ML19116A139

301 ADAMS-2A Draft Exam Material

ML19116A139
Person / Time
Site:	Watts Bar
Issue date:	04/26/2019
From:	NRC/RGN-II/DRS/OLB
To:	Tennessee Valley Authority
References
50-390/11-301, 50-391/11-301
Download: ML19116A139 (280)

Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Watts Bar Date of Examination: June 2011 Examination Level: RO LI SRO Operating Test Number: 1 Administrative Topic Tvne Jr

• Describe activity to be performed (See Note) Code A.1-1 RCS void determination per GO-b, Reactor Coolant System Drain And Fill Operations.

Conduct of Operations N,R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. 47.10/43.5/45.2/45.6 4.3/4.4 A.b-2 Complete ECP prediction using ICRR data.

2. 7.43 Ability to use procedures to determine the effects Conduct of Operations N,R on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel

.

depletion, etc. 41.10/43.6/45.6 4.7 4.3 A.2 Review 1-Sl-68-33, Measurement Of Reactor Coolant Pump Seal Injection Flow, for Approval.

Equipment Control N,R 2.2.37 Ability to determine operability and/or availability of safety related equipment. 47.7 / 43.5 / 45. 12 3.6/4.6.

A.3 Authorize a Radioactive Liquid Release.

Radiation Control M,R 2.3.6 Ability to approve release permits.47.12/45.70 2.0/3.8 A.4 Classify an Event (Earthquake and Radiation Release).

Emergency Procedures / M R Plan 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation. 47.10 / 43.5 / 45.17 2.7/4.5.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; <4 for SROs & RO retakes)

(N)ew or (M)odified from bank (> 1)

(P)revious 2 exams (< 1; randomly selected)

ES 301, Page 22 of 27

ES-301 Administrative Topics Outline Form ES-301-1 SRO Admin JPM Summary Al-I Given plant data, the applicant performs GO-lU, Reactor Coolant System Drain and Fill Operations, Appendix HH, RCS Void Determination. With the calculation completed, the applicant will determine whether or not additional RCS sweeps are required.

A.l-2 Given rod positions and source range counts, the applicant will calculate and plot the Inverse Count Rate Ratio data. After plotting the data, the applicant uses the data to determine the proper actions to take for the startup.

A.2 The applicant reviews l-Sl-68-33, Measurement Of Reactor Coolant Pump Seal Injection Flow, and determines that seal injection flow does NOT meet acceptance criteria, and that Section 6.3, Valve Adjustment is required to be performed.

A.3 The applicant reviews a Liquid Radioactive Waste Release Permit for approval. The applicant identifies three errors and determines that the permit may not be approved.

A.4 Given the data provided and associated references, the applicant evaluates EPIP-1, Emergency Plan Classification Flowpath, and determines that the earthquake satisfies conditions for an ALERT and the gas decay tank leak and associated radiological level satisfy the conditions for a SITE AREA EMERGENCY: The applicant implements EPIP-4, Site Area Emergency.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Watts Bar Date of Examination: June 2011 Exam Level: RO U SRO-l SRO-U Q Operating Test Number: I Control Room Systems@ (8 for RO); (7 for SRO-l); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

Fn

a. Borate the RCS per AOl.39,Rapid Load Reduction. A,N 004 A2.06 4.2/4.3 CFR: 41.5/43/5/45/3 /45/5 (1-M-6)

b. Synchronize IA-A DG to IA-A 6.9 Ky Shutdown Board per SOI-82.Ol,Diesel Generator (DG) C or 5, 0 6 lA-A.

064 A4.06 3.9/3.9 CFR: 41.7/45.5 to 45.8. (0-M-26)

c. Return NIS Power Range Channel N44 to service, per AOl-4, Nuclear Instrumentation A,M 7 Malfunctions.

015 A4.01 3.6/3.6 CFR: 41.7/45.5 to 45.8. (1-M-1 3)

d. Establish Manual Makeup to the VCT per SOl-62.02,Boron Concentration Control. M 2 004 A4.13 3.3/2.9 CFR: 41 .7/45.5 to 45.8 (1-M-6)

e. Place RHR Spray in Service per FR-Z.I, High Containment Pressure. A,D,L 4P 005 A4.01 3.6/3.4 CFR: 41 .7/45.5 to 45.8 (1-M-6)

f. Place Hydrogen Recombiner A in Low Power Standby per SOl-63.O1,Containment Hydrogen C or 5, M 5 Recombiners.

028 A4.01 4.0/4.0 CFR: 41.7/45.510 45.8. (1-M-10)

g. Isolate Cold Leg Accumulators using E-1, Loss of Reactor or Secondary Coolant. A, 0 3 011 EA1 .09 4.3/4.3 CFR 41.7/45.5 /45.6. (1-M-6)

h. RO ONLY(NIA)

In-Plant Systems@ (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U)

i. Local Control of 1-FCV-62-93, Charging Flow Control Valve. 0, R 004 A2.22 3.2/3.1 41.5/ 43/5 / 45/3 / 45/5+

j. Operate Steam Generator(SG) 1 Power Operated Relief Valve (PORV) Locally with N2 A,EN,M 4S 068 AA2.01 4.3/4.5 CFR: 41.7/ 45.5 / 45.6

k. Place Cation Bed in Service for High RCS Activity, per SOl-62.04, CVCS Purification System. D,E,R 9 076 AA2.02 2.8/3.4 43.5/45.13

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U fA)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect ftom bank < 9 I < 8 / < 4 (E)mergency or abnormal in-plant > 1 I > 1 I > 1 (EN)gineered safety feature - I - I > 1 (L)ow-Power I Shutdown > 1 / > 1 I > 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 I 1 (P)revious 2 exams < 3 < 3 / 2 (randomly selected)

(R)CA >1 I >1 / 1 (S)imulator

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Sum mary B.1 .a This is an ALTERNATE PATH JPM. The applicant will perform the actions to borate the Reactor Coolant System during a rapid load reduction. The applicant will determine the flow rate and the amount of boric acid to be added to the reactor coolant system for a 2%/mm load reduction rate to 80% power When it is discovered that the normal boration flow path is unavailable, the applicant will perform the actions contained in the RESPONSE NOT OBTAINED column. These actions will result in boration using 1-FCV-62-1 38, Emergency Borate Valve.

b.1 .b The applicant will synchronize the lA-A Diesel Generator to the lA-A 6.9 Ky Shutdown Board and then load the lA-A Diesel Generator to 4Mw, using SOI-82.01, Diesel Generator(DG) lA-A.

3.1 .c This is an ALTERNATE PATH JPM. The applicant will perform actions of AOI-4,Nuclear lnstwmentation Malfunctions, to return Power Range Channel N41 to service. The event which causes the JPM to be an alternate path JPM is the failure of Steam Generator (SG) 1, Main feedwater regulating valve open when the SG LEVEL-NIS BIAS controllers are returned to normal. The applicant will determine that SG level is rising uncontrollably, requiring the applicant to trip the reactor.

3.1 .d The applicant will perform SOI-62.02, Boron Concentration Control, Section 6.5, Manual Makeup, and raise VCT level from 35% to 40%, using REACTVV calculation VCT Makeup Calculation data provided.

This JPM was modified from an ALTERNATE PATH JPM to a normal JPM.

3.1 .e This is an ALTERNATE PATH JPM. The applicant will perform the alignment of the RHR system to establish RHR spray per FR-Z.l, High Containment Pressure. The event which causes the JPM to be an alternate path JPM was modified from the failure of 1-FCV-63-94 RHR TO CL 1&4 to a failure of 1-FCV-72-41 ,RHR SPRAY HDR B TO CNTMT. When the applicant attempts to open 1-FCV-72-41, the valve will NOT open. This requires entry into the RNO in order to place Train A RHR Spray in service.

B.1 .f The applicant will perform SOI-83.01, Containment Hydrogen Recombiners, Section 8.6, Placing Recombiner A in Low Power Standby (1 -H2C-83-1 ). This JPM requires the use of time compression to complete the task.

B.1 .g This is an ALTERNATE PATH JPM. The applicant performs E-l, Loss of Reactor or Secondary Coolant, Step 26 to determine if cold leg accumulators should be isolated. During the performance of the step, the applicant discovers that accumulators 1 and 2 cannot be isolated. The applicant enters the RESPONSE NOT OBTAINED column and performs actions to vent accumulators 1 and 2.

B.1.h ROONLY(NIA)

B.1 .i The applicant will take local control of l-FCV-62-93 using S01-62.0l, CVCS-Charging and Letdown, Section 8.5, Local Control of 1-FCV-62-93, CHARGING HEADER FLOW, raise the pressurizer level to 60% while maintaining contact with the Operator-At-the-Controls (OAC).

B.l .j This is an ALTERNATE PATH JPM. The applicant will take local control of Steam Generator (SG) 1 Power Operate Relief Valve (PORV) at the local N2 Control Station for #1 SG PORV with instructions to use the local N2 station per S0l-l.0l, Main Steam System, Section 8.2.3, Local Operation with Emergency Control Station (N2 Available). The nitrogen bottle placed in service first will have lower pressure than SOI criteria of 1200 psig, requiring the applicant to align another nitrogen bottle with sufficient pressure.

3.1 .k The applicant will place the cation demineralizer in service using SOI-62.04, CVCS Purification System, Section 8.2. Place Cation Bed in Service, due to reported high reactor coolant system activity.

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

1. 007 EA1.03 001 Given the following: /

- Unit 1 is operating at 15% power when an electrical grid disturbance results in a reactor trip because of a loss of offsite power.

- The operating crew completes performance of ES-0.1, Reactor Trip Response.

- Pressurizer Heater Banks C and D are energized.

When ES-0.1 is completed, which ONE of the following identifies how the RCS...

(1) temperature will stabilize and (2) pressure will be controlled?

Temperature Pressure A. Tavg at 557° F By auxiliary spray B. Tavg at 557° F By normal spray C Tcold at 557°F By auxiliary spray D. Tcold at 557°F By normal spray Page 1

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because RCS Tavg being maintained at 55 7°F would be correct if the reactor trip and the pressurizer pressure being controlled by the heaters and auxiliary spray is correct.

B. Incorrect, Plausible because RCS Tavg being maintained at 55 7°F and RCS pressure being controlled by the heaters and normal sprays would be correct if the RCPs had remained in service.

C. Correct, RCS Tcold will be maintained at 557°F because RCS will be in a Natural circulation condition due to the loss of offsite power. With no RCP running normal pressurizer spray is unavailable so the RCS pressure will be maintained by use of the pressurizer heaters and the auxiliary sprays. The steam dumps will be set to control SG pressure at 1092 psig (7107 psia) allowing Tcold to be maintained at 55 7°F.

D. Incorrect, Plausible because RCS Tcold being maintained at 55 7°F is correct and because the question requires knowledge of how RCPs are affected due to the loss of offsite power. If the RCPs in service the pressurizer pressure would be controlled by the heaters and normal sprays instead of the auxiliary spray being required.

Question Number: 1 Tier: 1 Group 1 K/A: 007 EA1.03 Reactor Trip Ability to operate and monitor the following as they apply to a reactor trip:

RCS pressure and temperature Importance Rating: 4.2 / 4.1 JO CFR Part 55: 41 .7 / 45.5 / 45.6 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to monitor and control RCS pressure and temperature following a reactor trip with reactor coolant pumps out of service.

Technical

Reference:

ES-0.1, Reactor Trip Response, Revision 0022 Proposed references None to be provided:

Page 2

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 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.

26. Given a set of plant conditions, use E-0, ES-0.0, ES-0.1, ES-0.2, ES-0.3, ESO.4 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 X Bank Question History: WBN bank question 007 EK3.01 001 modified from a question used on the 08/2010 NRC exam.

Comments:

Page 3

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

2. 008 AK2.03 002 Given the following:

- Unit 1 is operating at 100% power.

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

/-.-i-i Which ONE of the following describes thefAresponse of the pressurizer pressure control and level control systems?

Pressurizer Pressure Controller Pressurizer Level Controller 1 -PIC-68-340A 1 -LIC-68-339 A. output will INCREASE output will INCREASE B. output will INCREASE output will DECREASE C. output will DECREASE output will INCREASE D output will DECREASE output will DECREASE Page 4

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRA CTOR ANAL YSIS:

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

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

C. Incorrect, Plausible because the output of the master controller decreasing as the pressure drops below setpoint to turn on heaters is correct, but by design, with a 7/2 in vapor space leak in the PZR the heaters would not be able to maintain system pressure above the trip setpoint. Also, because a LOCA of this size, which is not in the vapor space, would result in the level controller output increasing to raise charging flow.

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

Page 5

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 2 Tier: 1 Group 1 K/A: 008 Pressurizer (PZR) Vapor Space Accident Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

Controllers and positioners Importance Rating: 2.5 I 2.4 IOCFRPart55: 41.7/45.7 JOCFR55.43.b: Not applicable KIA Match: This question matches the K/A by having the candidate determine how the pressuirzer master pressure controller and pressurizer level controller would respond to a vapor space LOCA.

