: 2 Group 1 KIA: 022 G2.2.36 Containment Cooling System (CCS)

Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Importance Rating: 3.1 I 4.2 10 CFR Part 55: 41.10 / 43.2 /45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the ability to analyze how maintenance activities on containment cooling equipment affect the status of limiting conditions for operations for a given operating Mode.

Technical

Reference:

TR 3.6.3, Lower Compartment Cooling (LCC) System, 09/30/95 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO300

3. Regarding Technical Specifications and Technical Requirements for this system:

a. Identify the conditions and required actions with completion time of one hour or less.

b. Explain the Limiting Conditions for Operation, Applicability, and Bases.

c. Given a status/set of plant conditions, apply the appropriate Technical Specifications and Technical Requirements.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question T1R363 001 modified Comments:

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010

41. 025 K1.01 041 Given the following:

- Unit 1 is operating at 100% power.

- Operators are in the process of placing Train A Containment Purge in service to lower containment.

Which ONE of the following damper(s) is opened st to ensure that the lower ice doors remain closed during startup of the purge in accordance with SOl-30.2, Containment Purge System?

A 1-FCV-30-2, Containment Purge Air Supply Fan 1A Discharge B. 1-FCO-30-IA, Containment Purge Air Supply Fan 1A Suction C. 1-FCV-30-213, Containment Purge Air Exhaust Fan 1A Discharge D. 1-FCV-30-61, Containment Purge Air Exhaust Fan IA Suction DISTRACTOR ANAL YSIS:

A. Correct, Opening this damper last is in accordance with the procedure. Need to minimize dip across the doors and keep lower containment pressure negative in respect to upper containment.

B. Incorrect Plausible because the damper is opened during placing the lower containment purge in service. This damper opens when the fan is started and would provide the same effect as the discharge damper, but would not be in accordance with SOI-30. 02.

C. Incorrect, Plausible because the damper is opened during placing the lower containment purge in service and the sequence could be mistaken.

D. lncorrect, Plausible because the damper is opened during placing the lower containment purge in service and the sequence could be mistaken.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 41 Tier: 2 Group 1 K/A: 025 K1.01 Ice Condenser System Knowledge of the physical connections and/or cause/effect relationships between the ice condenser system and the following systems:

Containment ventilation Importance Rating: 2.7* / 2.7*

10 CFR Part 55: 41.2 to 41.9/ 45.7 to 45.8 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the cause/effect relationship between the ice condenser and the containment ventilation system due to the physical connections between them.

Technical

Reference:

SOI-30.02, Containment Purge System, Rev. 0054 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO6 1 A

6. Describe the ice condenser doors and state at what pressures they open.

3-OT-SYSO3OC

12. Given certain plant conditions determine if SQl precautions and limitations for that system apply.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

42. 026 A4.05 042 Given the following conditions:

- Unit 1 was operating at 100% power when a LOCA occurred.

- While performing E-0, Reactor Trip or Safety Injection, the operating crew manually actuated Phase B Containment Isolation after the automatic signal failed.

- During performance of E-1, Loss of Reactor or Secondary Coolant, the operating crew is ready to stop the Containment Spray Pumps and place them in A-AUTO.

Which ONE of the following identifies...

(1) the signal(s) required to be reset to allow the pumps to be remain off when the Containment Spray Pump handswitches are placed back in A-AUTO after resetting.

and (2) the action required to reset the Containment Spray signal?

A. (1) Containment Spray Signal, only (2) Both Train A & B Reset pushbuttons are required to be pushed simultaneously.

B. (1) Phase B Signal and Containment Spray Signal (2) Both Train A & B Reset pushbuttons are required to be pushed simultaneously.

C (1) Containment Spray Signal, only (2) The Train A and B Reset pushbuttons can be pushed independently.

D. (1) Phase B Signal and Containment Spray Signal (2) The Train A and B Reset pushbuttons can be pushed independently.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. lncorrect, Plausthie because only the Containment Spray signal is required to be reset and for Phase B actuation, both the Train A and B pushbuttons must be actuated simultaneously.

B. Incorrect, Plausible because the Containment Spray signal is required to be reset and the Phase B signal when manually initiated will automatically start the containment spray pumps. Also, because when Phase B is manually actuateci both the Train A and B pushbuttons must be actuated simultaneously.

C. Correcl The signal can be reset and the pumps remain off when restored to A-AUTO after pushing the Containment spray reset pushbuttons and the Train A and B pushbuttons can be reset independently.

D. lncorrect Plausible because the Containment Spray signal is required to be reset and the Phase B signal when manually initiated will automatically start the containment spray pumps. Also, because when resetting the containment spray signaI the Train A and B pushbuttons can be reset independently.

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 Question Number: 42 Tier: 2 Group 1 KIA: 026 A4.05 Containment Spray System (CSS)

Ability to manually operate and/or monitor in the control room:

Containment spray reset switches Importance Rating: 3.5 / 3.5 10 CFR Part 55: 41.7 / 45.5 to 45.8 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the applicant to identify the manual actions required to reset the containment Spray actuation signal and restore the pumps to a standby conditions with their respect switches in A-AUTO.

Technical

Reference:

E-1, Loss of Reactor or Secondary Coolant, Rev 15 1-47W61 1-72-1 R7 1-47W611-88-1 R23 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO72A

16. Given a set of plant conditions, determine the correct response of the Containment Spray System.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN question SYSO72A.08 016 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

43. 039 K3.04 043 Given the following:

- Unit 1 in service at 100% power.

Which ONE of the following competes the statement below?

A leak in the high pressure stage (second stage) of MSR (1) will cause the temperature in the low pressure steam being supplied to the MFPT5 to (2)

A Al drop B. Al rise C. Cl drop D. Cl rise DISTRA CTOR ANAL YSIS:

A. Correct, a leak in MSR Al High Pressure stage will cause the temperature of the steam being supplied to the MFPTs to drop because the enthalpy of the superheated steam will drop as the higher pressure lower enthalpy steam leaks in to the shell side of the MSR.

B. Incorrect, Plausible because the leak in MSR A I causing the temperature change is correct and because the main steam temperature is higher, it could be concluded to cause the temperature of the steam to the MFP turbines to rise (when it actually has lower enthalpy and will cause the temperature to drop)

C. Incorrect, Plausible because MSR Cl/s physically the same heat exchanger as MSR Al but is not one of the MSRs that supply steam to the MFPTs but the reheated steam temperature dropping due to a tube leak in the high pressure stage is correct D. Incorrect, Plausible because MSR Cl is physically the same heat exchanger as MSR Al but is not one of the MSRs that supply steam to the MFPTs and because the main steam temperature is higher, it could be concluded to cause the temperature of the steam to the MFP turbines to rise (when it actually has lower enthalpy and will cause the temperature to drop)

Question Number: 43

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 Tier: 2 Group 1 KIA: 039 K3.04 Main and Reheat Steam System Knowledge of the effect that a loss or malfunction of the MRSS will have on the following:

MEW pumps Importance Rating: 2.5* I 2.6*

IOCFRPart55: 41.7/45.6 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the questions requires knowledge of the steam supplies to the MFPTs and how a leak in an tube bundle will affect to MSR outlet temperature.

Technical

Reference:

1-47W801-1 R41 47K1110-1A RO 47K1110-2A RO SQl-i .04, Moisture Separator Reheaters, Rev .0027 Proposed references None to be provided:

Learning Objective: 3-OT-SYS001C

8. Describe the precautions associated with the operation of the Moisture Separator Reheaters.

3-OT-SYSO46A

4. Identify the steam supplies to the main Feedwater Pump turbine and briefly explain how each is used to control Feed Pump output.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank questions SYSOOIC.01001 and SYSOO1C.10 001 combined and modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

44. 059 1(3.02 044 Given the following:

- Unit 1 is operating at 30% power when an instrument air line failure resulted in the #3 SG Main Feedwater reg valve closing.

- Operators tripped the reactor in anticipation of an automatic trip.

- Level in #3 SG dropped to 6% NR, and is recovering.

- Levels in 1, 2, and 4 SG5 dropped to 28% NR, and are recovering.

- RCS is stabilized at normal RCS no load temperature and pressure.

Which ONE of the following indicates the status of the Auxiliary Feedwater (AFW) pumps?

A. Only the lB-B AFW pump running.

B. Only the lA-A and lB-B AFW pumps running.

C. Only the TDAFW and 1 B-B AFW pumps running.

Dv All three AFW pumps running.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausthie because the lB-B AFW pump is the MD pump that supplies the

1. 3 SG and with only one SG NR level dropping to the start signal generated is only to motor driven pumps.

B. Incorrect, Plausible because the MD AFW pumps are the pumps that receive a start signal from a single SG NR level dropping below the AFW pump start setpoint.

C. Incorrect,, Plausible because the TDAFW pump and the lB-B MD pump are the two AFW pumps that can supply the #3 SG.

D. Correct, All three AFW pumps will be running. With the Reactor tripped and Tavg at normal post trip temperature a MFW isolation had occurred which tripped both MFPT generating a start signal to all AFW pumps.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 44 Tier: 2 Group I K/A: 059 K3.02 Main Feedwater (MEW) System KK3 Knowledge of the effect that a loss or malfunction of the MEW will have on the following:

AFW system Importance Rating: 3.6 I 3.7 IOCFRPart55: 41.7/45.6 IOCFR55.43.b: Not applicable K/A Match: KA matched because the question requires the applicant to determine the effect a failure of a main feedwater regulating valve will have on the AFW system Technical

Reference:

1-47W611-3-3 RiO 1-47W611-3-4 R18 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO3B

10. Identify the A-Auto start signals of the Motor-Driven AEW pumps.

17. Identify the Turbine-Driven Auxiliary Feedwater pump Auto start signals.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN Bank question 059 K3.02 with a distractor changed, stem re-formatted ( but not really modified) and correct answer relocated.

Comments:

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

45. 061 A1.04 045 Unit 1 heatup is in progress in accordance with GO-I, Unit Startup from Cold Shutdown to Hot Standby, following a Mode 5 outage with the following conditions:

- RCS is currently at 320°F and 530 psig.

- Auxiliary Feedwater (AFW) pumps IA-A and I B-B are in service.

- A large leak develops near the bottom of the Condensate Storage Tank (CST).

Which ONE of the following identifies the minimum CST level required by Technical Specification LCO 3.7.6, Condensate Storage Tank, and if the level continues to drop how AFW pumps lA-A and 1 B-B suction supply will be affected?

Minimum Level AFW Pumps A 200,000 gallons Suction will automatically swap to ERCW due to low pressure.

B. 200,000 gallons Suction will be manually transferred to ERCW when annunciator 63-A CST HDR TO AFW PMPS PRESS LO alarms.