Technical

Reference:

WOG E-1 background, Revision 2 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68C

8. Describe the operation of the master pressure controller.

15. Describe the response to a deviation from pressurizer level program.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 6

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

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

- A 200 gpm RCS leak is in progress.

- Containment pressure is stable at 3 psig.

Which ONE of the following completes the statements below?

Auxiliary Feedwater flow must be maintained greater than a specified setpoint until at least one S/G is above a minimum level of (1)

This level requirement ensures SIG ()

Li) t)

A. 29% NR tubes are covered in order to promote reflux cooling.

B. 29% NR inventory provides adequate secondary heat sink.

C. 39% NR inventory provides adequate secondary heat sink.

D. 39% NR tubes are covered in order to promote reflux cooling.

DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible because 29% NR is the correct minimum level for normal conditions in containment and reflux cooling is a mechanism for heat removal using the steam generators.

B. Incorrect, Plausible because 29% NR is the correct minimum level for normal conditions in containment and becuaseensuring adequate inventory to meet secondary heat sink requirements due to the steam generators being required during small break LOCA is correct.

C. Correct, the steam generators are required for heat removal during a small break LOCA and with containment pressure of 3.0 psig (higher than the adverse containment setpoint), the minimum water level in at least one steam generator is required to be at least 39% narrow range to ensure adequate inventory to meet secondary heat sink requirement.

D. Incorrect, Plausible because the correct minimum level for adverse containment is 39% NR, but the candidate incorrectly believes that reflux cooling is needed for heat removal.

Page 7

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 3 Tier: 1 Group 1 K/A: 009 EK2.03 Small Break LOCA Knowledge of the interrelations between the small break LOCA and the following: S/Gs.

Importance Rating: 3.0 / 3*3*

IOCFRPart55: 41.7/45.7 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the requirements for a secondary heat sink in the steam generators during a small break LOCA accident.

Technical

Reference:

ES-i .2, Post LOCA Cooldown and Depressurization, Revision 15 WOG E-1 Background HP Rev 4/30/2005 Proposed references None to be provided:

Learning Objective: 3-OT-EOP000 1

12. Discuss the purpose of ES-1.2 Post LOCA Cooldown and Depressurization.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN question EOPOO1.12 00i(with wording and formatting changes) Question was used on the 5/2008 exam Comments:

Page 8

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

4. 011 EK1.01 004 Which ONE of the following choices completes the statement below?

During a large break LOCA, Reflux boiling is A. significant for heat removal during a Cold Leg break, ONLY.

B. insignificant for heat removal during a Cold Leg break, ONLY.

C. significant for heat removal during both a Hot or Cold Leg break.

D insignificant for heat removal during either a Hot or Cold Leg break.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because there are LOCAs on the Cold leg where reflux boiling is required.

B. Incorrect, Plausible because there are LOCAs on the Hot leg where reflux boiling is required.

C. Incorrect, Plausible because there are LOCAs on both the Cold and Hot legs where reflux boiling is required.

D. Correct, During reflux boiling, water is heated and steam formed in the core, the water vapor/steam mixture rise in the hot leg and is cooled by the SG tubes. The cooling causes condensation and the water returns to the core via the hot leg.

During a large break LOCA the steam generator tubes would be emptied preventing the cooling but the RCS pressure would be low allowing sufficient ECCS flow to provide the required cooling without the steam generators.

Page 9

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 4 Tier: 1 Group 1 KIA: 011 EK1.01 Large Break LOCA Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA:

Natural circulation and cooling, including reflux boiling.

Importance Rating: 4.1 / 4.4 JOCFRPart55: 41.8/41.10/45.3 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question tests the ability of the applicant to recognize the operational implications of reflux boiling and relate the conditions that would require the present of refuel boiling as a required mechanism for core cooling during a LOCA. The KA asks for an operational implication of natural circulation and reflux boiling during a large break LOCA and the applicant is required to determine that there is no operational implication during this type of LOCA Technical

Reference:

WOG E-1 Background HP Rev. 2, 4/30/2005 Proposed references None to be provided:

Learning Objective: 3-OTEOP0 100,

12. Discuss the purpose of ES-1.2, Post LOCA Cooldown and Depressurization.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN Bank question 011 EK1 .01 003 modified Comments:

Page 10

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

5. 022 AA1.0$ 005 Given the following:

- Unit I is at 100% power at EOL.

- An automatic makeup to the VCT is in progress.

- Letdown flow is 75 gpm.

- 1-FC-3-139, BA TO BLENDER FCV-62-140 CONTROL, is in AUTO with a Pot setting of 4%.

- 1-HIC-3-142, PW TO BLENDER FCV-62-143 CONTROL, is in AUTO with a Pot setting of 35%.

- 1-LT-62-129A, VCT Level Transmitter, fails high.

In the absence of operator response, which ONE of the following describes the effect of the I -LT-62-J 29A failure on the actual VCT level and subsequent resu Its?

The VCT level will Av start dropping until the running CCP loses suction.

B. start dropping until the CCP suction transfers to the RWST.

C. rise until make-up automatically terminates, then slowly drop until automatic makeup is restarted; then cycle between the upper and lower setpoints.

D. rise until make-up automatically terminates, then start dropping and continue to drop until the CCP suction transfers to the RWST.

Page 71

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, the transmitter failing high will cause the VCT level to start dropping because the letdown will be diverted to the Holdup Tank. The VCT with automatic makeup in progress (with a water flow rate of 70 gpm and an acid flow rate of approximately 1.6 gpm) will result in the VCT level decreasing slowly until the VCT empties allowing the CCP to lose suction.

B. Incorrect, Plausible because the VCT level will start dropping and the level will drop to the suction transfer setpoint, but while the suction would normally swapover to the RWST, it cannot occur because the level transmitter is failed high preventing the transfer as identified in CAUTION I of ARI 709-A C. Incorrect, Plausible because prior to the transmitter failure, the VCT level would be rising and would continue to rise if the acid flow controller had been set higher (with a flow rate of greater than 5 gpm) (pot at approx 13% and higher). Then the makeup would terminate at 47%, and allow the level to start dropping until the makeup would auto start at the setpoint of 20%.

D. Incorrect, Plausible because prior to the transmitter failure, the VCT level would be rising and would continue to rise if the acid flow transmitter pot had been set higher (with a flow rate of greater than 5 gpm) (pot at approx 13% and higher). When the level reached 47%, the auto makeup would stop and, if the other level transmitter had been the one to fail, the auto makeup could not restart as the level dropped and the level would lower to the suction swapover setpoint.

VCT boric acid controller has a 0 to 40 gpm range as the controller pot is set from I to 100%.

With a setting of 4% the acid flow rate is approximately 7.6 gpm (40 x 0.04=7.6 gpm). This is confirmed by the values in Tl-59, Boron Tables and the placard mounted on 1-M-6.

Question Number: 5 Tier: 1 Group 1 K/A: 022 AA1.08 Loss of Reactor Coolant Makeup Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup:

VCT level Importance Rating: 3.4 / 3.3 10 CFR Part 55: 41.7 / 45.5 I 45.6 10CFR55.43b: Not applicable Page 12

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 K/A Match: K/A is matched because the reactor coolant makeup is maintained by maintaining suction to the CCPs from either the VCT or the RWST and without this process a loss of RCS makeup occurs. The question creates a condition where the VCT level will lost and the RWST supply will be blocked and the applicant must be able to determine this from the conditions (able to monitor VCT level as it applies to loss of reactor coolant makeup) given.

Technical

Reference:

ARI-1 09-115, CVCS & RHR RPS & ESF, Revision 18

-

Tl-59, Boron Tables, Revision 0007 SOl-62.02, Boron Concentration Control, Revision 0052 Placard on 1-M-6 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO62A

10. Explain the VCT level program.

12. Explain the automatic actuation logic and interlocks associated with the VCT outlet valves, FCV-62-J 32 and 133 and the CCP suction valves from the RWST, FCV-62-135 and 136.

28. Describe the modes of operation of the CVCS Boron Concentration and Reactor Makeup Control System.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Turkey Point bank question modified for use at WBN.

Comments:

Page 13

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

6. 025 AA2.02 006 Given the following:

- Unit 1 is at 220°F, with RHR Train A in service.

- The following annunciators are in alarm:

249-A - Ui SURGE TANK LEVEL HI/LO 179-A- CCS HXA i-RM-123 LIQ RAD HI

- CCS Surge Tank level is 86% and increasing.

- VCT level is decreasing at a rate of 3 gpm.

- VCT pressure is 18 psig.

Which ONE of the following identifies the heat exchangers that could be the source of the CCS in-leakage?

1. CVCS Letdown heat exchanger

2. RHR heat exchanger

3. Thermal barrier heat exchanger

4. Seal Water Return heat exchanger A. i,2,3and4 B 1,2, and 3 ONLY C. I and 2 ONLY D. 1 and 3 ONLY Page 14

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because each of the leaks would result in mixing of RCS water and CCS water but the seal water heat exchangers pressure would be lower than the CCS pressure. The seal water heat exchanger pressure would be at VCT pressure but could be at a higher pressure with other conditions (i.e. higher VCT pressure).

B. Correct, The RCS pressure, the RHR system pressure, and in the CVCS letdown heat exchanger pressure would be greater than CCS pressure and any of these could be the source of the CCS in-leakage.

C. Incorrect, Plausible because the CVCS letdown heat exchanger and the RHR heat exchanger could be at a higher pressure than CCS but they are not the only two in the list that would be higher.

D. Incorrect, Plausible because the CVCS letdown heat exchanger and the thermal barrier exchanger could be at a higher pressure than CCS but they are not the only two in the list that would be higher.

Question Number: 6 Tier: 1 Group 1 KIA: 025 AA2.02 Loss of Residual Heat Removal System (RHR)

Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

Leakage of reactor coolant from RHR into closed loop cooling water system or into reactor building atmosphere.

Importance Rating: 3.4 / 3.8 IOCFRPart55: 43.5/45.13 IOCFR55.43.b: Not applicable KIA Match: This question matches the K/A by having the candidate interpret the plant conditions and determine the source of a leak into the CCS system.

Technical

Reference:

AOl-i 5, Loss of Component Cooling Water, Revision 0032 ARI-241-253, CCS, Revision 0010 ARI-i 73-179, u-i Radiation Detectors Revision 0046 Page 15

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO7OA

7. Given a tube rupture in a CCS heat exchanger, describe the resulting flow path.

3-CT-AOl 1400

4. Describe 5 ways that RHR Cooling can be lost.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Kewaunee bank question used on Kewaunee 2004 NRC exam modified.

Comments:

Page 76

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 7.026 AK3.02 007 If a manual Safety Injection is initiated, which ONE of the following identifies a component or components that will have the cooIinc water isolated and the reason for the isolation?

Excess letdown heat exchanger as part of the required containment isolation.

B. Excess letdown heat exchanger to ensure adequate cooling flow to SI required components.

C. Containment Lower Compartment Coolers as part of the required containment isolation.

D. Containment Lower Compartment Coolers to ensure adequate cooling flow to SI required components.

DISTRACTOR ANALYSIS:

A. Correct, the Excess Letdown Heat Exchanger CCS supply and return valves receive an isolation signal from the Containment Phase A isolation signal which is generated on any Safety Injection signal and the reason is to ensure containment is isolated.

B. Incorrect, the Excess Letdown Heat Exchanger CCS supply and return valves receive an isolation signal from the Containment Phase A isolation signal which is generated on any Safety Injection signal but the isolation is not to ensure adequate cooling flow to SI required components. Plausible because the isolation occurring is correct and during a safety injection, cooling for the ESF loads is required.

C. Incorrect, the Containment Lower Compartment Cooler supply and return valves receive an isolation signal only if Containment Phase B isolation occurs not just a Safety Injection signal. If the isolation did occur it would be as a part of a containment isolation. Plausible because the isolation does occur during a Phase B containment isolation and the reason being to ensure containment is isolated is correct.

D. Incorrect, the Containment Lower Compartment Cooler supply and return valves receive an isolation signal only if Containment Phase B isolation occurs not just a Safety Injection signal; thus, the isolation is not to ensure adequate cooling flow to SI required components. Plausible because the isolation does occur during a Phase B containment isolation and following a safety injection, cooling for the ESF loads is required.

Question Number: 7 Page 77

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Tier: 1 Group 1 KIA: 026 AK3.02 Loss of Component Cooling Water (CCW)

Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water:

The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS Importance Rating: 3.6 / 3.9 JO CFR Part 55: 41.5,41.10/45.6/45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question knowledge of and the reason for the automatic alignments of the Component Cooling Water system that occur during a safety injection (ESFAS) actuation.

Technical

Reference:

1-47W611-63-1 R13 1-47W611-88-1 R23 N3-70-4002, Component Cooling System, Revision 15 WB-DC-40-34, Containment Isolation System, Revision 0009 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO88A

6. Describe the conditions which would cause each of the following to occur:

a. Containment Vent Isolation.

b. Phase A Containment Isolation.

c. Phase B Containment Isolation.