C. 116,000 gallons Suction will automatically swap to ERCW due to low pressure.

D. 116,000 gallons Suction will be manually transferred to ERCW when annunciator 63-A CST HDR TO AFW PMPS PRESS LO alarms.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Correct, The Tech Spec LCO requires the level to be equal to or greater than 200,000 gallons and if the leak resulted in the CST level continuing to drop the AFW pump IA-A and lB-B suction would automatically open the motor operated valves allowing ERCW to supply suction to the AFW pumps when the suction pressure to the respective pump dropped below setpoint for greater than 10 seconds.

B. Incorrect, Plausible because 200,000 gallons being the minimum required level is correct and annunciator 63-A CST HDR TO AFW PMPS PRESS LO will alarm when the pressure is below the required setpoint on the AFW pumps but the AFW pump suction transfer to ERCWis automatic.

C. Incorrect, Plausible because 116,000 gallons is the setpoint for annunciator 41-A CSTA LEVEL LO-LO alarm and the AFW pump suction transfer to ERCW being automatic when the pressure is below the required setpoint is correct.

D. Incorrect Plausible because 116,000 gallons is the setpoint for annunciator 41-A CSTA LEVEL LO-LO alarm and annunciator 63-A CST HDR TO AFW PMPS PRESS LO will alarm when the pressure is below setpoint but the transfer to ERCWis automatic.

Question Number: 45 Tier: 2 Group 1 K/A: 061 A1.04 Auxiliary / Emergency Feedwater (AFW) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including:

AFW source tank level Importance Rating: 3.9 / 3.9 10 CFR Part 55: 41.5/45.5 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the design limits of the CST which is the AFW source tankand the ability to predict the affect a lowering of the tank level will have on the AFW Pu m PS.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Technical

Reference:

Tech Spec 3.7.6, Condensate Storage Tank (CST)

ARI-36-42, Heaters, Turb Seal & Air, Revision 17 windows 41-A and 41-B ARI-57-63, Feedwater & Main Steam, Revision 17 window 63-A 1-47W611-3-3 RiO Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO3B

6. Identify the required CST volume needed for AFW operation as stated in Tech Specs and the basis for this volume.

8. Describe the automatic opening signal for the ERCW supply valves to the AFW system.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank questions SYSOO3B.05 006 and SYSOO3B.08 001 combined and modified.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

46. 062 A3.05 046 Given the follow conditions:

- Unit 1 is operating at 100% power when a loss of offsite power occurs.

- The DGs restore voltage to the Shutdown Boards and the Blackout Relays are sequencing loads back to the boards.

- A Safety Injection occurs during the blackout sequencing.

Which ONE of the following identifies the events that occur due to the Safety Injection actuation?

A. Loads already sequenced on will remain on. Timers for equipment not already started will continue to run and the timers for the additional loads to be started due to the Safety injection will start to sequence the loads on.

B Loads already sequenced on will remain on. The timers for equipment not already started will be reset to zero and the required loads will then be sequenced on.

C. Emergency Feeder breaker connecting the DG to the board will open causing the blackout sequence to restart. The Emergency Feeder breaker will reclose allowing the required loads to sequence on.

D. DG will remain connected to the board but the loads already sequenced on will trip. Timers for the required loads will be reset to zero and the required loads will then be sequenced on.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

When a Blackout has occurred and load sequencing is in progress, if an SI signal is received the Blackout sequence stops. The timers for equipment not yet started is reset to zero. Equipment already running remains in service and all other required equipment starts to sequence on.

A. Incorrect, Plausible because the loads already started do remain in service and additional equipment timers will be started to sequence equipment on.

B. Correct, Loads already started will remain in service. Timers for equipment not yet started will reset to zero and the required loads will be sequenced on by the timers.

C. Incorrect, Plausible because the breaker would open if the diesel generator had been feeding the board in parallel with another supply when the safety injection occurred. This would caused the board to blackout and start the sequence for loading the board D. Incorrecl Plausible because the DG does remain connected to the board and the load already connected would trip under conditions (see above) causing all timers to be reset prior to sequencing the equipment on.

Question Number: 46 Tier: 2 Group 1 K/A: 062 A3.05 AC Electrical Distribution System Ability to monitor automatic operation of the ac distribution system, including:

Safety-related indicators and controls Importance Rating: 3.5 / 3.6 IOCFRPart55: 41.7/45.5 IOCFR55.43.b: Not applicable K/A Match: KA is matched becuase the applicant is must possess the ability to monitor the automatic loading of safety-related equipment during all possible load sequence scenarios and determine that the sequencer has operated properly using Main Control Board indications for the various sequencer loads.

Technical

Reference:

1 -45W760-62-1 Ri 5 3-OT-STG-201 B, Operations Blackout and Load shed Logic Student Training Guide, Rev 1

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Proposed references None to be provided:

Learning Objective: 3-OT-SYS2O 1 B

16. Identify the equipment and associated (sequence timer) ST setpoint that will start with the following:

a. A blackout with a return of voltage

b. A blackout with a return of voltage together with a safety injection signal.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN question SYS2O1B.16 modified Correct answer relocated Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

47. 062 K2.01 047 Which ONE of the following identifies the normal and alternate power supplies to Reactor Coolant Pump # 3?

Normal Supply Alternate Supply A. USST IA RCP Start Bus A B. USST IA RCP Start Bus B C USST 1 B RCP Start Bus A D. USST iB RCP Start Bus B DISTRA CTOR ANAL YSIS:

A. Incorrect, the normal supply is not from USST IA but the alternate being from RCP Start Bus A is correct. Plausible because USST IA is the normal supply for two of the RCPs (#1 & #2) and the alternate supply is correct.

B. Incorrect the normal supply is not from USST IA and the alternate is not from RCP Start Bus B. Plausible because USST IA is the normal supply for two of the RCPs

(#1 & #2) and the RCP Start bus B is also the alternate supply for two of the RCPs

(#2 & #4)

C. Correct the normal supply is from the USST lB and the alternate is from RCP Start Bus A as identified on print l-45W720.

D. Incorrect, the normal supply is correct but the alternate is not from RCP Start Bus B. Plausible because USST lB being the normal is correct and the RCP Start Bus B is the alternate supply for two of the RCPs (#2 & #4).

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 47 Tier: 2 Group 1 K/A: 062 K2.01 AC Electrical Distribution System Knowledge of bus power supplies to the following:

Major system loads Importance Rating: 3.3 I 3.4 IOCFRPart55: 41.7 10CFR5543.b: Not applicable K/A Match: Applicant must identify the AC electrical distribution system normal and alternate power supplies to a Reactor Coolant Pump (major system load)

Technical

Reference:

1-45W720 R8 Proposed references None to be provided:

Learning Objective: 3-OT-68-5Y5068B

12. Identify the RCPs Normal and Alternate Power Supplies Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO68B.12 002 modified Comments:

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010

48. 063 A1.01 048 Which ONE of the following identifies batteries that will have loads shed during performance of AOl-40, Station Blackout and the benefit of shedding the load?

A. 125V DC Diesel Generator Batteries to provide battery capacity for DC control and flashing the generator field when conditions allow the restart of the diesel generator.

B. 125V DC Diesel Generator Batteries to provide battery capacity for prolonged continuous operation operation of the DC driven oil pumps on the diesel engines and turbocharger.

C 250V DC Station Batteries to provide battery capacity to support breaker operation for restoration of AC power to the Shutdown boards from the 500kV system.

D. 250V DC Station Batteries to allow adequate battery capacity to ensure emergency lighting in the Turbine Building while the AC power is not available.

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the 125V Vital Batteries will undergo load shedding when A 01-40, Station Blackout, is implemented (but not the 125V Diesel Generator Batteries) and the battery is a safety related battery needed for control and field flashing when the diesel generator is to be placed in service.

B. lncorrect, Plausible because the 125V Vital Batteries will undergo load shedding when A 01-40, Station Blackout, is implemented (but not the 125V Diesel Generator Batteries) and a DG battery is a safety related component that supply power to the DC oil pump to help in maintaining the DG ready to start.

C. Correct, The 250V DC Station Batteries will have load shedding initiated when A 01-40, Station Blackout, is implemented and the reason is to preserve the batteries capacity to provide to support breaker operation for restoration of AC power to the Shutdown boards from the 500kV system.

(see excerpt below)

D. lncorrect Plausible because 250V DC Station Batteries having load shedding initiated when A 01-40, Station Blackout is implemented is correct and providing turbine building lighting is important, but not correct during A 01-40 implementation.

AOI-40, Station Blackout 4.0 Discussion Analysis of the vital battery capacity, for the duration of the postulated 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Station Black Out (SBO) event, shows that the batteries are capable of powering all loads necessary to obtain and maintain the unit in the Safe Shutdown condition throughout the entire event, provided that the nonessential loads are shed from the emergency DC lighting cabinets to extend the useful life of the batteries. Nonessential loads are shed from the 250V Station Batteries as soon as possible after the event in order to ensure sufficient capacity to support breaker operation for restoration of AC power to the Shutdown Boards from the 500Kv system.

The nonessential loads are listed on Appendices A and B to this instruction. The...

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 63 Tier: 2 Group 2 K/A: 063 A1.01 DC Electrical Distribution System Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including:

Battery capacity as it is affected by discharge rate Importance Rating: 2.5 / 3.3 IOCFRPart55: 41.5/45.5 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the effect load (discharge rate) has on battery capacity and the ability to predict how and why action are taken to reduce the load on the battery.

Technical

Reference:

AOl-40, Station Blackout, Revision 13 Proposed references None to be provided:

Learning Objective: 3-OT-A014000

5. State what loads are removed from the dc systems to preserve battery capacity.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

49. 063 A2.01 049 Given the following conditions:

- Unit I is in Mode 3 controlling Tave with the SG PORV Pressure Indicating Controllers (PICs) on l-M-4 in manual.

- A ground develops on the GREEN indicating light socket in the control circuit for SG 3 PORV PCV-1-23.

Which ONE of the following identifies.

(1) the Battery Board ground indicator where the ground would be displayed and (2) if the ground later caused the control fuse to blow, how would the manual operation of the PORV be continued?

A& (1) 125v DC Vital Battery Board I (2) Control would be from the PlC B. (1) 125v DC Vital Battery Board I (2) Local control would be required C. (1) 250v DC Station Battery Board 1 (2) Control would be from the PlC D. (1) 250v DC Station Battery Board 1 (2) Local control would be required

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Correct, a ground of the control power would be indicated on the ground indicator on the 125v DC Vital Battery Board I and if the control power fuse blew the operator will still have control of the PORV from the PlC on the control board but the indicating lights would not be available.

B. Incorrect, Plausible because the location of the ground detector is correct and there are other conditions that would result in the PORVs having to be controlled from one of the 3 methods of local control identified in SQl-I. 01.

C. Incorrect, Plausible because there are many control circuits on the secondary side of the plant that use 250v DC for control and the operator still being able to control the valve form the PlC is correct.

D. Incorrect, Plausible because there are many control circuits on the secondary side of the plant that use 250v DC for control and there are other conditions that would result in the PQRVs having to be controlled from one of the 3 methods of local control identified in SOl-1.01.

Question Number: 49 Tier: 2 Group 1 KIA: 063 A2.01 DC Electrical Distribution System Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Grounds Importance Rating: 2.5 I 3.2*

10 CFR Part 55: 41.5 I 43.5 / 45.3 I 45.13 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of how a ground would be displayed and if the ground caused a loss of control power to a valve being operated, how the valve operation would be continued as identified in plant instructions.

Technical

Reference:

AOl-21.01, Loss of 125V DC Vital Battery BD 1, Rev 21 SQl-i .01, Main Steam System, Rev 0040 1-45W600-1-4 R8 1-47W611-i-1 R13

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO1A

22. Explain the operation of the atmospheric relief valve in auto, manual, and with loss of air pressure.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Question modified from Vogtle 2009 exam question 063A2.O1 1 Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

50. 064K1.03 050 Which ONE of the following describes the Fuel Oil system on one of the four diesel generator sets?

A Each of the DG engines has a priming pump that starts when the DG receives an emergency start signal.

B. Each of the DG engines has a priming pump that is running when the DG is in Standby Alignment.

C. There is one priming pump to supply both engines that starts when the DG receives an emergency start signal.

D. There is One priming pump to supply both engines that is running when the DG is in Standby Alignment.

DISTRACTOR ANAL YSIS:

A. Correct, There is a DC pump mounted on the Day Tank of each engine that is started when the DG receives an emergency start signal.

B. Incorrect, Plausible because each engine having its own pump is correct and there are other oil pumps (lube oil pumps) that do run while the DG is in Standby Alignment. Running the fuel oil pumps while in standby alignment would maintain the prime on the engines, but there are check valves in the suction line to the pumps to maintain the prime.

C. lncorrect Plausible because both engines are supplied by one pump running in the case of makeup to the day tanks from the 7 day tanks and the pump starting when an emergency start signal is generated is correct.

D. lncorrect, Plausible because both engines are supplied by one pump running in the case of makeup to the day tanks from the 7 day tanks and there are other oil pumps (lube oil pumps) that do run while the DG is in Standby Alignment. Running the fuel oil pumps while in standby alignment would maintain the prime on the engines, but there are check valves in the suction line to the pumps to maintain the prime.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 50 Tier: 2 Group 1 K/A: 064 K1.03 Emergency Diesel Generators (ED/G)

Knowledge of the physical connections and/or cause effect relationships between the ED/G system and the following systems:

Diesel fuel oil supply system Importance Rating: 3.6 I 4.0 10 CFR Part 55: 41.2 to 41.9 /45.7 to 45.8 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the physical connections of the pumps that supply fuel oil from the day tank to the DG engines and how the operation of the pumps is affected by the status of the DG (cause effect relationship)

Technical

Reference:

N3-82-4002, Standby Diesel Generator System, Rev.001 5 1-47W760-82-2 R17 1 -47W760-82-4 R20 Proposed references None to be provided:

Learning Objective: 3-OT-SYS82I

12. Identify eight systems to be checked or aligned for the diesel standby condition for SOl-82.01.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO82E.02 002 modified to match the KA.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 51.o64K4.o3o51 Given the following:

- 6.9kV Shutdown Board IA-A is being supplied by DG lA-A following a blackout signal.

- The operating crew is in the process of removing the diesel generator from service in accordance with SOl-82.01, Diesel Generator (DG) IA-A.

- The crew is ready to parallel the shutdown boards Normal feed to the diesel generator and 1-HS-57-42, NORMAL CSST C SYNC SWITCH, is in the SYN position.

- The synchroscope is rotating fast in the FAST direction.

Which ONE of the following identifies...

(1) the direction the diesel generator speed control switch will initially have to be manipulated to establish conditions for closing Shutdown Board IA-A NOR supply breaker, and (2) the mode of speed control after the normal supply breaker (1716) is closed in parallel with the diesel?

A. (1)raise (2) speed droop B. (1)raise (2) without speed droop C (1)lower (2) speed droop D. (1)lower (2) without speed droop

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. lncorrect the DG will need its speed lowered (not raised) by going to lower on the speed control switch to achieve slow in the FAST direction but after the breaker is closed the mode of speed control would be speed droop. Plausible because the system would appear to be the incoming and the mode of speed control is correct.

B. Incorrect, the DG will need its speed lowered (not raised) by going to lower on the speed control switch to achieve slow in the FAST direction and after the breaker is closed the mode of speed control would not be without speed droop (isochronous.)

It would be speed droop. Plausible because the system would appear to be the incoming and the mode of speed control would be without speed droop (isochronous) prior to closing the breaker.

C. Correct, Even though the grid would appear to be the incoming and the DG the running componen1, the DG is always incoming and therefore the DG will need its speed lowered by going to lower on the speed control switch to achieve slow in the FAST direction and after the breaker is closed the mode of speed control would be speed droop.

D. lncorrect, the DG will need its speed lowered by going to lower on the speed control switch to achieve slow in the FAST direction but after the breaker is closed the mode of speed control would not be without speed droop (isochronous.) It would be in speed droop. Plausible because the system would appear to be the incoming and the mode of speed control would be without speed droop (isochronous) prior to closing the breaker.

Question Number: 51 Tier: 2 Group 1 K)A: 064 K4.03 Emergency Diesel Generator (ED/G) System Knowledge of ED/G system design feature(s) and/or interlock(s) which provide the following:

Governor valve operation Importance Rating: 2,5 / 3.0 10 CFR Part 55: 41.7 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of EDIG system design features and interlocks which provide for the mode of operation of the EDIG governor valve operation.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Technical

Reference:

SOl-82.01, Diesel Generator(DG) lA-A, Rev. 0070 Proposed references None to be provided:

Learning Objective: 3-OT-SYS82I

6. Describe the following Diesel Generator Control Relays in terms of locations, energizing initiations, functions, setpoints, interlocks, timers, resets, power supplies, and effects on other relays and/or equipment as applicable.

m. Hydraulic Governor Speed Control

n. Electric Governor Controls and Relays-SRX & SLX Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN bank question 064 K4.03 050 used on 1/2009 retake exam.

Comments: Stem wording changed to match WBN and correct answer relocated.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

52. 073 K4.01 052 Given the following plant conditions:

- The RadwasteAUO is in the process of making a liquid radwaste release from the Monitor tank.

- 0-RM-90-122, WDS Liquid Effluent, has just been placed in service.

- The release flowrate has been adjusted with 0-ISV-77-660, Cooling Tower Blowdown Release Header Isolation.

- Effluent radiation levels rose sharply during the release, causing a high radiation alarm on 0-RM-90-122.

Which of the following would occur?

A Automatic isolation of 0-RCV-77-43, Cooling Tower Blowdown Radiation Release Control, to terminate the release.

B. Automatic isolation of the cooling tower blowdown diffuser valves, 0-FCV-27-1 00 and 101, which routes the effluent to the holding pond.

C. 0-RM-90-122 alarms on the local radwaste panel, 0-L-2, and the release will continue until the AUO manually isolates 0-ISV-77-660, Cooling Tower Blowdown Release Header Isol.

D. 0-RM-90-1 22 alarms in the Main Control Room, and requires the operator to manually isolate the cooling tower blowdown diffuser valves, 0-FCV-27-1 00 and 101, which routes the radwaste release to the holding pond.

DIS TRACTOR ANAL YSIS:

A. Correct High Rad sensed on the release line radiation monitor causes the automatic closure of O-RCV-77-43 to terminate the release.

B. Incorrect, Plausible because these valves are in the liquid release flowpath and the valves to receive an automatic close signal but the signal is generated from low river flow, not high radiation. 0-IS V-77-660 is a valve in the Cooling Tower blowdown flow path.

C. Incorrect, Plausible because there are liquid release points with radiation monitors that will cause an alarm but not an automatic termination. Thus requiring a manual action to isolate the release.

D. Incorrect, Plausible because there are liquid release points with radiation monitors that will cause an alarm but not an automatic termination. Thus requiring a manual action to isolate the release. 0-FCV-100 and -101 are valves in the Cooling Tower blowdown flow path.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 52 Tier: 2 Group 1 K/A: 073K4.01 Process Radiation Monitoring (PRM) System K4 Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following:

Release termination when radiation exceeds setpoint Importance Rating: 4.0 / 4.3 10 CFR Part 55: 41.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the design feature of a release path which provides for the termination of the release when radiation exceeds setpoint.

Technical

Reference:

SOI-77.01, Liquid Waste Disposal, Revision 0065 1-47W61 1-77-2 R5 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO77A

10. Describe the Monitor Tank; include location, typical drains it receives, number of pumps, and its process paths.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question SYSO9OA.27 001 with UNIDs corrected for valves Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

53. 076 A2.01 053 Given the following:

- Unit 1 is in Mode 1 when the 2B-B ERCW Strainer clogs.

- AOl-I 3, Loss of Essential Raw Cooling Water, is implemented.

Which ONE of the following identifies the CCS Heat Exchanger whose CCS outlet temperature would increase and a requirement that must be implemented if the Train-B Supply Headers were cross-tied in accordance with AOI-13 due to the strainer being clogged?

A. CCS Heat Exchanger B; Train-B Flow Balance valves must be repositioned.

B. CCS Heat Exchanger B; ERCW Strainer lB-B must be maintained in continous backwash.

C. CCS Heat Exchanger C; Train-B Flow Balance valves must be repositioned.

D CCS Heat Exchanger C; ERCW Strainer lB-B must be maintained in continous backwash.

DISTRA CTOR ANAL YSIS:

A. lncorrecI, Plausible because the ERCW loss is a Train B loss (however CCS heat exchanger B is a Train-A heat exchanger) and the repositioning of the flow balance valves is addressed in the procedure section for the cross-tie of the headers but it identifies that repositioning is not required.

B. Incorrect Plausible because the ERCW loss is a Train B loss (however CCS heat exchanger B is a Train-A heat exchanger) and maintaining in-service strainer (lB-B) in continuous backwash is correct.

C. lncorrect Plausible because CCS heat exchanger C losing cooling resulting in an increase in CCS temperature leaving the heat exchanger is correct and the repositioning of the flow balance valves is addressed in the procedure section for the cross-tie of the headers but it identifies that repositioning is not required.

D. Correct, CCS heat exchanger C would lose cooling resulting in an increase in CCS temperature leaving the heat exchanger and if the strainer cross-ties are used the in-service strainer (lB-B) is required to be maintained in continuous backwash This is a system modification made in the last Refueling outage.

Question Number: 53

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Tier: 2 Group 1 K/A: 076 A2.01 Service Water System (SWS)

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of SWS Importance Rating: 35* / 37*

10 CFR Part 55: 41 .5/43.5/45/3 /45/13 IOCFR55.43.b: Not applicable K/A Match: Questions require the applicant to predict the impact of a condition resulting in the loss of ERCW header flow due to a clogged strainer and actions allowed in procedures to mitigate the consequences of the malfunction.