9. Given a set of plant conditions, determine the correct response of the Containment Isolation System.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question 026 AK3.02 006 Comments:

Page 18

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Page 19

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 8.027 AG2.1.27 008 Given the following:

- Unit 1 is operating at 100% power with all controls in automatic.

- 1-XS-68-340D, PZR PRESS CONTROL CHANNEL SELECT, is selected to the P1-68-340 & 334 position.

- A step-load reduction to 50% of full power occurs due to a turbine control malfunction.

- During the load reduction 1-PT-68-334, Pressurizer Pressure Transmitter, fails LOW.

Which ONE of the choices below completes the following two statements relative the the transient?

In accordance with N3-68-4001, Reactor Coolant System, the pressurizer pressure control equipment is designed to respond to the given conditions without (1)

When 1-PT-68-334 failed, the pressurizer spray valves would (2)

Li)

A. a reactor trip go closed.

B a reactor trip continue to respond to pressure changes.

C. opening a PORV go closed.

D. opening a PORV continue to respond to pressure changes.

Page 20

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the pressurizer pressure control system is designed to provide satisfactory operation without a reactor trip and the valves would have gone closed if the failure had been on 7-PT 68-323 instead of 7-PT 68-334.

B. Correct, The pressurizer pressure control system is designed to provide satisfactory operation without a reactor trip during a step-load reduction of 50%

of full power with automatic reactor control (7 0%) and steam dump (40%). (See below). Also, the failure of 7-PT-68-334 with the identified selector switch operation would not affect the operation of the pressurizer spray valves. The valves are being controlled by 7-PT-68-323 would continue to operate as required.

C. Incorrect, Plausible because opening the pressurizer PORV is not a usual occurrence of the control system in response to an expected transient and the valves would have gone closed it the failure had been on 1-PT-68-323 instead of 7-PT 68-334.

D. Incorrect, Plausible because opening the pressurizer PORV is not a usual occurrence of the control system in response to an expected transient and the pressurizer spray valves continuing to operate as required is correct.

3.2.4 Pressurizer (PZR)

The PZR provides a point in the RCS where liquid and vapor can be maintained in equilibrium under saturated conditions for RCS pressure control. The PZR and its sprays, heaters, PORVs, safety valves, and surge line make up the pressure control equipment. This equipment is designed to accommodate changes in system volume and to limit changes in system pressure due to RCS loop temperature variation during all modes of operation. In particular, the equipment is designed to provide satisfactory operation without reactor trip when the RCS is subjected to the following transients:

a. Loading or unloading at 5% of full power per minute, with automatic reactor control.

b. Instantaneous load transients of >10% of full power (not exceeding full power), with automatic reactor control.

c. Step-load reduction of 50% of full power with automatic reactor control (10%) and steam dump (40%).

Page 27

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 8 Tier: 1 Group 1 K/A: 027 AG2.1.27 Pressurizer Pressure Control (PZR PCS) Malfunction Knowledge of system purpose and/or function.

Importance Rating: 3.9 / 4.0 10 CFR Part 55: 41.7 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the Pressurizer Pressure Control system function and how a detector malfunction will affect system operation.

Technical

Reference:

AOl-i 8, Malfunction of Pressurizer Pressure Control System, Revision 0022 N3-68-4001, Reactor Coolant System, Revision 0030 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68C

8. Describe the operation of the master pressure controller.

22. Explain the operation of major system components.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 22

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

9. 029 EG2.4.21 009 The plant has experienced an ATWS and boration is in progress per FR-S.1, Nuclear Power Generation! ATWS.

Which ONE of the following identifies the two criteria that must be satisfied to verify the reactor is subcritical per FR-S. 1?

A. All reactor trip breakers OPEN and All rod bottom lights ON.

B. All rod bottom lights ON and Intermediate Range SUR negative.

C. Source Range detectors energized and Source Range SUR negative.

Dv Power Range detectors less than 5% and Intermediate Range SUR negative.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible if applicant thinks that the reactor trip breakers must be open and all rod bottom lights on before the reactor can be declared subcritical, however that is not one of criteria per FR-S. I used to verify that the reactor is subcritical.

B. Incorrect, Plausible if applicant thinks that you must have the rod bottom lights lit before the reactor can be determined to be subcritical, however that is not one of criteria per FR-S. I used to verify that the reactor is subcritical. Also plausible because the intermediate range SUR being negative is one of the required criteria.

C. Incorrect, Plausible if applicant thinks that you must have the source range energized and indicating a negative SUR before the reactor can be determined to be subcritical, however these are not the criteria per FR-S. I used to verify that the reactor is subcritical.

D. Correct, Per step 20 of FR-S. I, the procedure has the crew Check the reactor subcritical: by verifying that Power Range channels less than 5% and Intermediate range startup rate negative.

Page 23

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 9 Tier: 1 Group 1 K/A: 029 EG 2.4.21 Anticipated Transient Without Scram (ATWS)

Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Importance Rating: 4.0 I 4.6 JOCFRPart55: 41.7/43.5/45.12 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the applicant to identify the parameters and logic used to assess the reactivity control safety function during an ATWS condition.

Technical

Reference:

FR-S.1, Nuclear Power Generation/ATWS, Revision 0020 Proposed references None to be provided:

Learning Objective: 3-OT-FRS0001

9. Given a set of plant conditions, use FR-S.1, FR-S.2 and the Critical Safety Function Status Trees to correctly diagnose and implement: Action Steps, RNOs, Notes and Cautions.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: SQN bank question FR-S-1-B.8.A 001 Comments:

Page 24

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

10. 038 EA1.37 010 Given the following:

- A SGTR occurs on Unit 1.

- Following the termination of Safety Injection in E-3, Steam Generator Tube Rupture, a loss of off-site power occurs.

- The crew is ready to maintain the RCS pressure at less than the ruptured steam generator pressure in accordance with E-3.

- Conditions require auxiliary spray to be used.

- 1-PIC-68-340B, LOOP 2 SPRAY CONTROL, is in MANUAL and at 50%.

Which ONE of the following completes the statement below?

To prevent exceeding the cyclic limit, the maximum allowed z\T between the pressurizer and the charging flow is (1) and if 1-PIC-68-340B is adjusted to 0% it will (2) the potential for thermal shock of the presurizer spray nozzle when the auxiliary spray is opened.

c)

A. 100°F increase B. 100°F decrease C 320°F increase D. 320°F decrease Page 25

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because if zlT between Pzr-Spray is greater than 100°F, Spray use is restricted under the conditions in 1-SI-0-8 (See below) and placing the controller output to 0% is correct.

B. Incorrect, Plausible because iMT between Pzr-Spray is greater than 100°F, Spray use is restricted under the conditions in 1-51-0-8 (See below) and there are several valves that are closed when the respective controller output is at 100%. Distractor requires knowledge of how the valves must be positioned and how the controller acts to position the valve as well as knowing that the valves being opened would result in the spray flow entering the pressurizer being reduced.

C. Correct, The normal maximum AT between the pressurizer and the spray is 320°F and Appendix A for aligning does contain the note stating Aux spray flow can be maximized by closing the normal pressurizer spray valve(s). The valves are closed when the respective controller output is at 0%. So, by positioning the normal spray valve, the potential for thermal shock would increase.

D. Incorrect, Plausible because the normal maximum AT between the pressurizer and the spray being 320°F is correct and there are several valves that are closed when the respective controller output is at 100%. Distractor requires knowledge of how the valves must be positioned and how the controller acts to position the valve as well as knowing that the valves being opened would result in the spray flow entering the pressurizer being reduced.

SOI-68.03 3.0 PRECAUTIONS AND LIMITATIONS A. If AT between Pzr-Spray is greater than 100°F, Spray use is restricted under the conditions in 1-SI-0-8 (To avoid cyclic stress, spray flow is NOT normally initiated if Pzr-Spray AT is greater than 100°F). Maximum Pzr-Spray AT is 320°F.

Page 26

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 10 Tier: 1 Group 1 KIA: 038 EA1.37 Steam Generator Tube Rupture (SGTR)

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

Controlling of thermal shock during PZR spray operation Importance Rating: 3*5* / 34 JO 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 applicant to operate and monitor the operation of PZR spray with conditions that have the potential for thermal shock of the pressurizer spray nozzle.

Technical

Reference:

E-3, Steam Generator Tube Rupture, Revision 23 SOl-68.03, Pressurizer Pressure and Spray Control System, Revision 0021 1-SI-0-8, Monitoring Component Cyclic or Transient Limits, Revision 0007 Proposed references None to be provided:

Learning Objective: 3-OT-EOPO300

5. Given a set of plant conditions, use E-3, ES-3.1, ES-3.2, and ES-3.3 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 for the WBN 06/2011 NRC exam.

Comments:

Page 27

0612011 Watts Bar SRO NRC License Exam 6/22/2011 11.054 AA2.O1 011 Given the following:

- Unit 1 is operating at 60% power.

- An AFWline break in the South Valve Vault causes a Unit 1 trip.

Which ONE of the following identifies...

(1) the sequence of actions that occur when the unit trips, and (2) if resetting the Feed Water Isolation would require cycling the reactor trip breakers?

A. (1) The reactor will trip causing the turbine to trip.

(2) Would require cycling.

B. (1) The reactor will trip causing the turbine to trip.

(2) Would NOT require cycling.

C (1) The turbine will trip causing the reactor to trip.

(2) Would require cycling.

D. (1) The turbine will trip causing the reactor to trip.

(2) Would NOT require cycling.

Page 28

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because the feedwater isolation signal will cause SG levels to start dropping and if the conditions that cause the automatic turbine trip are not recalled correctly, the reactor would trip on SG low level causing the turbine to trip and the reactor trip breakers being required to be cycled to clear the isolation signal is correct.

B. Incorrect, Plausible because the feedwater isolation signal will cause SG levels to start dropping and if the conditions that cause the automatic turbine trip are not recalled correctly, the reactor would trip on SG low level causing the turbine to trip and there are EWI signals that do not require the reactor trip breakers to be cycled to clear the isolation signal (e.g. EWI due to reactor trip coincident with Lo Tavg).

C. Correct, The FWI signal causes both the MEW pumps to trip, Both MFPs tripping causes the turbine to trip which then causes a reactor trip. The reactor trip breakers do have to be cycled to clear the FWI signal generated as a result of the valve room flood switch operation.

D. Incorrect, Plausible because the turbine would trip causing the reactor to trip and there are EWI signals that do not require the reactor trip breakers to be cycled to clear the isolation signal (e.g. FWI due to reactor trip coincident with Lo Tavg).

Page 29

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 11 Tier: 1 Group 1 KIA: 054 AA2.01 Loss of Main Feedwater (MFW)

Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW):

Occurrence of reactor and/or turbine trip Importance Rating: 4.3 I 4.4 IOCFRPart55: 43.5/45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to determine how the loss of main feed water affects the occurrence of the reactor and turbine trip.

Technical

Reference:

1 -47W61 1-3-2 R22 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO3A

12. List the equipment affected by a Feedwater Isolation Signal.

13. Describe the three steps to reset FW Isolation Signal.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank questions SYS 003A.13 002 and SYSOO3A.11 010 modified Comments:

Page 30

06/2011 Waifs Bar SRD NRC License Exam 6/22/2011

12. 055 EK1.02 012 Given the following plant conditions:

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

- During the rapid depressurization of all intact SGs to reduce RCS pressure to 300 psig, an overshoot occurs.

- RCS is reduced to 195 psig before the depressurization is stabilized.

What is the potential implication that could result from this overshoot in SG depressurization?

A. The RCP seals may be damaged.

B Natural circulation may be impeded.

C. Unacceptable upper head voiding may occur.

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

DIS TRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the RCP seals are a concern during the loss of aIIAC power and is the reason the steam generators are depressurized (to reduce RCS pressure) to help protect the seals. But this is not the implication of overshooting the depressurization.

B. Correct, Nitrogen gas accumulation being discharged from the CLAs will inhibit natural circulation C. Incorrect, Plausible because upper head voiding will occur after pressurizer level is lost. This is an acceptable consequence to minimize RCS inventory loss through RCP seal degradation.

D. Incorrect, Plausible because the rate of cooldown may be a concern; but the final pressure/temperature is evaluated.

Page 31

06/2011 Watts Bar SRD NRC License Exam 6/22/2011 Question Number: 12 Tier: 1 Group 1 KIA: 055 EK1.02 Loss of Offsite and Onsite Power Knowledge of the operational implications of the following concepts as they apply to the Station Blackout:

Natural circulation cooling Importance Rating: 4.1 / 4.4 IOCFRPart55: 41.8/41.10/45.3 IOCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of conditions that could impact (operational implications) natural circulation cooling during a Station Blackout.

Technical

Reference:

ECA-0.0, Loss of Shutdown Power, Revision 21 Proposed references None to be provided:

Learning Objective: 3-OT-ECA0000

06. Explain why the operator is directed to maintain RCS press greater than 250 psig during RCS cooldown (ECA-0.0).

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: SQN bank question 055 EK1.02 012 with stem changes to make applicable for WBN (ECA title and the pressure range)

Comments:

Page 32

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

13. 056 AG2.4.47 013 Given the following time line:

0800 - Unit I Condensate Storage Tank (CST) is determined to be at the Tech Spec minimum level.

0800 - The WBN Hydro switchyard is completely de-energized due to an electrical storm which also causes a Unit 1 reactor trip.

1200 - The RCS is at normal No Load temperature and pressure when the decision is made to cool the RCS down to Mode 5.

1300 - RCSTavgis532°F.

1400 - RCSTavgis5O7°F.

1500 - RCSTavgis482°F.

1600 - RCS Tavg is 457°F.

Which ONE of the following identifies...

(1) the volume of water in the Unit 1 CST at 0800 and (2) if the CST will have sufficient water to allow the cooldown to RHR cut-in to be reached without additional makeup to the tank if the current cooldown rate is maintained?

w A. 210,000 gal. The CST is designed to have sufficient water to reach RHR at the current cooldown rate.

B. 210,000 gal. Additional makeup will be required to reach RHR at the current cooldown rate.

C. 200,000 gal. The CST is designed to have sufficient water to teach RHR at the current cooldown rate.

D 200,000 gal. Additional makeup will be required to reach RHR at the current cooldown rate.

Page 33

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because 210,000 gallons level is setpoint for the CST low level alarm and, if the times or the cooldown rate had been different, the required volume would be present.

B. Incorrect, Plausible because 210,000 gallons level is setpoint for the CST low level alarm and additional water being required to reach RHR at the current cooldown rate is correct.

C. Incorrect, Plausible because the minimum level is correct and, if the holding time had been shorter or the cooldown rate had been higher, the required volume would be present.

D. Correct, The minimum level is 200,000 gallons and the design basis of the minimum level is to allow maintaining Hot Standby for 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> followed by a 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> cooldown at 50°F/hour to RHR cut in conditions. The plant has already been at hot standby conditions for 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> and the cooldown rate will not provide cool down within the required time frame.

Page 34

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 13 Tier: 1 Group 1 KIA: 056 AG2.4.47 Loss of Offsite Power Emergency Procedures / Plan Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Importance Rating: 4.2 / 4.2 10 CFR Part 55: 41.10 / 43.5 / 45.12 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability recognize trends during a loss of offsite power event in an accurate and timely manner using information provided on MCR instrumentation.

Technical

Reference:

ARI-36-42, Heaters, Tutb Seal & Air, Rev. 0018 N3-3B-4002, Auxiliary Feedwatet System, Revision 0015 N3-2-4002, Condensate System, Revision 0015 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO2A

3. State the purpose and the capacity of the condensate storage tank

33. Given the status of the Condensate Storage Tank and the appropriate Technical Specification, determine if operability requirements are met and what actions, if any, are required Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSOO2A.35 001 modified for WBN 06/2011 NRC exam Page 35

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

14. 057 AK3.01 014 Given the following plant conditions:

- Unit 1 isat 100% power.

- Alarms received indicate that an electrical board has failed.

- All trip status lights are OFF on panel 1-XX-55-5 on 1-M-5.

Which ONE of the following identifies...

(1) which electrical board failed and (2) the reason that manipulation of controls in the Auxiliary Control Room (ACR) is required?

A. 120 VAC Vital Instrument ACR Auxiliary Feedwater Controllers for Power Board 1-I S/G 3 and 4 have swapped to MANUAL and require adjustment to ensure an operable heat sink is maintained.

B. 120 VAC Vital Instrument ACR Auxiliary Feedwater Controllers for Power Board 1-Il S/G 1 and 2 have swapped to MANUAL and require adjustment to ensure an operable heat sink is maintained.

C. 120 VAC Vital Instrument CHARGING FLOW controllers, Power Board 1-I 1-HIC-62-93A and 1-HIC-62-89A cannot be controlled from MCR requiring control to be established in the AUX position.

D. 120 VAC Vital Instrument CHARGING FLOW controllers, Power Board 1-Il 1-HIC-62-93A and 1-HIC-62-89A cannot be controlled from MCR requiring control to be established in the AUX position.

Page 36

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

A. Incorrect, The board failure is correct, and the ACR AFW controllers do swap to MANUAL, but the SG AEW controllers are incorrect (Board 1-I would shift controllers for S/G I and 2).

B. Incorrect, The board failure is incorrect, and the ACR AFW controllers do swap to MANUAL, but the SG AFW controllers are incorrect (Board 1-Il would shift controllers for S/G 3 and 4).

C. Correct, A 01-25.07 states that the listed flow control valves controllers are failed, and Appendix C, Alternate Control of Letdown and Charging specifically addresses placing the transfer switches for 1-PC V-62-93 and 1-PC V-62-89 in the AUX position on Panel 1-L-1 18 and 1-L-1 IA respectively.

D. Incorrect, The board failure is incorrect, but the reason for the controllers operation is correct.

Question Number: 14 Tier: 1 Group 1 KIA: 057 AK3.01 Loss of Vital AC Electrical Instrument Bus Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus:

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

Importance Rating: 4.1 / 4.4 10 CFR Part 55: 41.5/41.10/45.6/45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of an action contained in the AOl for a loss of an Vital AC Electrical Instrument Bus.

Technical

Reference:

AOI-25.01, Loss of 1 20V AC Vital Instrument Power Boards 1-I and 2-I, Revision 31, &

Appendices A & C.

Proposed references None to be provided:

Learning Objective: 3-OT-A012500

1. Demonstrate ability to recognize a loss of any 120V Page 37

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Learning Objective: 3-CT-AC 12500

1. Demonstrate ability to recognize a loss of any 1 20V AC Vital Power Bd, including effects on equipment and controls (SOER 81-02).

2. Analyze Symptoms for loss of Vital Power Bd 1-I, and evaluate their importance to system operation per AOl-25 series (SOER 81- 02).

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN bank question 057 AK3.01 011 used on 5/2008 exam with minor format changes.

Comments:

Page 38

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

15. 058 AA2.03 015 Given the following:

- Unit 1 is at 100% power when the following annunciator windows alarm.

196-A - DG CONTROL POWER FAILURE 197-B - DG BATTERY ABNORMAL

- An AUD sent to check, reports that the following 2 breakers OPEN in the DG IA-A 125V DC Battery Distribution Panel:

1-BKR-82-A, DG IA-A 125V DC SUPPLY FROM BAIT 1A 1-BKR-82-A/6, DG lA-A 125V DC SUPPLY FROM BATT CHGR IA

- The condition has NOT been corrected.

Which ONE of the following identifies how D/G lA-A will respond to a loss of voltage on 6.9kV Shutdown Board lA-A?

A. The DG will automatically start and connect to the shutdown board, but control from 0-M-26 will NOT be available.

B. The DG will automatically start but the generator field will NOT be flashed as the DG speed increases.

C. The DG will NOT automatically start but could be started manually from its Local Control Panel.

D The DG will NOT automatically start and could NOT be started from its Local Control Panel.

Page 39

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because a loss of 125v DC vital (different DC supply) will cause the DG to start and because the loss of the controls on Q-M-26 is correct.

B. Incorrect, Plausible because a loss of 125v DC vital (different DC supply) would cause the DG to start and the loss of the DG 725v DC will result in no DC to flash the generator field.

C. Incorrect, Plausible because the DG not starting automatically is correct and other conditions would allow the DG to be started from the local control panel.

D. Correct, With the battery and the charger separated from the distribution panel, there will be not control voltage on the local control panel nor any potential for the auxiliary relays (including the power to the air start relays and solenoids) required to start the DG. Thus, the DG will not start automatically and cannot be started from its local control panel.

Page 40

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 15 Tier: 1 Group 1 K/A: 058 AA2.03 058 Loss of DC Power Ability to determine and interpret the following as they apply to the Loss of DC Power:

DC loads lost; impact on ability to operate and monitor plant Importance Rating: 3.5! 3.9 10 CFR Part 55: 43.5/45.13 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the ability to determine what DC loads are lost from plant conditions and how the loads lost impact the ability to monitor and operate the plant.

Technical

Reference:

ARI-195-201, Diesel GEN lA-A, Revision 13 SOl-82.01, Diesel Generator(DG) lA-A, Revision 0074 1 -45W760-82-2 Ri 7 1-45W760-82-3 RiO 1-45W760-82-4 R20 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO82A

10. Given a set of plant conditions, determine the correct response of the Diesel Generator or Diesel Generator Support Systems.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question 058 AK3.01 012 modified.

Comments:

Page 47

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

16. W/E04 EK2.2 016 Given the following:

- A LOCA outside Containment has occurred.

- The crew is performing ECA-1 .2, LOCA Outside Containment.

Which ONE of the following is:

(1) the first system to be addressed while performing the steps to identify and isolate the leak, and (2) the parameter directed to be monitored to determine if the leak has been isolated?

(1) (2)

First System Isolated RCS Parameter Monitored A. Residual Heat Removal subcooling B Residual Heat Removal pressure C. Safety Injection subcooling D. Safety Injection pressure Page 42

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the RHR system is the heat removal system that is addressed first when performing ECA-1.2, LOCA Outside Containment and RCS subcooling would be rising following isolation of the leak.

B. Correct, Because the leak is most likely to be on the low pressure piping of the RHR system, it is the heat removal system that is addressed first when performing ECA-1.2, LOCA Outside Containment and the parameter checked to determine if the leak has been isolated following an attempt to isolate is the RCS pressure.

C. Incorrect, Plausible because the Safety Injection system is the heat removal system that is performing the cooling function for the core when performing ECA-1.2, LOCA Outside Containment and RCS subcooling would be rising following isolation of the leak.

D. Incorrect, Plausible because the Safety Injection system is the heat removal system that is performing the cooling function for the core when performing ECA-1.2, LOCA Outside Containment and RCS pressure rising following isolation of the leak is correct.

Question Number: 16 Tier: 1 Group 1 KIA: W/E04 EK2.2 LOCA Outside Containment Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

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

Importance Rating: 3.8 / 4.0 IOCFRPart55: 41.7/45.7 JOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the proper operation the plants heat removal systems (as to the order of priority to be addressed) during a LOCA outside containment.

Page 43

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

ECA-1 .2, LOCA Outside Containment, Revision 0005 Proposed references None to be provided:

Learning Objective: 3-OT-ECAO 101

01. Identify and explain the major actions of procedures ECA-1.1 and 1.2.

08. Given a set of plant conditions, use procedures ECA-1 .1 and 1.2 to identify any required procedure transition.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Vogtle bank question WEO4EK2.2 used on the Vogtle 2010 exam with 2 distractors changed and wording changed in the stem to allow use at WBN.

Comments:

Page 44

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

17. W/E05 EK1.3 017 Given the following:

- RCS Bleed and Feed has been initiated on Unit 1 in accordance with FR-H.1, Loss of Secondary Heat Sink.

- All SG WR levels are 21% and lowering.

- Phase B windows are LIT on both MASTER ISOL SIGNAL STATUS PNLs on 1-M-6.

- RCS Loop WR temperature is 551°F and rising.

- Core Exit TC temperatures are 554°F and rising.

- Conditions have been established to allow feed flow using the condensate system.

- 2 Hotwell pumps and 2 Condensate Booster pumps have been started.

Which ONE of the following identifies the strategy the crew should use to initially re-establish secondary flow?

Establish feed flow to A. all SGs and feed at <100 gpm/SG until the respective SC WR level is >25%.

B. a selected SG and feed at <100 gpm until WR level is > 25%.

C. all SGs and feed at the maximum rate.

D a selected SG and feed at the maximum rate.

Page 45

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because feeding all steam generators could have been correct prior to the initiation of bleed and feed After bleed and feed initiated, feeding at less than 100 gpm until NR level is> 25% would have been correct if the Core Exit TCs had not been rising.

B. Incorrect, Plausible because feeding a selected steam generator is correct and feeding at less than 100 gpm until NR level is > 25% would have been correct if the Core Exit TCs had not been rising.

C. Incorrect, Plausible because feeding all steam generators could have been correct prior to the initiation of bleed and feed and after bleed and feed initiated, feeding with no restrictions is correct.

D. Correct, the conditions direct that a selected steam generator will be fed and that the rate of feed will be at the maximum rate.

Question Number: 17 Tier: 1 Group 1 KIA: W/EO5EK1.3 Loss of Secondary Heat Sink Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink)

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

Importance Rating: 3.9 / 4.1 JUCFRPart55: 41.8/41.10/45.3 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires of the operational implications of the Phase B windows being LIT on the Master Isolation Signal Panels and along with the other conditions in the question to determine how many steam generators will be fed and at what rate of feed flow.