Technical

Reference:

AOl13, Loss of Essential Raw Cooling Water, Rev 38 SOl-67.01, Essential Raw Cooling Water System, Revision 0107 1 -47W845-1 R56 1 -47W845-2 R76 Proposed references None to be provided:

Learning Objective: 3-OT-A0l1300

8. Demonstrate ability/knowledge of AOl, to correctly:

b. Respond to Action steps.

3-OT-SYSO67A

3. Describe the ERCW System flow path from the river to the cooling tower basin and discharge holding pond including:

a. Interfaces

b. Major components

c. Paths to and from the Auxiliary Building Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question History: New question Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

54. 078 K2.02 054 Which ONE of the following identifies the power supply to the B Aux Air Compressor?

A 480V C & A Vent Board 2B1-B B. 480V C & A Vent Board IBI-B C. 480V Reactor MOV Board 2B1-B D. 480V Reactor MOV Board 1B1-B DISTRA CTOR ANAL YSIS:

The Auxiliary Air Compressors are the compressors used to provide air to safety related components following a loss of the normal control air system. There are two of these compressors. One for each Train and the compressor supplies the associated train equipment of both Unit I and Unit 2. The B Auxiliary Air Compressor is powered from a Unit 2 board and supplies air to the safety related Train B components on Unit I.

A. Correct, the B Auxiliary Air Compressor is supplied from 480V C & A Vent Board 2B1-B.

B. Incorrect, Plausible because this board is the corresponding board on the opposite unit.

C. Incorrect, Plausible because this board is a similar board that supplies safety related equipment.

D. Incorrect, Plausible because this board is a similar board that supplies safety related equipment.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 54 Tier: 2 Group 1 KIA: 078 K2.02 Instrument Air System (lAS)

Knowledge of bus power supplies to the following:

Emergency air compressor Importance Rating: 3*3* /35*

IOCFRPart55: 41.7 IOCFR55.43.b: Not applicable K/A Match: K/A is matched by requiring the applicant to identify the bus power supply to one the Auxiliary Air Compressors whih are the compressors that supply essentail air loads.

Technical

Reference:

SOI-32.02, Auxiliary Air System, Rev 0019 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO32A

4. Describe the station air compressors; include cooling methods, lubrication, capacities, power supplies, stages, and operation.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN bank question SYS 032B.06 002 with the correct answer relocated and one distractor replaced. No significant change to the stem.

Comments:

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

55. 103 A3.01 055 c Given the following:

- A large break LOCA has occurred on Unit 1.

- During performance of a procedure the position of 1-HS-81-12, Primary Water To PRT and Standpipes, is required to be checked closed.

Which ONE of the following indicates indication(s) available to determine the valve position?

A. On 1-M-5, only B. On 1-M-5 or on the CVI CNTMT ISOL STATUS PNL on 1-M-6 C On 1 -M-5 or on the Phase A CNTMT ISOL STATUS PNL on 1 -M-6 D. On 1-M-5 or on the Phase B CNTMT ISOL STATUS PNL on 1-M-6 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because there is indication on 1-M-5 but it is not the only indication of the valve position in the MCR.

B. Incorrect, Plausible because there is indication on 1-M-5 and on the Containment Isolation Status Pane) but the indication is on the Phase A isolations not the CVI isolations.

C. Correct, 1-FCV-81-12, Primary Water to Containment, receives a signal to close directly from the Phase A Containment Isolation signal and there are indications that can be monitored on both 1-M-5 and on the Phase A lights on 1-M-6 Containment Isolation Status Panel.

D. Incorrect Plausible because there is indication on 1-M-5 and on the Containment Isolation Status PaneI but the indication is on the Phase A isolations not the Phase B isolations.

Question Number: 55 Tier: 2 Group 1 K/A: 103 A3.01 Containment System Ability to monitor automatic operation of the containment system, including:

Containment isolation Importance Rating: 3.9 I 4.2

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 IOCFRPart55: 41.7/45.5 IOCFR55.43.b: Not applicable K/A Match: Question requires the applicant to be able to monitor the status of a component during conditions where containment isolation has occurred.

Technical

Reference:

1-47W61 1-88-1 R23 1-47W611-81-1 R7 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO88A

7. Correctly locate control room controls and indications associated with the Containment Isolation system, including:

a. Phase A and CVI hand switches/controls.

b. Phase B Actuation and reset controls.

c. Status Panel configuration and switch position indications associated with Phase A, Phase B, CVI.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

56. 002 K6.07 056 Given the following conditions:

- Unit 1 is in MODE 3 with RCS at 557° F and 2235 psig.

- Reactor trip breakers are closed.

Which ONE of the following identifies...

(1) a combination of RCP5 that could be shutdown for maintenance leaving the pressurizer sprays effective in controlling pressure and (2) if Tech Spec LCO 3.4.5 RCS Loops MODE 3 would be required to be entered while the 2 RCP5 were shutdown?

Pressurizer Sprays LCO 3.4.5 A. RCPs #1 and #3 Entry required B RCPs #1 and #4 Entry NOT required C. RCPs #2 and #3 Entry required D. RCPs#2and#4 Entry NOT required

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because if RCP #1 and #3 are shutdown, RCP #2 remains in service to provide effective pressurizer spray and if the 2 RCPs were shutdown in Mode I or 2 then the LCO entry would be required. The mode applicability could be mistaken for conditions when reactor trip breakers are closed because the breakers being closed does affect other T/S LCOs B. Correct, With the RCP #1 and #4 shutdown, the pressurizer sprays will remain effective because the #2 RCP will still be in service. As identified in a NOTE in ES-O. I, the pressurizer sprays (either loop spray valve is effective if Loop 2 RCP in service or if Loops 1, 3, & 4 RCPs in service. The Tech Spec requirement for Mode 3 with the Reactor Trip breakers closed is for at least 2 RCPs to be running thus with RCP #2 and #3 running the Tech Spec is met and LCO entry is not required.

C. Incorrect, Plausible because there is a spray line supplied from Loop #1 and RCP

1. 1 will be running in this choice and if the 2 RCPs were shutdown in Mode I or 2 then the LCO entry would be required and the mode applicability could be mistaken for conditions when reactor trip breakers are closed because the breakers being closed does affect other T/S LCOs D. Plausible because there is a spray line supplied from Loop #1 and RCP #1 will be running in this choice and because the LCD entry not being required is correct.

ES-O.1 NOTE Either Loop 1 or 2 pzr spray valve is effective for Loop 2 RCP in service or for Loops 1, 3, & 4 RCPs in service.

LCD 3.4.5 Two RCS loops shall be OPERABLE, and either:

a. Two RCS loops shall be in operation when the Rod Control System is capable of rod withdrawal; or

b. One RCS loop shall be in operation when the Rod Control System is not capable of rod withdrawal.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 56 Tier: 2 Group 2 K/A: 002 K6.07 Reactor Coolant System (RCS)

Knowledge of the effect or a loss or malfunction on the following RCS components:

Pumps Importance Rating: 2.5 I 2.8 IOCFRPart55: 41.7/45.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of how the RCS (pressurizer sprays and Tech Spec) is affected by a loss of pumps (RCPs)

Technical

Reference:

ES-0.1, Reactor Trip Response, Revision 22 Tech Spec 3.4.5 GO-i, Unit Startup From Cold Shutdown To Hot Standby, Revision 0066 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68B

03. Given the RCS condition/status and number of RCPs/RHR pumps in service, use Tech Specs to determine if operability requirements are met and if actions are required.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question Comments: Should the first or be an of in the KA text?

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

57. 014 A1.04 057 Given the following:

- During a power reduction from 100% power, a control Bank D rod became misaligned from the bank step counter by greater than 12 steps.

Which ONE of the following annunciator windows would alarm due to the condition and a problem the misalignment presents?

A. 86-C CERPI TROUBLE Non-conservative Rod Insertion Limit calculation B. 86-C CERPI TROUBLE Challenge to core power distribution limits C. 83-D PLANT COMPUTER GENERATED ALARM (SEE ICS)

Non-conservative Rod Insertion Limit calculation D 83-D PLANT COMPUTER GENERATED ALARM (SEE ICS)

Challenge to core power distribution limits DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because window 86-C CERPI Trouble is caused by other conditions in the rod position indicating system and because the alarm is generated as a result of rod demand position versus calculated position and the conservatism of the calculated position is discussed in the rod insertion limit alarms in the same AR!.

B. Incorrect, Plausible because window 86-C CERPI Trouble is caused by other conditions in the rod position indicating system and because the conditions causing a challenge to the core power distributions is correct.

C. Incorrect, Plausible because window 83-D PLANT COMPUTER GENERA TED ALARM (SEE ICS)i is correct and because the alarm is generated as a result of rod demand position versus calculated position and the conservatism of the calculated position is discussed in the rod insertion limit alarms in the same ARI.

D. Correcl A rod being greater than 12 steps from the demanded position (step counter) will cause window 83-D PLANT COMPUTER GENERATED ALARM (SEE ICS) to alarm and misaligned rods present a challenge to core power distribution limits.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 57 Tier: 2 Group 2 KIA: 014A1.04 Rod Position Indication System (RPIS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including:

Axial and radial power distribution Importance Rating: 3.5 3.8 IOCFRPart55: 41.5/45.5 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of RPI system annunciators and knowledge of the effect on a core power distribution limit of a rod misalignment identified by the system.

Technical

Reference:

ARI-81-87, NIS & Rod Controls, Rev. 0032 Windows 83-D, 86-C, 87-A, 87-B Proposed references None to be provided:

Learning Objective: 3-OT-A010200

06. [Describe adverse effects of a Misaligned rod at power. (SOER 84-02, Rec 7b)]

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

58. 015 A3.03 058 Given the following:

- Unit 1 reactor startup is in progress and the OAC announces the reactor is critical.

- Window 65-D, P-6 INTERM RANGE PERMISSIVE, alarms on the Bypass, Intk & Permissive annunciator panel.

Which ONE of the following identifies...

(1) the condition required to cause window 65-D to light and (2) how the window will be affected when the operator manually blocks the Source Range Reactor Trip?

A. (1) When either IRM rises above 1.66 X10  % power 4

(2) The window will go DARK.

B (1) When either IRM rises above 1.66 X10  % power 4

(2) The window will remain LIT.

C. (1) Only when both IRMs rise above 1.66 X10  % power 4

(2) The window will go DARK.

D. (1) Only when both IRMs rise above 1.66 X10  % power 4

(2) The window will remain LIT.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the window being lit when either lR channel is greater than or equal to 1.66 X 1O% RTP is correct and the permissive will go DARK when condition are met as power is dropping B. Correct, As identified in AR! for window 65-D Relay K1701 (to light the window) is actuated when either IR channel is greater than or equal to 1.66 X 1O% RTP and when the Source Range Trip is blocked the P-6 INTERM RANGE PERMISSIVE window will remah lit.