Technical

Reference:

FR-H. 1, Loss of secondary Heat Sink, Revision 18 Proposed references None to be provided:

Learning Objective: 3-OT-FRH0001 Page 46

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 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: Bank question WEO5EK1.3 2 from Vogtle 2009 exam (28) with stem conditions and wording changes for use atWBN.

Comments:

Page 47

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

18. W!E12 1K3.1 01$

Given the following:

- Following a steam line break outside containment on Unit 1, the crew is performing ECA-2.1, Uncontrolled Depressurization of All Steam Generators.

- The RCS temperature is at T-sat for the S/G pressure.

- The RCS pressure is 1490 psig and slowly dropping.

Which ONE of the following completes the statements below during the performance on ECA-2.1?

An overall goal of the action to Control Feed Flow is intended to (1)

The Reactor Coolant Pumps (2)

A (1) minimize further cooldown of the Reactor Coolant System.

(2) should remain in service because the pressure drop is caused by the S/G depressurization.

B. (1) minimize further cooldown of the Reactor Coolant System.

(2) are required to be stopped to prevent core uncovery if the pumps were to trip later in the transient.

C. (1) maintain minimum FR-0, Status Trees, requirement for heat sink.

(2) should remain in service because the pressure drop is caused by the S/G depressurization.

D. (1) maintain minimum FR-0,Status Trees, requirement for heat sink.

(2) are required to be stopped to prevent core uncovery if the pumps were to trip later in the transient.

Page 48

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, Feedwater flow is controlled to minimize the cooldown of the RCS (even if minimum heat sink is not maintained) and in accordance with the Caution in ECA 2.7, the RCPs will remain in service if the pressure drop is caused by S/G depressurization as indicated by RCS temperature at T-sat for S/G pressure.

B. Incorrect, Plausible because feedwater flow being minimized to control the cooldown of the RCS is correct and the RCPs are removed if the RCS drops to less than 7500 psig in other procedures due to the potential for uncovering the core in the event the pumps trip later in the accident.

C. Incorrect, Plausible because feedwater flow being controlled to maintain minimum heat sink would be correct during performance of other instructions and the RCPs remaining in service is correct.

D. Incorrect, Plausible because feedwater flow being controlled to maintain minimum heat sink would be correct during performance of other instructions and the RCPs are removed if the RCS drops to less than 1500 psig in other procedures due to the potential for uncovering the core in the event the pumps trip later in the accident.

Question Number: 18 Tier: 1 Group 1 K/A: W/E12EK3.1 Uncontrolled Depressurization of all Steam Generators Knowledge of the reasons for the following responses as they apply to the (Uncontrolled Depressurization of all Steam Generators.)

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: 3.5 / 3.9 10 CFR Part 55: 41.5 /41.10, 45.6, 45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the reasons as they apply to an Uncontrolled Depressurization of all Steam Generators for controlling feedwater flow and for not stopping the RCPs when condition of the RCS pressure would result in tripping the pumps under most conditions while in the EOP network.

Page 49

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

ECA-2.1, Uncontrolled Depressurization of All Steam Generators, Revision 12 ECA-2.1, Uncontrolled Depressurization of All Steam Generators, WOG Background H P-Rev.2, April 30, 2005 Proposed references None to be provided:

Learning Objective: 3-OT-ECAO2O1

1. Describe the major actions of ECA-2.1, Uncontrolled Depressurization of all Steam Generators.

3. Discuss the reasons for maintaining a minimum flow to the S/Gs during the uncontrolled depressurization of all SGs.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question ECAO2O1 002 modified for use on the WBN 06/20 1 1 exam.

Comments:

Page 50

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

19. 005 AA1.01 019 Given the following:

- Unit 1 is operating at 70% power.

- D Bank Group 2 Control Rod H8 is 14 steps lower than the rest of D bank.

- The rod has been misaligned for 90 minutes.

- The crew has been directed to realign the rod in accordance with AOl-2, Malfunction of Reactor Control System.

- Annunciator 86-A, CONTROL ROD URGENT FAILURE, alarms when the rod realignment begins.

Which ONE of the following identifies which D bank lift coil disconnect switch(es) was/were opened in accordance with AOI-2 and the source of the 86-A alarm?

A Alt except Rod H-8 were disconnected.

Group 1 is the source of the 86-A alarm.

B. All except Rod H-8 were disconnected.

Group 2 is the source of the 86-A alarm.

C. Only Rod H-8 was disconnected.

Group I is the source of the 86-A alarm.

D. Only Rod H-8 was disconnected.

Group 2 is the source of the 86-A alarm.

Page 51

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, In accordance with A 01-2, since the rod has been misaligned for greater than one hour, the rod is moved to be in alignment with the bank and because Group I is the source of the alarm.

B. Incorrect, Plausible because H-a is the only disconnect switch that will not be disconnected and because Group 2 is in a condition where all rods in the group will supply the same feedback when movement is initiated.

C. Incorrect, Plausible because all D Bank lift coil disconnect switches would have been disconnected except H-8 if the alignment had been performed within one hour and because Group us the source of the alarm.

D. Incorrect, Plausible because all D Bank lift coil disconnect switches would have been disconnected except H-8 if the alignment had been performed within 60 minutes and because Group 2 is in a condition where all rods in the group will supply the same feedback when movement is initiated.

Page 52

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 19 Tier: 1 Group 2 K/A: OO5AAI.01 Inoperable/Stuck Control Rod Ability to operate and/or monitor the CRDS as it applies to the inoperable /

stuck control rod.

Importance Rating: 3.6 / 3.4 10 CFR Part 55: 41.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the applicant to properly align the control rod disconnect switches when re-aligning an inoperable rod and understand why an annunciator will be received as the alignment is initiated.

Technical

Reference:

ARI-81-87, NIS & Rod Controls, Revision 0033 AOl-2, Malfunction of Reactor Control System, Rev. 37 Proposed references None to be provided:

Learning Objective: 3-OT-A010200

08. Explain Operator Actions for an RCCA Misalignment.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Turkey Point bank question 005AA1 .01 used on 2009 exam with changes to make question apply to WBN.

Comments:

Page 53

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

20. 028 AG2.2.44 020 Given the following:

- Unit 1 is operating at 30% reactor power with Tavg and Tref matched.

- Rod Control is in MANUAL for system trouble shooting.

- The Tavg Auctioneering unit fails to a value of 571 °F.

Which ONE of the following completes the statement below?

To restore pressurizer level to the program level for 30% power, the DAC will throttle 1-HIC-62-93A, CHARGING FLOW PZR LEVEL CONTROL in the (1) direction and the action to throttle the valve will cause RCP seal flow to (2)

A. OPEN rise B. OPEN drop C. CLOSE rise D CLOSE drop Page 54

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because the change in programmed level compared to required level could be calculated wrong (or the 2 series valves in the charging line could be reversed) and if the manual action had been to throttle 1-HIC-62-89, CHRG HDR RCP SEALS FLOW CONTROL, the RCP seal injection flow would have increased.

1-HIC-62-93A and 1-HIC-62-89A are both used to restore control of the pressurizer level.

B. Incorrect, Plausible because the change in programmed level compared to required level could be calculated wrong (or the 2 series valves in the charging line could be reversed) and if the manual action had been to throttle 7-HIC-62-89, CHRG HDR RCP SEALS FLOW CONTROL, the RCP seal injection flow would have increased.

1-HIC-62-93A and 1-HIC-62-89A are both used to restore control of the pressurizer level. Also because the RCP seal injection flow effect is correct.

C. Incorrect, Plausible because the OAC throttling the level control valve closed is correct and if the manual action had been to throttle 1-HIC-62-89, CHRG HDR RCP SEALS FLOW CONTROL, the RCP seal injection flow would have increased.

1-HIC-62-93A and 1-HIC-62-89A are both used to restore control of the pressurizer level.

D. Correct, With reactor power at 30% steady state power, Tavg is 565.8°F and pressurizer level setpoint is 35.5%. When the Tavg auctioneering unit fails to an output of 571°F, the pressurizer level setpoint changes to 47.8%. The pressurizer level control system will throttle the level control valve open to raise pressurizer level to the increased program setpoint. To restore the correct level for 30% power the OAC will place 1-HIC 93A in manual and throttle the valve closed. This action will result in a drop in the RCP seal injection flows because the seal injection line is downstream of the level control valve.

Question Number: 20 Tier: 1 Group 2 KIA: 028 AG2.2.44 Pressurizer (PZR) Level Control Malfunction Equipment Control Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Importance Rating: 4.2 / 4.4 IOCFRPart55: 41.5/43.5/45.12 IOCFR55.43.b: Not applicable Page 55

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: K/A is matched because the question requires the applicant to interpret control room indications (pressurizer level control setpoint) to verify the status of the pressurizer level control valve and determine the needed operator action to correct for the failure and the action affects plant and system conditions (RCP seal Flow).

Technical

Reference:

AOl-2, Malfunction of Reactor Control System, Revision 37 AOl-20, Malfunction of Pressurizer Level Control System, Revision 31 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO68C

15. Describe the response to a deviation from pressurizer level program.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question A012000.03 005 modified Comments:

Page 56

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

21. 032 AK1.01 021 Given the following:

- Unit 1 startup in progress.

- Reactor is critical in the source range.

- Intermediate Range monitors read 2 X 10% power.

- 12OVAC Vital Instrument Power Board 1-Il deenergizes.

Which ONE of the following describes effects associated with the loss of the instrument power board?

A. Reactor remains critical - Only one SRM is energized.

B. Reactor remains critical - Both SRMs remain energized.

C Reactor trips - Only one SRM is energized.

D. Reactor trips - Both SRMs remain energized.

DISTRA CTOR ANALYSIS:

A. Incorrect, Only one SRM would be energized but the reactor would trip due to Nl-32 failure. Plausible because lithe failure had been a different Instrument Power Board, the reactor could have remained critical and only one SRM being energized is correct as the question is written.

B. Incorrect, The reactor would trip due to Nl-32 failure and only one source range monitor would be energized. Plausible because lithe failure had been a different Instrument Power Board, the reactor could have remained critical and the failure would have been different on the SRMs.

C. Correct, Nl-32 and Nl-36 would fail causing bistables to trip. Failure of either one of these instruments would cause a reactor trip with the reactor power as stated in the question and with NI-32 failed, only SR instrument Nl-31 would be energized.

D. Incorrect, The reactor tripping is correct but due to NI-32 losing power but only one source range monitor would be energized. Plausible because the failure does result in the reactor tripping and if the failure had been a different Instrument Power Board, the failure would have been different on the SRMs.

Page 57

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 21 Tier: 1 Group 2 KIA: 032 AK1.01 Loss of Source Range Nuclear Instrumentation Knowledge of the operational implications of the following concepts as they apply to Loss of Source Range Nuclear Instrumentation:

Effects of voltage changes on performance Importance Rating: 2.5 / 3.1 IOCFRPart55: 41.8/41.10/45.3 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because it requires knowledge of the operational implications of a voltage change resulting in the loss of a source range instrument while the reactor is operating in thesource range.

Technical

Reference:

AOl-4, Nuclear Instrumentation Malfunctions, Revision 0029 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO92A

10. Identify all indications, alarms, permissives and trips associated with the SRMs.

32. Discuss excore instrument failures and their effects on the plant with and without operator action Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN bank question 032 AK1 .01 020 used on SQN 1/2009 RETAKE exam.

Comments:

Page 58

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

22. 033 AA2.12 022 Given the following:

- Unit 1 is in Mode 2 with startup in progress in accordance with GO-2, Reactor Startup, following a refueling outage.

- Due to damage to a source, only I source is installed in the core.

- During the past hour the l-M-4 Control Board indicators for NI-i 35 and NI-136 (1-NI-92-135A & 1-XI-92-136A) have trended as follows:

Time Nl-135 NI-136 0915 2.0 x i 02 1.0 x i 0930 4.5x102 8.0x102 0945 4.8x102 1.2x101 1000 5.6x102 1.4x101 Which ONE of the following indicates the first time the intermediate range monitors would not be able to meet channel check requirements?

REFERENCE PROVIDED A. 0915 B. 0930 C 0945 D. 1000 Page 59

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, 0975 is plausible if the math process is not understood. If the candidate incorrectly uses the sum of 1.0 plus 2.35 instead of the product, the result is greater than the highest reading IRM reading 3.0 (while the product of 7.0 and 2.35 is not greater) and it is the first time it occurs.

B. Incorrect, 0930 is plausible lithe note is misapplied because it is not understood that the reactor is critical and because it is the first time IRM-736 is not lower than IRM-135. See

• below.

C. Correct, See math below. 0945 is the first time the lowest reading IRM times 2.35 is not greater than the highest reading IRM.

D. Incorrect, See math below. 7000 is plausible because the lowest reading IRM times 2.35 is not greater than the highest reading IRM but it is not the first time.

Time NI-135 NI-136 Lowest

.2) 0915 2.0 x 1 02 1 .0 x 102 (1.0 x 1 times 2.35 = 2.352 2.0 0930 4.5 x 102 8.0 x 102 (4.5 x 102) times 2.35 = 1.4f1 > 8.0

(*but IRM-136 is not lower than IRM-135) 0945 4.8 x 102 1.2 x j1 (4.8x 102) times 2.35= 1.12841 not> 1.21 1000 5.6 x 10.2 1.4 x 101 (5.6 x 102) times 2.35 = 1.316 not> 1.41 Page 60

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 22 Tier: 1 Group 2 K/A: 033 AA2.12 Loss of Intermediate Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

Maximum allowable channel disagreement Importance Rating: 2.5* / 3.1*

10 CFR Part 55: 43.5 / 45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the maximum allowable channel disagreement during the failure of an Intermediate Range Nuclear Instrumentation.

Technical

Reference:

1-Sl-0-2A-02, 1900-0700 Shift and Daily Surveillance Log Mode Two, Rev 0034 Proposed references 1-Sl-0-2A-02, 1900-0700 Shift and Daily Surveillance to be provided: Log Mode Two page 22 Learning Objective: 3-OT-A010400

4. Explain Indications, Auto Actions, and Operator actions for an Intermediate Range Monitor (IRM) failure.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question 033 AA2.08 083 modified for use as an RO question.

Comments:

Page 61

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

23. 061 AK3.02 023 Given the following:

- Unit 1 is operating at 100% power.

- Annunciator 184-B, SEP 0-RM-1 02/1 03 RAD HI, alarms.

- 0-RM-90-102, Spent Fuel Pit, has only the GREEN light LIT and indicates 4 xl Ol mr/hr.

- 0-RM-90-103, Spent Fuel Pit, has the GREEN, AMBER, and RED indicating lights LIT and indicates 6.0 X 1 Q2 mr/hr.

Which ONE of the following identifies the ABGTS fan that the operator would have to start manually and the reason the Annunciator Response Instruction directs the ABGTS fans to be started?

A (1) Train AABGTS fan.

(2) To limit dose rates at the site boundary to less than federal limits.

B. (1) TrainAABGTSfan.

(2) To ensure any ABSCE boundary release is monitored by a PAM qualified radiation monitor.

C. (1) Train BABGTSfan.

(2) To limit dose rates at the site boundary to less than federal limits.

D. (1) Train BABGTSfan.

(2) To ensure any ABSCE boundary release is monitored by a PAM qualified radiation monitor.

Page 62

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, The Spent Fuel Pit Area radiation monitors are trained; 0-RM-90-102 is the Train A monitor and 0-RM-90-103 is the Train B monitor. Since 703 detects high radiation, ABGTS fan B automatically starts but the operator will have to start ABGTS fan A manually. The reason the ARI directs the fans to be started is to limit dose rates at the site boundary to less than federal limits by drawing down the ABSCE boundary pressure.

B. Incorrect, Plausible because having to manually start ABGTS Ian A is correct and the ABGTS discharge is routed out the shield building stack which is a PAM monitor.

C. Incorrect, Plausible because having to manually start ABGTS fan B would be correct if the status of the radiation monitors were reversed and the reason the ARI directs the fans to be started is to limit dose rates at the site boundary to less than federal limits by drawing down the ABS CE boundary pressure is correct.

D. Incorrect, Plausible because having to manually start ABGTS fan B would be correct if the status of the radiation monitors were reversed and the ABG TS discharge is routed out the shield building stack which is a PAM monitor.

Question Number: 23 Tier: 1 Group 2 K/A: 061 AK3.02 Area Radiation Monitoring (ARM) System Alarms Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms:

Guidance contained in alarm response for ARM system Importance Rating: 3.4 I 3.6 ID CFR Part 55: 41.5,41.10 /45.6 / 45.13 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the questions requires knowledge of the reason for action (Start ABGTS fans) required in an alarm response for ARM system.

Technical

Reference:

ARI-1 80-1 87, Common Radiation Detectors, Revision 0032 N3-3OAB-4001, Auxiliary Building Heating, Ventilation, Air Conditioning System (30, 31, 44), Revision 0031 Page 63

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO3OB

1. State the design basis of the ABGTS system in accordance with FSAR section 6.2.3.

2. State the function of the ABGT system in accordance with the system description.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO9O 004 modified for use on the 06/2011 exam.

Comments:

Page 64

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

24. 067 AK3.02 024 Given the following plant conditions:

- The Unit is at 100% power.

- AOl-30.1, Plant Fires, is in progress due to the report of a fire on site.

- An action required in the AOl is to check the status of both trains of Control Room Isolation (CR1) on the 1-M-6 Master Isolation Signal Status Panel.

Which ONE of the following identifies the reason for checking the status of the CR1?

A. To determine if Main Control Room abandonment is currently required due to the fire.

B. To ensure that an isolation of the Main Contol Room has automatically occurred.

C. To determine where to assemble the AUOs in preparation for an Appendix R fire being declared.

Db To evaluate the need for aligning CREVS suction to the other side of the Control Building.

Page 65

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, The action is not to determine if a MCR abandonment is or will be required. It is for determining adequacy the CREVS suction if a CR! has initiated.

Plausible because if a CR! had not initiated when needed, it could result in an MCR abandonment event but whether to or not to abandon the MCR would not be based on the CR! status.

B. Incorrect, The action is not to ensure the CR! has initiated. It is for determining adequacy the CREVS suction if a CR! has initiated. Plausible because the Master Status Panel is checked under other conditions (i.e. Safety injection actuation) to determine whether a manual actuation is required.

C. Incorrect, The action is not to determine where A UO reporting for Appendix R staging will be assembled. It is for determining adequacy ofthe CREVS suction if a CR! has initiated. Plausible because the pre-staging of the Appendix R AUOs is directed in the AOl.

D. Correct, the action is there in case the fire has resulted in a CR! that an evaluation can be made to determine if the suction intake to the CREVS should be realigned to the other side to the plant.

Page 66

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 24 Tier: 1 Group 2 K/A: 067 AK3.02 Plant fire on site Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site:

Steps called out in the site fire protection plan, FPS manual, and fire zone manual Importance Rating: 2.5 / 3.3 10 CFR Part 55: 41.5, 41.10 / 45.6 /45.13 10CFR55.43b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the reason for an step directed in the site protection procedure for responding to a fire on site.

Technical

Reference:

AOl-30.1, Plant Fires, Rev 0010 Proposed references None to be provided:

Learning Objective: 3-OT-A013000

5. State the major actions of AOl-30.1 PLANT FIRES.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question A013000 003 with correct answer relocated and minor wording changes in the answer choices.

Comments:

Page 67

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011

25. WiiiO EA1.3 025 Given the following:

- A reactor trip from 100% power and loss of Offsite power occurred 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago.

- ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (with RVLIS), is in progress.

- Pressurizer pressure is 785 psig and being lowered by auxiliary spray.

- RVLIS - ICCM PLASMA DISPLAYs indicate:

- Reactor vessel level at 81%.

- Core exit TC5 at 520°F.

Assuming NO additional operator action, which ONE of the following predicts the expected RVLIS and PZR level trends as the depressurization continues?

RVLIS PZR Level A. Increase Increase B. Increase Decrease C. Decrease Decrease D Decrease Increase Page 68

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible if the applicant determines that the bubble in Rx vessel head will decrease with the use of aux spray, and PZR level would increase with the use of aux spray.

B. Incorrect, Plausible if the applicant determines that the bubble in Rx vessel head will decrease with the use of aux spray, thus RVLIS level would increase and PZR level would decrease.

C. Incorrect, Plausible since Rx vessel head bubble would grow causing RVLIS to decrease but PZR level would increase not decrease.

D. Correct, With aux spray in service as the PZR pressure is decreased the bubble in Rx vessel head will grow. The expected plant response for this condition is that as the bubble in Rx head increases, RVLIS will decrease and PZR level will increase.

Page 69

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 25 Tier: 1 Group 2 K/A: W/E1OEA1.3 Natural Circulation with Steam Void in Vessel with/without RVLIS Ability to operate and/or monitor the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS):

Desired operating results during abnormal and emergency situations.

Importance Rating: 3.4 / 3.7 JO CFR Part 55: 41.7 / 45.5 I 45.6 IOCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate determine the expected plant response to lowering RCS pressure during Natural Circulation conditions.

Technical

Reference:

ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (with RVLIS), Rev 11 Proposed references None to be provided:

Learning Objective: 3-OT-EOP0000

13. Given plant conditions occurring as a result of depressurization during a natural circulation cooldown, determine whether or not RCS voiding is taking place.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN Bank question WE1O EA1.3 027 used on SQN Feb 2010 exam. Wording changed to apply to WBN and the correct answer relocated.

Comments:

Page 70

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

26. W/113 EK2.1 026 Given the following plant conditions:

- Unit 1 is operating at 100% power with the TD AFW pump tagged.

- A reactor trip occurs.

- The crew has entered FR-H.2, Steam Generator Overpressure, based on the Heat Sink CSF Status Tree.

- SG #2 pressure is 1230 psig, level is 32% NR and both are slowly rising.

- The pressure in the other 3 SGs is currently 1140 psig and stable.

Which ONE of the following is the status of SG #2 blowdown valves when FR-H.2 is entered and an action which will be initiated during the performance of FR-H.2?

A. Both of the blowdown valves would be open.

Isolate AFW flow to #2 SG until steam release path is established.

B. Both of the blowdown valves would be open.

Establish minimum AFW flow to #2 SG until steam release path is established.

C.v Only one of the blowdown valves would be open.

Isolate AFW flow to #2 SG until steam release path is established.

D. Only one of the blowdown valves would be open.

Establish minimum AFW flow to #2 SG until steam release path is established.

Page 77

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRA CTOR ANAL YSIS:

A. Incorrect, Plausible because no safety injection has occurred and the ID AFW pump being out of service removes another signal for the valves to close and the AFW being isolated to SO #2 by a procedure step until a steam release path is established is correct.

B. Incorrect, Plausible because no safety injection has occurred and the TD AEW pump being out of service removes another signal for the valves to close and there are other conditions where minimum AFW flow is established with the level below the normal level control setpoint as is the case in this question.

C. Correct, because the AEWP IA-A started, one of the SGBD valves on SG #2 would be closed and step 6 directs the AFW to be isolated to 50 #2 until a steam release path is established.

D. Incorrect, Plausible because only one of the SGBD valves on SG #2 being open is correct (AFW pump IA-A started which closed the outside valve) and there are other conditions where minimum AEW flow is established with the level below the normal level control setpoint as is the case in this question Question Number: 26 Tier: 1 Group 2 KIA: W/E13EK2.1 Steam Generator Overpressure Knowledge of the interrelations between the (Steam Generator Overpressure) and the following:

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

Importance Rating: 3.4 / 3.7 10 CFR Part 55: 41.7 / 45.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the automatic features associated with the SGBD valves, the interlocks associated with AFW pump starts, and the manual actions associated with control of AFW flow while performing FR-H.2 for steam generator overpressure.

Technical

Reference:

FR-H.2, Steam Generator Overpressure, Revision 6 Page 72

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

FR-H.2, Steam Generator Overpressure, Revision 6 1-45W600-57-8 R15 1-45W600-1-3 R15 Proposed references None to be provided:

Learning Objective: 3-OT-FRH0001

22. Explain the purpose for and basis of each step in FR-H.1, FR-H.2, FR-H.3, FR-H.4 and FR-H.5.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 73

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

27. W/E16 EG2.1.19 027 Given the following:

- Following an accident on Unit 1, the operator is using ICS screen 4RMI, Gaseous Rad Monitoring, to check the status of 1-RM-90-273 and 1-RM-90-274, Lower Containment Post Accident Radiation Monitors.

- 1-RM-90-274 data is displayed in BLUE text with a BLACK background.

Which ONE of the following identifies...

(1) the status of 1-RM-90-274 data and (2) the containment condition that will result in the monitors temporarily displaying unreliable readings?

Li)

A. BAD a rapid pressure change BAD a rapid temperature change C. SUSPECT a rapid pressure change D. SUSPECT a rapid temperature change Page 74

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because BAD data being displayed in BLUE text with a BLACK background is correct and the containment pressure can change rapidly during an accident but a change in pressure is not the cause of the radiation monitors providing unreliable data for a short period of time.

B. Correct, BAD data is displayed in BLUE text with a BLACK background and the containment post accident radiation monitors accuracy is affected by rapid increases and decreases in containment temperature and will provide unreliable data for a shod period of time following the rapid change.

C. Incorrect, Plausible because ALARM data is presented in RED text with a black background and the containment pressure can change rapidly during an accident but a change in pressure is not the cause of the radiation monitors providing unreliable data for a short period of time.

D. Incorrect, Plausible because ALARM data is presented in RED text with a black background and a rapid change in the containment temperature causing the radiation monitors to provide unreliable data for a shod period of time is correct.

Question Number: 27 Tier: 1 Group 2 KIA: W/E16 EG2.1.19 High Containment Radiation Conduct of Operations Ability to use plant computers to evaluate system or component status.