C. Incorrect, Plausible because both IR channels being less than 1.66 X 1O% RTP is a condition required to allow the SRMs to automatically reinstate as power is dropping and the permissive will go DARK when condition are met as power is dropping D. lncorrect Plausible because both IR channels being less than 1.66 X 1O% RTP is a condition required to allow the SRMs to automatically reinstate as power is dropping and the window remaining lit is correct.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 58 Tier: 2 Group 2 K/A: 015 A3.03 Nuclear Instrumentation System (NIS)

Ability to monitor automatic operation of the NIS, including:

Verification of proper functioning/operability Importance Rating: 3.6 I 3.9 10 CFR Part 55: 41.7 / 45.5 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the ability to recognize conditions that will allow the manual blocking of an automatic function of an NI and the ability to verify from permissive windows the proper functioning of permissives.

Technical

Reference:

ARI-64-70, Bypass, Intlk, & Permissive, Rev. 0009 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO92A

10. Identify all indications, alarms, permissives and trips associated with the SRMs.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO92A.10 002 modified Comments:

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010

59. 016 K3.06 059 Given the following:

- Unit I is in MODE 4.

- Steam Generator levels are being maintained at setpoint using the motor-driven AFW pumps.

Which ONE of the choices below completes the following two statements?

1-PCV-3-122, AUX FEEDWATER PMP lA-A DISCHARGE PRESS CONTROL, will CLOSE if the measured differential pressure signal that inputs to 1-PDIC-3-122A, AFW PMP A-A Disch Press Control, fails The valve could then be reopened by placing 1-PDIC-3-122A in manual and the controller output.

A Low; Raising B. Low; Lowering C. High; Raising D. High; Lowering

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Correct, the system pressure signal is a differential pressure across the AFW pump IA-A. The valve purpose is to prevent pump runout by maintaining a differential pressure across the pump. If the differential pressure across the pump was indicating low, the valve would close to prevent potential for pump runout and to correct the MCR operator would be required to place the PlC in manual and open the valve by raising the output signal from the PlC.

B. Incorrect; Plausible because the signal failure low causing the valve to close is correct and other plant controllers are designed so that lowering the output signal will result in the valve being opened (ex. I-HIC-62-93)

C. Incorrect; Plausible because if the controller function was to protect against overpressure then a high failure of the pressure input would cause the valve to close (There are PCVs on plant systems that control in this manner.) and raising the controller output in manual for this controller to open the valve is correct.

D. Incorrect, Plausible because if the controller function was to protect against overpressure then a high failure of the pressure input would cause the valve to close (There are PCVs on plant systems that control in this manner.) and other plant controllers are designed so that lowering the output signal will result in the valve being opened (ex. I-HIC-62-93)

Question Number: 59 Tier: 2 Group 2 K/A: 016 K3.06 Non-Nuclear Instrumentation System (NNIS)

Knowledge of the effect that a loss or malfunction of the NNIS will have on the following:

AFW system Importance Rating: 35* / 37*

10 CFR Part 55: 41.7 / 45.6 IOCFR55.43.b: Not applicable K/A Match: KA is matched because he question requires knowledge of how a failure of the pressure input (Non-Nuclear Instrumentation) effect the AFW system and how the operator would compensate for the failure of the instrument Technical

Reference:

1-47W611-3-3 RiO Picture of MCR controllers 1-PDIC-3-122A

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Technical

Reference:

1-47W611-3-3 RIO Picture of MCR controllers l-PDIC-3-122A

&1 -HIC-62-93 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO3B

19. Identify the pressure at which the Motor-Driven AFW pump discharge pressure control valve should be set.

37. Correctly locate control room controls and indications associated with the AFW system, including: (NOTE 2)

a. AFWPs

b. AFW regulating valve

c. SIG level

d. CST level Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

60. 028 K2.01 060 Which ONE of the following describes the direct source of 480v power to the Containment Hydrogen Recombiners?

A Reactor Vent Boards B. Reactor MOV Boards C. C & A Bldg Vent Boards D. Aux Building Common Board Buses DISTRACTOR ANAL YSIS:

A. Correct, the hydrogen recombiners are supplied directly from the Reactor Vent boards.

B. Incorrect, Plausible because the Reactor MOV Boards supply equipment plant primary side equipment.

C. Incorrect, Plausible because the C &A Vent Boards supply equipment plant primary side equipment.

D. lncorreci, Plausible because the aux Building Common Board Buses supply equipment plant primary side equipment.

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 Question Number: 60 Tier: 2 Group 2 KIA: 028 K2.01 Hydrogen Recombiner and Purge Control System (HRPS)

Knowledge of bus power supplies to the following:

Hydrogen recombiners Importance Rating: 2.5* I 2.8*

IOCFRPart55: 41.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched becasue the question requires the applicant to identify the power supply for the Hydrogen Recombiners.

Technical

Reference:

SO1-83.01, Containment Hydrogen Recombiners Rev. 0015 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO83A No objective identified Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN Bank question 028 K2.01 001 with one distractor changed, very minor stem change, and the correct answer relocated Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

61. 035 G2.1.20 061 Given the following:

- Unit us at 100% power.

- SGBD is in service with 1-HS-15-44, SG BLOWDOWN DISCH TO CTBD, is in the OPEN position.

- 1-HS-15-44 is to be returned to AUTO position in accordance with SQl-I 5.01, Steam Generator Blowdown System.

Which ONE of the following identifies...

(1) if the hand switch key would be required to change the position of the switch from OPEN toAUTO and (2) how the automatic isolation on the valve is affected when the hanswitch position is changed to AUTO?

A. (1) Yes, the key is required for all operations of the handswitch.

(2) A high rad isolation will be placed in service.

B (1) Yes, the key is required for all operations of the handswitch.

(2) A low dilution flow isolation will be placed in service.

C. (1) No, the key is only required when going from AUTO to OPEN.

(2) A high rad isolation will be placed in service.

D. (1) No, the key is only required when going from AUTO to OPEN.

(2) A low dilution flow isolation will be placed in service.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the switch does require a key to change the position of the switch from OPEN to AUTO and the isolation due to high radiation is another automatic isolation of the SGBD to the cooling tower blowdown line.

B. Correct, SGBD handswitch 1-HS-15-44 is a key lock switch that requires a key to allow the position of the switch to be changed and if the position is changed from OPEN to AUTO the low dilution flow isolation function will be placed in service.

C. Incorrect Plausible because the requirement for needing a key to change the position of the switch from AUTO to OPEN is removing a condition that would automatically isolate the system overboard flow path (same concept as needing a key to bypass interlocks on refueling equipment) and the isolation due to high radiation is another automatic isolation of the SGBD to the cooling tower blowdown line.

D. lncorrecI, Plausible because the requirement for needing a key to change the position of the switch from AUTOto OPEN is removing a condition that would automatically isolate the system overboard flow path (same concept as needing a key to bypass interlocks on refueling equipment) and placing the low dilution flow isolation in service when going from OPEN to AUTO is correct.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 61 Tier: 2 Group 2 K/A: 035 G2.1.20 Steam Generator System (S/GS)

Ability to interpret and execute procedure steps.

Importance Rating: 4.6 / 4.6 IOCFRPart55: 41.10/43.5/45.12 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of required conditions when executing steps associated with the SG systems (SGBD).

Technical

Reference:

SQl-I 5.01, Steam Generator Blowdown System, Revision 0057 Proposed references None to be provided:

Learning Objective: 3-QT-SYSO15A

5. Describe the automatic response of the Steam Generator Blowdown System to a high radiation signal detected by the blowdown radiation monitor (RM-90-120, 121).

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN Bank question SYSO15A.05 002 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

62. 056 G2.1.32 062 Which ONE of the following identifies the normal Condensate operating temperature limit in accordance with SOl-2&3.01, Condensate and Feedwater System, and the purpose of the limit?

A. 140°F, to limit backpressure in the MFPT Condensers.

Bb 140°F, to prevent damage to Condensate Demin Resin.

C. 150°F, to limit backpressure in the MFPT Condensers.

D. 150°F, to prevent damage to Condensate Demin Resin.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because 140°F being the normal limit is correct and with Condensate flowing through the MFPT condenser tubes, elevated condensate temperature will cause higher backpressure in the MFPT condensers.

B. Correci The normallimit as identified in SOI-2&3.01 Precautions and Limitations is 140°F operating above 140°F can damage Condensate Demin resin.

C. Incorrect, Plausible because 150°F is a temperature identified in SOl-2&3.Q1 Precautions and Limitations section (see below) that results in isolation of the a SGBD flow path into the condensate system and with Condensate flowing through the MFPT condenser tubes, elevated condensate temperature will cause higher backpressure in the MFPT condensers.

D. lncorrect Plausible because 150°F is a temperature identified in SOl-2&3.01 Precautions and Limitations section (see below) and to prevent damage to the Condensate Demin resin is the reason for the normal temperature limit.

SO l-2&3 .01 Precautions and Limitations B. Common/Misc Systems

6. To protect the condensate demins, 1-TS-1 5-43 alarms at greater than or equal to 145°F blowdown fluid temperature and closes 1-FCV-15-43 at greater than or equal to 150°F.

C. Condensate

3. Normal Condensate operating temp limit is 140°F. Operating above 140°F can damage Cond Demin Resin.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 62 Tier: 2 Group 2 K/A: 056 G2.1.32 Condensate System Ability to explain and apply system limits and precautions.

Importance Rating: 3.8 / 4.0 10 CFR Part 55: 41.10/43.2/45.12 10CFR5543.b: Not applicable K/A Match: KA is matched because the question requires the knowledge a condensate system limit and precaution and the ability to explain the reason for the limit.

Technical

Reference:

SOl-2&3.01, Condensate and Feedwater, Revision 0109 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO2A

31. Discuss precautions and limitations necessary for operation of the condensate system, per SO 1-2 &

3.01, CONDENSATE AND FEEDWATER SYSTEM.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question SYSO14A.08 003 changed but not called a bank modified question.

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

63. 072 K4.03 063 With Unit 1 operating at 100% power the following occurs:

- Annunciatorwindow 184-B SEP 0-RM-102/103 RAD HI alarms.

- 0-RM-90-102, Spent Fuel Pit Area, has the RED HIGH light lit.

- 0-RM-90-1 03, Spent Fuel Pit Area, has only the GREEN OPERATE light lit.

Which ONE of the following identifies the status of the ABGTS Cleanup Fans and the release point for any release from the Aux Building Ventilation system?

A Only Train A ABGTS fan running and the release point is through a Shield Building Vent.

B. Only Train A ABGTS fan running and the release point is through the Auxiliary Building Vent.

C. Only Train B ABGTS fan running and the release point is through a Shield Building Vent.

D. Only Train B ABGTS fan running and the release point is through the Auxiliary Building Vent.

DISTRA CTOR ANAL YSIS:

A. Correct, the Spent Fuel Pit radiation monitors are Train specific. O-RM-90-102 is for Train A. The detection of high radiation by the monitor will cause an isolation of the Auxiliary Building and start ABGTS Cleanup Fan A. The fan will release through the Unit I Shield Building Vent.