Importance Rating: 3.9 I 3.8 IOCFRPart55: 41.10/45.12 IOCFR55.43.b: Not applicable KIA Match: KJA is matched because the question requires the ability to evaluate data displayed on the ICS (plant computer system) to identify the status of the data.

Technical

Reference:

3-OT-SYS261, Integrated Computer System, Revision 6 ARI-265-268, Post Accident Radiation, Revision 7 WBN Integrated Computer System Users Guide Proposed references None Page 75

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 to be provided:

Learning Objective: 3-OT-SYS26J

5. Describe how to tell if a data point is good, bad, or suspect, based on screen item display colors.

3-OT-SYSO9OA

07. Determine Interlocks and/or cause-effect relationships between the Rad Monitoring Systems (ARM & Process) and the areas they monitor.

Include HVAC systems and area isolations.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/201 1 NRC exam Comments:

Page 76

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

28. 003 K4.04 028 Given the following:

- Unit I is at 100% power, steady state, operation.

- RCP seal injection flows are:

RCP#1 RCP#2 RCP#3 RCP#4 10.2 gpm 7.8 gpm 9.0 gpm 12.2 gpm Which ONE of the following identifies how 1-HIC-62-89A, Charging Seal Water Flow Controller, should be adjusted to establish all RCP seal injections flows within the normal range per SOl-68.02, Reactor Coolant Pumps?

A. Raise output to throttle open 1-FCV-62-89 to raise flow on RCP #2.

B Raise output to throttle close 1 -FCV-62-89 to raise flow on RCP #2.

C. Lower output to throttle open 1 -FCV-62-89 to lower flow on RCP #4.

D. Lower output to throttle close 1 -FCV-62-89 to lower flow on RCP #4.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because raising the output is correct but the action would not throttle the valve open to raise flow.

B. Correct, raising the output would throttle valve closed which would raise the flow on RCP #2, which is lower than the normal range of 8-73 gpm.

C. Incorrect, Plausible because lowering the output would throttle the valve open but that action will lower the flow not raise the flow as is needed.

D. Incorrect, Plausible because the valve needs to be closed to raise flow but lowering the output would open the valve.

501-68.02 Section 3.0 PRECAUTIONS AND LIMITATIONS P. Maintain seal injection temp less than 130°F, and flow 8 to 13 gpm to each RCP.

Page 77

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 28 Tier: 2 Group 1 K/A: 003 K4.04 Reactor Coolant Pump System (RCPS)

Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following:

Adequate cooling of RCP motor and seals Importance Rating: 2.8 / 3.1 10 CFR Part 55: 41.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the knowledge of the normal limits for seal water flow to the Reactor Coolant Pumps.

Technical

Reference:

SOI-68.02, Reactor Coolant Pumps, Revision 0034 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68B

05. Describe the RCPs Seal Injection System, including:

a. Flowpath!Components

b. Flowrate

c. Purpose Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: SQN bank question 003 K4.04 modified to make a different answer correct.

Comments:

Page 78

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

29. 004A4.18 029 Given the following:

- While performing AOl-34, Immediate Boration the DAC places 1-HS-62-138A, EMER BORATE, to OPEN.

- The OAC releases the handswitch 2 seconds after the RED indicating light is LIT.

Which ONE of the following identifies...

(1) how the Emergency Borate Valve will respond and (2) the panel in the main control room where emergency boration flow can be verified?

A (1) The valve will stop open travel when the control switch is released.

(2) on 1-M-5 B. (1) The valve will stop open travel when the control switch is released.

(2) on 1-M-6 C. (1) The valve will continue to travel until it reaches full open.

(2) on 1-M-5 D. (1) The valve will continue to travel until it reaches full open.

(2) on 1-M-6 Page 79

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Correct, the Emergency Borate Valve is one of a few MOVs that does not have a seal-in circuit. The valve will stop open travel when the handswitch is released.

The flow indication (1-Fl-62-137A, EMERG BORATE FLOW) is located on Panel I -M-5.

B. Incorrect, Plausible because the valve stopping open travel when the handswitch is released is correct and there is a boron flow indication on 1-M-6 but it is the flow indication (1-Fl-62-139) used when boration is through the blender.

C. Incorrect, Plausible because most MOVs will continue the open travel until stopped by a torque or limit switch when full open and the boron flow indication being on 7 -M-5 correct.

D. Incorrect, Plausible because most MOVs will continue the open travel until stopped by a torque or limit switch when full open and there is a boron flow indication on 7-M-6 but it is the flow indication (7-Fl-62-739) used when boration is through the blender.

Page 80

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 29 Tier: 2 Group 1 KIA: 004 A4.18 Chemical and Volume Control System Ability to manually operate and/or monitor in the control room:

Emergency borate valve Importance Rating: 4.3 / 4.1 JO CFR Part 55: 41.7 / 45.5 to 45.8 IOCFR55.43.b: Not applicable K/A Match: Question requires the ability to manually operate and monitor the emergency borate valve in the control room as to how the valve reacts to control switch operation and how expected results can be monitored.

Technical

Reference:

Control Board pictures 1-47W61 1-62-2 R8 1 -45W760-62-5 Ri 1 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO62A

29. Discuss how to perform manual immediate boration, include when this is necessary.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank SRO question TR0312 001 modified for the RO exam Comments:

Page 87

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

30. 004 K6.01 030 Given the following:

- A Reactor trip occurs, subsequent to a loss of offsite power.

- ES-0.2, Natural Circulation Cooldown, has been implemented.

- The operating crew has determined 1500 gallons of boric acid must be injected to meet Cold Shutdown RCS boron concentration (CB) calculated using REACTINW.

Which ONE of the following identifies the requirement for RCS boration in accordance with ES-0.2?

A The entire amount of boric acid must be injected prior to the initiation of the cooldown.

B. The entire amount of boric acid plus 15% must be injected prior to the initiation of the cooldown.

C. The cooldown can begin during the boron addition but the entire amount of boric acid is required to be injected prior to reaching Cold Shutdown.

D. The cooldown can begin during the boron addition but the entire amount of boric acid plus 15% is required to be injected prior to reaching Cold Shutdown.

Page 82

06/2017 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Correct, ES-U. 2 step 3 requires the initiation of RCS boration to cold shutdown based on determination of cold shutdown boron concentration. Step 9 which initiates the RCS coo/down identifies that the coo/down cannot begin until the RCS boron concentration if greater than the cold shutdown boron concentration and the total number of ga/Ions of boron calculated in step 3a have been added.

B. Incorrect, Plausible because the coo/down cannot begin until the RCS boron concentration is greater than the cold shutdown boron concentration and the total number of ga/Ions of boron calculated in step 3a have been added but while there is no requirement for an additional 15% to be added there is a caution prior to step 9 that identifies that the actual RCS boron concentration will be as much as 75%

higher than the calculated RCS boron concentration.

C. Incorrect, Plausible because, with forced circulation, a cooldown can begin prior to the RCS boron concentration exceeding the cold shutdown boron concentration and because the entire boration needed for Cold Shutdown being added is a correct statement.

D. Incorrect, Plausible because, with forced circulation, a cooldown can begin prior to the RCS boron concentration exceeding the cold shutdown boron concentration and the total gal/on of boron needed for Cold Shutdown being added and the caution prior to step 9 identifying that the actual RCS boron concentration will be as much as 15% higher than the calculated RCS boron concentration could be misinterpreted.

Page 83

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 30 Tier: 2 Group 1 K/A: 004 K6.01 Chemical and Volume Control System Knowledge of the effect of a loss or malfunction on the following CVCS components:

Spray/heater combination in PZR to assure uniform boron concentration Importance Rating: 3.1 / 3.3 IOCFRPart55: 41.7/45.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the knowledge of the effect (change in initiation of RCS cooldown and boron injection requirement via the CVCS) due to a loss of pressurizer spray to assure uniform boron concentration.

Technical

Reference:

ES-0.2, Natural Circulation Cooldown, Rev 21 Proposed references None to be provided:

Learning Objective: 3-OT-EOP0000

12. Given the results of RCS boron concentration samples taken while on natural circulation, explain why the sample taken from the RCS cold leg is higher than the pzr CB following RCS boration

15. Explain the purpose for and basis of each step in E-0, ES-0.0, ES-0.1, ES-0.2, ES-0.3, and ES-0.4.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question EOP0000.12 001 modified Comments:

Page 84

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

31. 005 A2.02 031 Given the following plant conditions:

- Unit 1 is in Mode 5, solid water operation, with Train A RHR in service.

- RCS temperature is 180°F.

- RCS pressure is 320 psig.

Subsequently:

- RCS pressure begins increasing uncontrolled.

Which ONE of the following identifies...

(1) the pressure at which the RHR suction relief valve is set to open, and, (2) the first action directed by AOl-I 4, Loss of RHR Shutdown Cooling?

Relief Opens First Action A. 370 psig Stop RHR Pump B. 370 psig Stop Charging Pump C. 450 psig Stop RHR Pump D 450 psig Stop Charging Pump Page 85

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because 370 psig is the pressure identified in AOl-14 Section 3.4 as the pressure requiring the RHR to be stopped and the RHR pump(s) are stopped during performance of Step 3 of A 01-14 Section 3.4.

B. Incorrect, Plausible because 370 psig is the pressure identified in AOl-74 Section 3.4 as the pressure requiring the RHR to be stopped in Step 3 and step I of Section 3.4 requires the charging pumps to be stopped if the plant is in water solid operation.

C. Incorrect, Plausible because the setpoint for the RHR suction relief valve is 450 psig and the RHR pump(s) are stopped during performance of Step 3 of A0I-14 Section 3.4.

D. Correct, the setpoint for the RHR suction relief valve is 450 psig and AOl-14 Section 3.4, RCS Shutdown Cooling During RHR Shutdown Cooling Step I directs the charging pump to be stopped if the plant is in water solid operation.

Question Number: 31 Tier: 2 Group 1 K/A: 005 A2.02 Residual Heat Removal System (RHRS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Pressure transient protection during cold shutdown Importance Rating: 3.5 I 3.7 10 CFR Part 55: 41.5/43.5/45.3/45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of when the RHR suction relief valve will open when RCS pressure raising uncontrolled (ability to predict the impact) and the knowledge of the procedure used to mitigate the consequences of the malfunction causing the pressure transient.

Technical

Reference:

AOl-14, Loss of Shutdown Cooling, Revision 0037 Proposed references None Page 86

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 to be provided:

Learning Objective: 3-CT-AOl 1400

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

a. Recognize Entry conditions.

b. Respond to Action steps.

c. Respond to Contingencies (RNO column).

U. Respond to Notes & Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question A011400.04 001 modified by changing stem, first half of the question being asked, distractors and correct answer.

Comments:

Page 87

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

32. 006 K6.03 032 Given the following:

- Unit 1 is operating at 100% power when a LOCA occurs.

- The crew is performing ES-I .3, Transfer to Containment Sump.

- 1-FCV-63-175, SI PMP B RECIRC TO RWST, cannot be closed from the MCR.

As the performance of ES-1 .3 continues, which ONE of the following identifies how the valve failure affects the...

(1) isolation of Safety Injection Pump recirculation flow and (2) alignment of the RHR pump I B-B to supply suction the SIPs?

A (1) Can be isolated from the MCR.

(2) Can be aligned to supply suction to the SIPs with 1-FCV-63-175 open.

B. (1) Must be isolated locally.

(2) Can be aligned to supply suction to the SIPs with 1-FCV-63-175 open.

C. (1) Can be isolated from the MCR.

(2) Can NOT be aligned to supply suction to the SIPs until 1-FCV-63-175 is closed.

D. (I) Must be isolated locally.

(2) Can NOT be aligned to supply suction to the SIPs until I-FCV-63-175 is closed.

Page 88

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS IRA CTOR ANALYSIS:

A. Correct, there is another valve (1-FCV-63-4) in series with 1-FC V-63-175 that can be operated from the MCR. 1-FCV-63-l 75 being closed is an interlock for opening 7-FCV-63-1 7 which allows the RHR pump 18-B to supply suction to the S/Ps but there is an alternate interlock for the series valve being closed that if made will allow 7-FCV-63-71 to be opened.

B. Incorrect, Plausible because there are valve failures that will require local isolation and while 7-FCV-63-175 being closed is an interlock for opening 1-FCV-63-7 I there is a parallel circuit to the interlock allowing the RHR pump lB-B to supply suction to the SIPs when the series valveis closed.

C. Incorrect, Plausible because there is a series valve that can be closed from the MCR to isolate the recirc flow path and because 7-FCV-63-1 75 being closed is interlocked for opening 1-FCV-63-1 I (allowing the RHR pump lB-B to supply suction to the SIPs).

D. Incorrect, Plausible because there are valve failures that will require closing locally to isolate a flow path and because 1-FCV-63-175 being closed is interlocked for opening 1-FCV-63-1 I (allowing the RHR pump lB-B to supply suction to the SIPs).

Page 89

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 32 Tier: 2 Group 1 K/A: 006 K6.03 Emergency Core Cooling System (ECCS)

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

Safety Injection Pumps Importance Rating: 3.6 / 3.9 IOCFRPart55: 41.7/45.7 I OCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of how the ECCS is affected by the failure of a Safety Injection Pump recirculation valve to close (malfunction) during the alignment to cold leg recirculation.

Technical

Reference:

ES-i .3, Transfer To Containment Sump, Revision 17 Proposed references None to be provided:

Learning Objective: 3-OT-EOPO100

20. Discuss and justify the priority of usage given to procedure ES-i .3, Transfer to RHR Containment Sump Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question EOPO100.20 009 modified Comments:

Page 90

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

33. 007 A3.01 033 Which ONE of the following identifies the pressure that the relief valve on the discharge of the RHR pumps will start relieving and the tank where the flow through the valve will be routed?

Pressure Tank A. 550 psig RCDT B. 550 psig PRT C. 600 psig RCDT D 600 psig PRT DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible because 550 psig is the maximum RHR pump discharge pressure allowed to be maintained in accordance with the System Operating Instruction when RHR system is in service and the RCDT is an RCS tank inside containment (like the PRT) which does receive flow and leakoifs from RCS related components.

B. Incorrect, Plausible because 550 psig is the maximum RHR pump discharge pressure allowed to be maintained in accordance with the System Operating Instruction when RHR system is in service and the PRT being the tank that receives flow passing through the valve is correct.

C. Incorrect, Plausible because 600 psig is the pressure that the valve starts relieving and the RCDT is an RCS tank inside containment aike the PRT) which does receive flow and leako ifs from RCS related components.

D. Correct, the RHR discharge relief valve starts relieving at 600 psig and is routed to the PRT.

Page 91

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 33 Tier: 2 Group 1 KIA: 007 A3.01 Pressurizer Relief Tank/Quench Tank System (PRTS)

Ability to monitor automatic operation of the PRTS, including:

Components which discharge to the PRT Importance Rating: 2.7 /2.9 JO CFR Part 55: 41.7 / 45.5 to 45.8 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the ability to monitor the automatic operation of the RHR discharge pressure and PRT level to know lithe relief valve is relieving (which did occur at SQN allowing approximately 10,000 gallons to be passed to the PRT prior to termination)

Technical

Reference:

1-47W811-1 R52 1-46W813-1 R43 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68C

21. Describe the flow path of sources of supply, discharges, vents, drains, leakoff, and connections/penetrations that intertie this system to other systems.

Cognitive Level:

Higher Lower X Question Source:

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

Comments:

Page 92

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

34. 008 A3.04 034 Given the following:

- Unit I RCS is at 320°F and 350 psig with plant heatup to normal RCS operating temperature and pressure in progress.

- CCS is in a normal alignment with CCS pumps lA-A and C-S running.

- CCS IA-A pump suffers a catastrophic shaft failure.

Which of the following completes the two statements below?

CCS pump 1 B-B will (1) to supply the Unit 1 Train A CCS supply header.

Tech Spec 3.7.7, Component Cooling System, LCO entry (2)

A (1) automatically start (2) is currently required B. (1) automatically start (2) is NOT currently required C. (1) be requited to be manually aligned (2) is currently required D. (1) be requited to be manually aligned (2) is NOT currently required Page 93

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, With the CCS in a normal alignment, the lB-B pump will be aligned to Train A and will automatically start when the pressure drops to 40 psig restoring flow to Train A supply header but the lB-B pump starting does not restore Tech Spec operability to CCS Train A, thus, the LCO is required to be entered because the LCO is required in Mode 4.

B. Incorrect, Plausible because the 18-B pump automatically starting is correct and because the CCS T/S LCO is not required in all operational modes but is required in the current Mode 4 condition.

C. Incorrect, Plausible because the 18-B pump would have been required to be manually realigned to supply flow if the C-S pump had been the pump that failed and because the CCS T/S LCO being required in the current Mode 4 condition is correct.

D. Incorrect, Plausible because the lB-B pump would have been required to be manually realigned to supply flow if the C-S pump had been the pump that failed and because the CCS T/S LCO is not required in all operational modes but is required in the current Mode 4 condition.

Question Number: 34 Tier: 2 Group 1 KIA: Component Cooling Water System (COWS)

Ability to monitor automatic operation of the COWS, including:

Requirements on and for the COWS for different conditions of the power plant Importance Rating: 2.9 / 3.2 10 CFR Part 55: 41.7 I 45.5 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the knowledge of how the system is normally aligned, the ability to monitor the automatic operation of the COWS (a low pressure condition resulting in an automatic pump start) and the TIS required actions resulting from a failure on the system.

Technical

Reference:

SOl-70.01, Component Cooling Water System, Revision 0064 Tech Spec 3.7.7, Component Cooling System Page 94

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO7OA

3. Describe the CCS pumps, include power supply, pump type, capacity, lubrication, and logic.

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

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 X Lower Question Source:

New Modified Bank X Bank Question History: WBN Bank question SYSO7OA.15 004 modified.

Comments:

Page 95

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

35. 008 K1.05 035 Given the following plant conditions:

- The Unit I Component Cooling Water System (CCS) Surge Tank level is decreasing.

Which ONE of the choices below completes the following statement?

The normal make-up to the CCS surge tank is supplied from the (1) and if the emergency makeup supply was needed to maintain a level in the surge tank, it would (2)

Li)

A Demineralized Water System require local action to align the supply to the tank B. Demineralized Water System open automatically as tank level drops C. Primary Water System require local action to align the supply to the tank D. Primary Water System open automatically as tank level drops DISTRA CTOR ANAL YSIS:

A. Correct, The normal makeup supply is from the Demineralized Water System and if the emergency supply from the ERCW system was needed, the spool piece would have to be installed and a manual valve opened.

B. Incorrect, Plausible because the normal supply is from the Demineralized Water System and while the emergency makeup is required to be manually aligned, the normal makeup will open automatically as the tank level drops.

C. Incorrect, Plausible because the Primary Water System is the normal supply for other systems on the primary side of the plant and the use of the emergency makeup supply requiring action to align is correct.

D. Incorrect, Plausible because the Primary Water System is the normal supply for other systems on the primary side of the plant and while the emergency makeup is required to be manually aligned, the normal makeup will open automatically as the tank level drops.

Page 96

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 35 Tier: 2 Group 1 KIA: 008 K1.05 Component Cooling Water System (COWS)

Knowledge of the physical connections and/or cause-effect relationships between the COWS and the following systems: Sources of makeup water.

Importance Rating: 3.0 / 3.1 JO CFR Part 55: 41.2 to 41.9 / 45.7 to 45.9 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the applicant must identify the physical connection for normal make-up to the OCS surge tank and the cause-effect relationship for the emergency makeup supply as the tank level drops.

Technical

Reference:

1-47W61 1-70-1, R9 1 -47W859-1, R46 SOl-70.01 Component Cooling Water System, Revision 0064 ARI-241-253, Revision 10 AOl-i 5, Loss of Component Cooling Water (OCS),

Revision 0032 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO7OA

9. Describe the CCS Surge Tanks; include purpose, capacity, and method of makeup to them.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO7OA.09 002 modified. The fill in the blank statement was modified to require identification of the normal supply (instead of the emergency supply) and the cause -effect relationship was change to the use of the emergency supply (instead Page 97

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question History: WBN bank question SYSO7OA.09 002 modified. The fill in the blank statement was modified to require identification of the normal supply (instead of the emergency supply) and the cause -effect relationship was change to the use of the emergency supply (instead of the normal supply.) This changed the correct answer and the location of the correct answer is also in a different location.

Comments: Question SYSO7OA.09 002 was used on the 12/2009 NRC exam.

Page 98

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

36. 010 A2.01 036 Given the following:

- Unit us at 100% power with an RCS dilution in progress.

- Pressurizer pressure is 2235 psig with both spray valves closed.

- Annunciator 14-E M-l THRU M-6 MOTOR TRIPOUT alarms.

- Operator notes RCS pressure slowly dropping.

- There are NO other annunciators in alarm.

Which ONE of the following identifies...

(1) the breaker trip that caused the motor tripout alarm and (2)the action, if taken, associated with 1-PIC-68-340A, PZR PRESS MASTER CONTROL that would restore RCS pressure in accordance with AOl-18, Malfunction of Pressurizer Pressure Control System?

A. (1) Centrifugal Charging Pump.

(2) Lower the controller output.

B. (1) Centrifugal Charging Pump.

(2) Raise the controller output.

C (1) Pressurizer heater Control Group C.

(2) Lower the controller output.

D. (1) Pressurizer heater Control Group C.

(2) Raise the controller output.

Page 99

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible because the charging pump trip would cause the motor trip out alarm and the pressurizer pressure would start to drop along with the level (due to seal leakoff); however a charging pump trip would also cause additional annunciators (Lo Charging Flow.) Lowering the controller output is correct to energize the backup heaters.

B. Incorrect, Plausible because the charging pump trip would cause the motor trip out alarm and the pressurizer pressure would start to drop along with the level (due to RCP seal leakoff); however a charging pump trip would also cause additional annunciators(Lo Charging Flow.) Raising the controller output can be done but it would cause spray valves to open and reduce pressure further.

C. Correct, while the pressurizer heater breaker is not a motor breaker, if the heater breaker trips it will bring it the motor trip out annunciator. Following the breaker trip, the pressurizer pressure will start dropping and AOl-18 directs the pressure to be monitored and trending to desired pressure. If pressure is not trending to desired pressure, then the pressure is to be restored using manual control of 1-PIC-68-340A, Master Pressure Controller or PZR Spray controllers or Pressurizer Heaters. To raise Pressure the controller output must be lowered. ARI for low pressure also directs use of the controller.

D. Incorrect, Plausible because the pressurizer heater breaker trip will bring in the motor trip out annunciator and the controller output can be raised but that action would cause spray valves to open and reduce pressure further.

Question Number: 36 Tier: 2 Group 1 KIA: 010 A.2.0l Pressurizer Pressure Control System Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Heater failures Importance Rating: 3.3 / 3.6 10 CFR Part 55: 41.5 / 43.5 / 45.3 / 45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the applicant to predict the impact of a pressurizer heater failure (malfunction) by Page 700

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: KJA is matched because the question requires the applicant to predict the impact of a pressurizer heater failure (malfunction) by determining the action required to identify the failure and then to take manual action in accordance with the procedure to stabilize the unit as a result of the failure.

Technical

Reference:

AOl-18, Malfunction of Pressurizer Control System, Revision 0022 ARI-8-14, Aux Power, Revision 0008 Proposed references None to be provided:

Learning Objective: 3-OT-SYS1800

5. Explain the Operator Actions for dropping RCS pressure.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN bank question SYSO68C 030. Changed location of correct answer by swapping C and D, also swapped A and B distractors. Minor wording changes in the stem and swapped heater bank that trips.

Comments:

Page 701

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

37. 012 K5.01 037 Which ONE of the following identifies (1) two parameters monitored to ensure compliance with the Tech Spec LCO 3.4.1, DNB Limits and (2) a reactor trip which provides protection against Departure from Nucleate Boiling (DNB)?

A. Thermal power and Al Low RCS flow B. Thermal power and Al NIS Flux High (Low setpoint)

C RCS temperature and pressure Low RCS flow D. RCS temperature and pressure NIS Flux High (Low setpoint)

Page 702

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

The parameters monitored by T/S LCO 3.4. 1, DNB reactor coolant temperature, reactor coolant pressure, reactor coolant flow rate. Thermal power does affect DNB and the AED is used to adjust the OTzIT trip setpoint, which is used to prevent DNB, in order to account for power imbalances in the upper and lower portions of the core, but neither thermal power nor OTAT are parameters included in the LCO 3.4. 1 A. Incorrect, Plausible because the thermal power affects DNBR, and delta flux potentially affects the OTrIT trip setpoints and because the Low RCS flow trip provides protection against DNB.

B. Incorrect, Plausible because the thermal power affects DNBR, and delta flux potentially affects the OTzIT trip setpoints and because the NIS Flux Hi (low setpoint) is a reactor trip associated with parameters (thermal power and Al) that are associated with DNB C. Correct, Both RCS temperature and pressure are parameters monitored to ensure compliance with DNB and the Low RCS flow trip protects against DNB.

D. Incorrect, Plausible because both RCS temperature and pressure are parameters monitored to ensure compliance with the DNB Tech Spec and because the NIS Flux Hi (low setpoint) is a reactor trip associated with parameters (thermal power and Al) that are associated with DNB.

Question Number: 37 Tier: 2 Group 1 K/A: 012 K5.01 Reactor Protection Knowledge of the operational implications of the following concepts as the apply to the RPS.

DNB Importance Rating: 3.3 / 3.8*

JOCFRPart55: 41.5/45.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires applicant to recall factors affecting DNB and reactor trips designed to prevent DNB.

Page 103

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

Technical Spec 3.3.1, RTS