B. Incorrect, Plausible because the Train A ABGTS fan running is correct and the Auxiliary Building Exhaust fans normally release through the Auxiliary Building Vent.

C. Incorrect, Plausible because the Train B ABGTS fan would be running if O-RM-9O-I03 had been the monitor with high radiation and the release point be a Shield Building Vent is correct.

D. Incorrect, Plausible because the Train B ABGTS fan would be running if O-RM-9O-103 had been the monitor with high radiation and the Auxiliary Building Exhaust fans normally release through the Auxiliary Building Vent.

Question Number: 63 Tier: 2 Group 2 KIA: 072 K4.03

08/2010 Watts Bar RD NRC Exam as Submitted 7/2/2010 K/A: 072 K4.03 Area Radiation Monitoring (ARM) System Knowledge of ARM system design feature(s) and/or interlock(s) which provide for the following:

Plant ventilation systems Importance Rating: 3.2*! 3.6*

10 CFR Part 55: 41.7 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of design features and interlocks associated with the Area Radiation Monitoring (ARM) System to control ventilation equipment and flow paths during conditions when radiation is detected by only one of the two area radiation monitors on the system.

Technical

Reference:

ARI-1 80-1 87, Common Radiation Detectors, Revision 31 Window 184-B 1-47W610-90-1 R37 1-47W61 1-30-5 R7 1-47W611-30-6 R13 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO3OB

12. Describe the automatic start signals for the ABGTS.

13. Given a set of plant conditions, determine the correct response of the ABGT system.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO3O 002 modified Comments:

08/2010 Watts Bar RD NRC Exam as Submiffed 7/2/2010

64. 079 K1.01 064 Which ONE of the following identifies a condition on the Control And Station Air System that will result in Station and Control Air Compressor A starting in a fully loaded condition?

A. 125v DC vital Battery Board II is deenergized with 0-XS-32-5049, STA AIR COMPR PANEL CNTL POWER XFER, in NORMAL when the compressor is started.

B. 0-HS-32-25D, STATION AIR COMPR A MAN/AUTO SELECTOR, is in HAND when the compressor is started.

C. Compressor starting with the 0-HS-32-125, STATION AIR COMPRESSOR A/B/C UNLOADING SOL, in the OFF position when the compressor is started.

D Control and Station Air System is in. a totally depressurized condition when the compressor is started.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because a 125v DC Vital Battery Board Ills the normal power supply to the control panel and a loss of the power supply does affect the air compressors ability to load.

B. Incorrect, Plausible because placing the handswitch to HAND position does affect the compressor operation related to Ioaded but would not cause it to start fully loaded. The switch is HAND with the compressor running prevent the compressor form stopping if it has been running for >1 Ominutes unloaded.

C. Incorrect, Plausible because the handswitch is normally in the ON position and placing the switch to OFF does affect the loading process. See SOI-32.O1 Precaution E below D. Correct As identified in SOI-32.O1 Precaution D (see below), if the air compressor is started with the system totally depressurized there would be no air pressure to allow the compressors to start unloaded. The unloading valves required air pressure to hold the valve open preventing the compressor from loading.

SO 1-32.01 Precautions and Limitations D. With system totally or partially depressurized, insufficient air press may be supplied to unloaders, causing compressors to fully load when started.

E. Loss of power to Compressors A, B, or C unloader solenoid causes compressor to unload. If this occurs, compressor can be manually loaded by closing the unloading solenoid isolation valve, and opening the local unloading solenoid vent valve. Both isol and vent valves are located upstream of the solenoid valve.

G. 0-HS-32-125, hand loading panel switch is normally ON except when it is desired to take the Auto-Sequence relays out of service. Placing switch in OFF removes the auto loading and/or starting of the A, B, or C compressors from sequencer relay cabinet.

Question Number: 64 Tier: 2 Group 2 KIA: 079 K1.01 Station Air System (SAS)

Knowledge of the physical connections and/or cause effect relationships between the SAS and the following systems:

lAS Importance Rating: 3.0 / 3.1 10 CFR Part 55: 41.2 to 41.9 / 45.7 to 45.8 IOCFR55.43.b: Not applicable

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 KIA Match: KA is matched because the question requires knowledge of the connections to and how the Station & Control Air Compressors are loaded/unloaded by the pneumatic instrument air sequencer and unloading solenoids.

Technical

Reference:

SOl-32.01, Control Air System, Rev. 0050 AOl-21.02, Loss of 125V Vital Battery BD II, Revision 20 1-47W846-1 R38 1-47W848-1 R25 1 -47W848-2 R29 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO32A

9. Explain how the sequencing device operates to control the control air compressors in Hand and Auto.

10. Explain how the control air compressors are affected by a loss of power to the sequencer.

11. Explain how the control air compressors may be loaded manually.

Cognitive Level:

Higher X Lower Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

65. 086 A2.04 065 With Unit 1 operating at 100% reactor power the following occurs:

- The fire protection system detects a fire in the 2A-A Diesel Generator Room but the automatic fire suppression system fails to actuate.

- Personnel at the DG building confirm the fire exists and report the Manual-electric push button outside the room door has failed to initiate the system.

Which ONE of the following identifies...

(1) an impact of the failure of the fire suppression system to initiate has on the fire dampers and doors, and (2) the actions directed in SOl-39.02, DG C02 System. to manually initiate C02?

A. (1) Fire dampers and doors will remain open until C02 is available to release the latches.

(2) Place DG Room 2A-A Pilot Valve to open using the lever outside the room door which allows C02 pressure to open both the room routing valve and the master routing valve in the C02 tank room.

B. (1) Fire dampers and doors will close when the heat from the fire melts fusible links.

(2) Place DG Room 2A-A Pilot Valve to open using the lever outside the room door which allows C02 pressure to open both the room routing valve and the master routing valve in the CO2 tank room.

C. (1) Fire dampers and doors will remain open until C02 is available to release the latches.

(2) Open the master routing valve in the CO2 tank room, then place DG Room 2A-A Pilot Valve to open using the lever outside the room door.

Dv (1) Fire dampers and doors will close when the heat from the fire melts fusible links.

(2) Open the master routing valve in the CO2 tank room, then place DG Room 2A-A Pilot Valve to open using the lever outside the room door.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because without C02 being released the latches will not release to initiate closing the dampers & doors. Also because after the master valve is set to open, there will be C02 available to open the room valve (operation opposite of what is stated in choice)

B. lncorrect, Plausible because the fire dampers and doors being released due to the fusible links melting is correct and because after the master valve is set to open, there will be C02 available to open the room valve (operation opposite of what is stated in choice)

C. Incorrect, Plausible because without C02 being released the latches will not release to initiate the closing the dampers & doors and because open the master valve then placing the room pilot valve to open is correct.

D. Correct, In addition to having a release latch operated by the C02 as the header is pressurized, the fire dampers and doors are equipped with a fusible link that will melt the restraint cabling to release the doors and dampers. Both the room valve and the master valve must be manually actuated to release C02 into the room. The master valve is opened which provides C02 to allow the room valve to be opened when its pilot valve is set to the open position.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 65 Tier: 2 Group 2 K/A: 086 A2.04 Fire Protection System (FPS)

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: 41.5/43.5/45.3/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires the prediction all components of the fire protection system would be impacted by the fire and how the system would be initiated in accordance with procedrures to mitigate the consequences of the fire.

Technical

Reference:

SOl-39.02, DG 002 System, Revision 17 3-OT-SYSO39A Proposed references None to be provided:

Learning Objective: 3-OT-SYS39A

14. Describe the automatic, manual and manual electric methods of 002 fire protection initiation listing the sequence of events when a high temperature is detected in an area of CO 2 protection.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

66. G2.1.15 066 Which ONE of the following identifies the maximum time Standing Orders and Shift Orders should normally remain in effect in accordance with 0DM-i .0, Standing Orders and Shift Orders?

Standing Orders Shift Orders A. 60 days 7 days B. 60 days 30 days C. 1 year 7 days D 1 year 30 days DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because 60 days is a time period for other conditions in operations (SI frequency, LERs) and 7 days is a time period for other things in operations (LCO times, SI times, etc.)

B. Incorrect, Plausible because 60 days is a time period for other conditions in operations (SI frequency, LERs) and 30 days being the maximum time for a Shift Order is correct.

C. Incorrect, Plausible because 1 year being the maximum normal time for a Standing Order is correct and 7 days is a time period for other things in operations (LCO times, SI times, etc.)

D. Correct, Operations Directive Manual (0DM) -1 states that a Standing Order should not remain in effect beyond 1 year and a Shift Order should not be listed for more than 30 days (see excerpt below) 0DM-i, Standing Orders and Shift Orders I. Standing Orders G. Standing Orders should not be in effect for more than one year. They may be extended beyond one year at the discretion of the Operations Superintendent.

II. Shift Orders C. Shift Orders should not be listed for more than 30 days.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 66 Tier: 3 Group n/a K/A: G2.1.15 Conduct of Operations Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.

Importance Rating: 2.7 / 3.4 10 CFR Part 55: 41.10 /45.12 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of administrative requirements (maximum time restrictions) for standing orders and night orders, Technical

Reference:

0DM-i, Standing Orders and Shift Orders, Revision 0 Proposed references None to be provided:

Learning Objective: 3-OT-SPP1 000 Conduct of Operations No objective Identified Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question SPP1000.01 006 updated to current procedure and format changed Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

67. G2.1.18 067 Given the following:

- An individual is in the process of reactivating his/her license in accordance with OPDP-10, License Status Maintenance, Reactivation and Proficiency for Non-Licensed Positions.

Which ONE of the following completes the two statements below?

In addition to being documented in Appendix A, Return to Active Status Checklist, narrative log entries are required to be made for (1)

After the license reactivation is complete and the individual is performing the first shift turnover in accordance with OPDP-1, Conduct of Operations, the requirement for review of the Operating Logs is (2) or last shift held, whichever is less.

A. (1) ppjy at the beginning and end of each shift completed.

(2) 3 days B. (1) çjjjy at the beginning and end of each shift completed.

(2) 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> C (1) at the beginning and end of each shift completed and for the plant tour.

(2) 3 days D. (1) at the beginning and end of each shift completed and for the plant tour.

(2) 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the log entry required for the beginning/end of each shift worked (and the plant tour is documented in the appendix) and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> is a time that appears in the reactivation procedure.

B. Incorrect, Plausible because the log entry required for the beginning/end of each shift worked (and the plant tour is documented in the appendix) and 3 days is correct.

C. Correct, OPDP-10 requires narrative log entries for both the beginning/end of each shift worked during reactivation and also for the completion plant tour. The requirements appear in the body of the instruction and in the Appendix used for documentation of completion. OPDP- I requires a review of the Operating Logs for the last 3 days or last shift helci whichever is less.

D. Incorrect, Plausible because the log entry requirement is correct and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> is a time that appears in the reactivation procedure.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 67 Tier: 3 Group n/a KIA: G2.1.18 Conduct of Operations Ability to make accurate, clear, and concise logs, records, status boards, and reports.

Importance Rating: 3.6 / 3.8 IOCFRPart55: 41.10/45.12/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of activities that are required to be included in the operating log in order that the log will be accurate and allow reactivation process to be validated and documented.

Technical

Reference:

OPDP-10, License Status Maintenance, Reactivation and Proficiency for Non-Licensed Positions, Rev 0001 OPDP-1, Conduct of Operations, Rev 0015 Proposed references None to be provided:

Learning Objective: 3-OT-SPP1000

12. Describe the information to be recorded in Operating logs.

21. Describe the responsibilities and requirements for reactivation of an inactive license. (OPDP-10)

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question PAIO2O.03 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

68. G2.1.45 068 Given the following:

- Unit 1 is operating at 100% reactor power.

- Annunciator window 89-E PZR SPRAY TEMP LO alarms.

Which ONE of the following identifies...

(1) the condition that would cause the alarm and (2) a diverse indication that would validate the condition did exist?

Condition Diverse Indication A. Pressurizer spray Pressurizer surge line bypass flow LOW temperature RISING B Pressurizer spray Pressurizer surge line bypass flow LOW temperature DROPPING C. Pressurizer spray Pressurizer surge line bypass flow HIGH temperature RISING D. Pressurizer spray Pressurizer surge line bypass flow HIGH temperature DROPPING

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the condition being caused by low pressurizer spray bypass flow is correct and the pressurizer surge line temperature will change but the affect will not be for the temperature to rise.

B. Correct, the alarm would be caused by low pressurizer spray bypass flow and the low flow would cause reduced pressurizer out flow through the pressurizer surge line causing the surge line temperature to drop.

C. Incorrect, Plausible because the alarm is caused by improper flow though the pressurizer spray bypass line but it is not excessive flow through the line (it is low flow) and the pressurizer surge line temperature will change but the affect will not be for the temperature to rise when the spray bypass flow is low.

D. Incorrect, Plausible because the alarm is caused by improper flow though the pressurizer spray bypass line but it is not excessive flow through the line (it is low flow) and the pressurizer surge line temperature will is correct when the spray bypass flow is low.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 68 Tier: 3 Group n/a KIA: G2.1.45 Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication.

Importance Rating: 4.3 / 4.3 10 CFR Part 55: 41.7/43.5 I 45.4 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of how to interpret changes to MCR indications during operation and how to use a diverse indication to confirm the response of an indication is correct.

Technical

Reference:

ARI-88-94, Reactor Coolant System, Rev 19 Windows 89-D and 89-E Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68C

3. Describe the purposes of the Manual Bypass Pressurizer Spray Throttle Valves.

22. Explain the operation of major system components.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN question SYSO68C.22 031 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submiffed 7/2/2010

69. G 2.2.15 069 While performing a system valve checklist for verification of system alignment, the AUO reports a valve has been identified as being out of the position required by the checklist being performed.

Which ONE of the following identifies pjjjy approved plant processes which could be controlling the valve in the different position in accordance with SPP-10.1, System Status Control?

A. Caution Tag TACF Work Order B Hold Order TACF Work Order C. Caution Tag Hold Order Work Order D. Caution tag Hold Order TACF

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because a TACF and a work order are correct and the Caution tag identifies abnormal conditions and can be placed on valves when off normal conditions exist. But as identified in SPP-1O. 1, Caution tags shall not be used to authorize, control or establish equipment status changes.

B. Correct As identified in SPP-1O. 1, System Status ControI the processes that change the status of plant equipment are approved plant procedures, clearance (Hold Order), work order, or TA CF.

C. Incorrect, Plausible because a hold order and a work order are correct and the Caution tag identifies abnormal conditions and can be placed on valves when off normal conditions exist. But as identified in SPP-1O. 1, Caution tags shall not be used to authorize, control or establish equipment status changes.

D. Incorrect, Plausible because a work order and a hold order are correct and the Caution tag identifies abnormal conditions and can be placed on valves when off normal conditions exist. But as identified in SPP-1O. 1, Caution tags shall not be used to authorize, control or establish equipment status changes.

sPP-10.1 3.3.1 .C Performance. If during the performance of a checklist, a component is not in the specified position/status:

1. The operator shall notify the UO.

2. The UO shall determine if the component should be repositioned based on a review including the following:

a. Procedures currently in effect.

b. Clearances affecting the component.

c. Temporary Alterations.

d. Work currently in progress.

3. The UO shall notify the US of the deviations.

4. The US shall determine if the condition needs to be documented in accordance with the TVAN procedure on Corrective Action Program.

3.5 Work Documents 3.5.1 Work Documents Affecting Status A. Work activities that change the status of components must be authorized and documented by one of the following:

1. SPP-1O.2, Clearance Program.

2. SPP-6.1, Work Order Process.

3. Approved plant procedures.

4. SPP-9.5, Temporary Alterations.

page 5

12. Caution tags shall not be used to authorize, control or establish equipment status changes.

13. Ensuring that Caution tags are placed on valves when a valve dog is placed in an off normal status (engaged or disengaged). Caution tags should be placed on other devices as deemed necessary to identify that an off normal status exists.

D. The Responsible Individual shall:

1. Ensure procedures and work documents restore systems and equipment to the correct status.

2. Ensure all activities that change the status of plant equipment are authorized by an approved plant procedure, clearance, work order or TACF.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 69 Tier: 3 Group n/a KIA: Equipment Control 2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

Importance Rating: 3.9 I 4.3 IOCFRPart55: 41.10/43.3/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the processes that can be used to change the status of plant equipment while maintaining configuration control on a system when performing system lineups, tag-outs or maintenance activities.

Technical

Reference:

SPP-10.1, System Status Control, Rev. 4 Proposed references None to be provided:

Learning Objective: 3-OT-SPP1 001

07. Identify the requirements for maintaining Status Control as to:

a. How Status Control is initially established.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question SPPI 001.13 001 modified Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

70. G2.2.18 070 During a unit outage, which ONE of the following identifies...

(1) the lowest level of reduced Defense in Depth identified by Outage Risk Assessment Management (ORAM) that requires a contingency plan to be in place prior to entering the condition and (2) how protected equipment logged in the OSSDM 4.0, Operational Defense in Depth Assessment, will be tracked when the equipment is required to be protected following completion of the outage?

Note:

0DM Operations Directive Manual OSSDM Outage and Site Scheduling Directive Manual Risk Level requiring a Contingency glan Protection tracking A. Yellow Protective devices will be transferred to 0DM 4.0, Protected Equipment, log.

B. Yellow OSSDM log will remain open until equipment protection is not required.

Orange Protective devices will be transferred to 0DM 4.0, Protected Equipment, log.

D. Orange OSSDM log will remain open until equipment protection is not required.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because a risk level of Yellow is a condition where a reduction in the Defense in Depth level is occurring and transferring any protected equipment to the 0DM 4.0 log is correct.

B. Incorrect, Plausible because a risk level of Yellow is a condition where a reduction in the Defense in Depth level is occurring and keeping the OSSDM 4.0 log open to track the equipment is a means of protection.

C. Correc1, In accordance with SPP-7.2, Outage Management, a contingency Plan is required if the risk level reaches an Orange condition which is where the Defense in Depth level is significantly reduced and any equipment being protected at the completion of an outage will be transferred to the 0DM 4.0 log in accordance with ODM4. 0, Protected Equipment.

D. Incorrect, Plausible because a risk level of Orange requiring a contingency plan is correct and the Outage Group has many responsibilities during an outage and keeping the OSSDM 4.0 log open to track the equipment is a means of protection.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 70 Tier: 3 Group n/a K/A: G2.2.18 Equipment Control Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Importance Rating: 2.6 I 3.9 IOCFRPart55: 41.10/43.5/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of the process for managing risk assessments for activities during shutdown operations.

Technical

Reference:

SPP-7.2, Outage Management, Rev.0018 0DM 4.0, Protected Equipment, Revision 0 Proposed references None to be provided:

Learning Objective: No objective identified Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 71.G2.3.5o71 With Unit I operating at 100% power the following annunciator windows alarm:

174-B, 1-RR-90-1 AREA RAD HI 174-E, 1-RR-90-1 AREA MONITORS INSTR MALE If a momentary loss of power to 1-RM-90-61, Incore lnstr Room, caused the alarms, which ONE of the following identifies...

(1) the status of the radiation monitors OPERATE light (GREEN Light) on 0-M-12 and (2) when responding to the alarms, the location where the source check for the monitor can be performed?

GREEN light Source Check A. LIT Main Control Room B. LIT Locally at the monitor C DARK Main Control Room D. DARK Locally at the monitor DISTRACTOR ANAL YSIS:

A. lncorrec1, Plausible because the green light would be LIT if High Rad alarm was due to other conditions and because the Main Control Room being the location where a source check can be performed is correct.

B. Incorrect, Plausible because the green light would be LIT if High Rad alarm was due to other conditions and because the area monitor does have local indications and other monitor to have local controls.

C. Correct As identified in ARI -1748 Corrective Action [3], IF loss of power indicated (green light out and I 74-E LIT), THEN GO TO of I 74-E and the Math Control Room on O-M- 12 is the location of the controls for source checking the area monitor.

D. lncorrect Plausible because the green light being DARK is correct and because the area monitor does have local indications and other monitor to have local controls.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 71 Tier: 3 Group n/a K/A: G2.3.5 Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance Rating: 2.9 / 2.9 10 CFR Part 55: 41.11 I 41.12 I 43.4 I 45.9 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires the ability to determine status of the monitor from alarms & indications and the knowledge of where operations pertaining to the monitor can be performed.

Technical

Reference:

ARI-1 73-1 79, U-i Radiation Detectors, Rev. 0045 windows 174B and 174E SOl-90.04, Area Radiation Monitors, Rev. 0007 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO9OA

05. Describe the purpose of a source check and how to perform a source check on a Rad monitor Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

72. G2.3.11 072 Following a Unit 1 reactor trip from 100% power, the plant is being maintained in Hot Standby in accordance with GO-5, Unit Shutdown From 10% Power To Hot Standby, when the following conditions develop:

- Annunciator windows lit:

175-B VAC PMP EXH 1-RM-119 RAD HI 178-A SG BLDN 1-RM-120/121 LIQ RAD HI

- Steam Generator (SG) parameters are as follows:

SG1 SG2 SG3 SG4 NRLevel 42% 35% 38% 33%

(stable) (rising) (lowering) (rising)

AFW Flow 80 gpm 0 gpm 0 gpm 230 gpm Which ONE of the following is an action required to be taken to minimize the radiation release?

A. Raise #2 SG Atmospheric relief valve setpoint.

B. Raise #3 SG Atmospheric relief valve setpoint.

C. Isolate the Steam Supply from the #1 SG to the TD AFW Pump turbine.

D. Isolate the Steam Supply from the #4 SG to the TD AFW Pump turbine.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Correct, Steam Generator #2 is the ruptured steam generator as identified by the level rising with no AFW flow and an action to control the release of radiation is to raise the setpoint of the #2 SG Atmospheric relief valve.

B. lncorreci, Steam Generator #2 is the ruptured steam generator (not steam generator #3) but if the #3 SG had been ruptured raising the setpoint of the #3 SG Atmospheric relief valve would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.

C. Incorrect, Steam Generator #2 is the ruptured steam generator (not steam generator #1) but if the #1 SG had been ruptured isolating the steam supply from the SG#1 to the TD AFW pump turbine would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.

D. IncorrecI, Steam Generator #2 is the ruptured steam generator (not steam generator #4) but if the #4 SG had been ruptured ensuring the TD AFW pump turbine was not being supplied from the #4 SG would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.

Question Number: 72 Tier: 3 Group n/a K/A: G2.3.ii Radiation Control Ability to control radiation releases.

Importance Rating: 3.8 / 4.3 10 CFR Part 55: 41.11 /43.4/45.10 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of actions required to prevent (control) a radiation release during abnormal conditions on the unit.

Technical

Reference:

AOl-33, Steam Tube Leak, Revision 32 GO-5, Unit Shutdown From 10% Power To Hot Standby, Revision 0032 TI-i 2.04, Users Guide For Abnormal And Emergency Operating Instructions, Revision 0008

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Proposed references None to be provided:

Learning Objective: 3-OT-A013300

8. Given a set of plant conditions, use AOl-33 to correctly:

a. Recognize Entry Conditions.

b. Identify Required Actions.

c. Respond to Contingencies (RNO).

d. Observe and Interpret Cautions and Notes.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank (audit exam spring 2009 question G2.3.11 072 modified) stem conditions changed to make a different answer correct, another distractor changed and the other distractors relocated..

Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

73. G2.4.16 073 In accordance with TI-12.04, Users Guide For Abnormal and Emergency Operating Instructions, which ONE of the choices below completes the following statements?

Emergency Operating Instruction (EQ I) Network must be entered if a Reactor Trip or a Safety Injection occurred while the plant was operating in While performing ES-0.1, Reactor Trip Response, an AOl be performed concurrently with the EOl without the EOI in effect directing the AOl usage.

A. (1) Modes 1, 2, 3, and 4 (2) may B. (1) Modes 1,2, 3, and4 (2) may NOT C (1) Modes 1,2, and 3, pjjjy (2) may D. (1) Modesl,2,and3,ppjy (2) may NOT

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because implementation of the EQIs if a Reactor Trip or a Safety Injection occurs during Mode 1, 2 or 3 is correct and there is no requirement that the AOl use be directed in the EQI. (See below)

B. lncorrect Plausible because implementation of the EQ/s during Mode 4 is correct if a complete loss of shutdown power occurs and there is no requirement that the AOl use be directed in the EQI. (See below)

C. Correct, Tl-12.04 identifies that the EOls are to be implemented whenever a reactor trip or a safety injection is initiated with the unit in Modes 1, 2, or 3. It also provides for use of the AQIs performance of the EQ/s without the EOl directing the AOl use but the AOls must be on a not-to-interfere basis. There are cases where AOl use is directed by the EOl in effect, but no requirement that the AOl use must be directed in the EQI. (See below)

D. Incorrect, Plausible because implementation of the EQIs during Mode 4 is correct if a complete loss of shutdown power occurs and an AOl can be implemented during performance of the EQ/s without the EQI directing the AOl use but the AQIs must be on a not-to-interfere basis. (See below)

TI-I 2.04, Users Guide For Abnormal And Emergency Operating Instructions 2.1.2 Mode Applicability of the EOIs The EOls are written to mitigate emergency transients initiated when the unit is at hot or power conditions.

A. The guidance for operator action in the EOls assumes that the safety-related equipment required by Tech Specs in Mode 1 or Mode 2 is available for use.

B. The operating crew should implement the EOl network whenever reactor trip or safety injection events are initiated with the unit in Modes 1, 2, or 3.

C. The operating crew should implement the EOl network for the complete loss of shutdown power event with the unit in Modes 1, 2, 3, or 4.

2.8 Use of AOIs while in EOIs

1. During performance of the ES-0.1, if plant conditions warrant implementation of an AOl, then the required AOl may be performed concurrently (on a not-to-interfere basis) with the EOls.

2. When running an AOl concurrently with an EOI (ECA-0.0, ES-0.I, etc.) the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOl if another Unit Supervisor is NOT available. If the BOP/CRO operator performs an AOl, he/she should consult directly with the Unit Supervisor and give them the status as required by the AOl.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 73 Tier: 3 Group n/a K/A: G2.4.16 Emergency Procedures I Plan Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

Importance Rating: 3.5 I 4.4 IOCFRPart55: 41.10/43.5/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of how EOI usage is coordinated by AOl usage.

Technical

Reference:

Tl-12.04, Users Guide For Abnormal and Emergency Operating Instructions, Revision 8 Proposed references None to be provided:

Learning Objective: 3-OT-Tl1204

1. Describe the conditions for entry into the EOP network.

Cognitive Level:

Higher Lower X Question Source:

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

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

74. G 2.4.34 074 Given the following:

- A small break LOCA is in progress on Unit I and a loss of all offsite power has occurred.

- DG 1 B-B failed when it attempted to start.

- The crew has transitioned to ES-i .2, Post LOCA Cooldown and Depressurization, and the step to ISOLATE cold leg accumulators is being performed.

- The operator has been dispatched to perform ES-i .2 Appendix B, CLA Breaker Operation to restore power to the cold leg accumulator isolation valves (i-FCV-63-118, -98, -80, -87).

Which ONE of the following identifies the status of the CLA isolation valves and the operational status when the step is completed?

A. All four CLA isolation valves can be energized.

Close all accumulator isolation valves and the breakers are left energized.

B. All four CLA isolation valves can be energized.

Close all accumulator isolation valves and then the breakers are de-energized.

C. Only two CLA isolation valves can be energized.

Close the energized isolation valves and then vent all four of the accumulators.

D Only two CLA isolation valves can be energized.

Close the energized isolation valves and the two accumulators without power to the valves will be vented.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorreci, Plausible because normally all four would be energized and the procedure leaving power on the breakers is correct.

B. lncorrect Plausible because normally all four would be energized and the procedure directing the power be removed after the valves are closed would be typical of many ormally open breakers operations in the plant.

C. Incorrect, Plausible because only the two Train A valves being able to be closed is correct and the procedure does list all 4 of the CLA nitrogen valves but only directs opening the nitrogen vent valves on unisolated accumulators.

D. Correct, Only 2 of the isolation vales can be closed because there is no power available for the Train B valves. Thus procedure directs that the CLA be vented if its isolation valve cannot be closed.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 74 Tier: 3 Group n/a K/A: G 2.4.34 Emergency Procedures I Plan Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Importance Rating: 4.2 / 4.1 IOCFRPart55: 41.10/43.5/45.13 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of tasks performed outside the Main Control Room and the operational affect of performing the task due to a degraded power supply condition..

Technical

Reference:

ES-i .2, Post LOCA Cooldown and Depressurization, Rev. 14 Proposed references None to be provided:

Learning Objective: 3-OT-EOPOIOO

6. Explain the basis for isolating the CLAs when RCS press decreases to less than 250 psig.

8. Given a set of plant conditions, use E-i, ES-i .1, ES-i .2, ES-i .3, and ES-i .4 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes, and Cautions.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Vogtle 06/2009 question Comments:

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010

75. G 2.4.35 075 After a Safety Injection on Unit 1 resulting from a SG #1 steam line break, the following occurs:

- An AUO is dispatched to perform E-2, Faulted Steam Generator Isolation, Attachment 1 (E2) because 1-FCV-1-4, SG #1 MSIV, failed to close from its handswitch.

- The valve again failed to close when control is transferred in the Auxiliary Control Room and the AUO proceeds to remove the control fuses.

Which ONE of the following identifies...

(1) the location of the fuses that are removed during performance of the Attachment and (2) an operational effect if the valve closes following fuse removal?

A. (1) In 125V DC Vital Battery Board I fuse column, only.

(2) An alternate means to determine valve status would be required because 1-HS-1-4A indicating lights would be DARK.

B. (1) In 125V DC Vital Battery Board I fuse column, only.

(2) The valve status could be determined by the GREEN indicating light on 1-HS-1-4A being LIT.

C. (1) In 125V DC Vital Battery Boards I and II fuse columns.

(2) An alternate means to determine valve status would be required because 1-HS-1-4A indicating lights would be DARK.

D (1) In 125V DC Vital Battery Boards I and II fuse columns.

(2) The valve status could be determined by the GREEN indicating light on 1-HS-1-4A being LIT.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the MSIVs having two control circuits, one on 125V DC Vital Battery Board I and the other on 125V DC Vital Battery Board!! is different than most other components (where there is only one control circuit) and in most plant control circuits when the control power is removed the Indicating lights on the handswitch go DARK.

B. lncorrect Incorrect, Plausible because the MSIVs having two control circuits, one on 125V DC Vital Battery Board I and the other on 125V DC Vital Battery Board II is different than most other components (where there is only one control circuit) and the indicating lights on the handswitch remaining able to provide indication of the valve status after the fuses are removed is correct.

C. lncorrect, Plausible because there being two control circuits, one on 125V DC Vital Battery Board I and the other on 125V DC Vita! Battery Board Ills correct and in most plant control circuits when the control power is removed the indicating lights on the handswitch go DARK.

D. Correct The attachment directs the fuses to be removed from both control circuits.

One is supplied from 125V DC Vital Battery Board I and the other circuit is from 125V DC Vital Battery Board II. If the MSIV closes after the fuses are removed the indicating lights on the handswitch will remain able to provide indication of the valve status and if the MSIV closes, the Green light on the handswitch will be LIT.

08/2010 Watts Bar RO NRC Exam as Submitted 7/2/2010 Question Number: 75 Tier: 3 Group n/a K/A: G2.4.35 Emergency Procedures / Plan Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Importance Rating: 41.10 I 43.5 I 45.13 10 CFR Part 55: 3.8 I 4.0 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of the required actions by an auxiliary operator required during performance of emergency procedures and how the performance of the actions effect the indications available to the MCR operator (operational effect)

Technical

Reference:

E-2, Faulted Steam Generator Isolation, Rev 11 1-45W600-1-5 R8 1-45W600-1-6 R4 Proposed references None to be provided:

Learning Objective: 3-OT-EOP0200

5. Given a set of plant conditions, use procedure E-2 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question EOPO200.02 016 modified.

Comments: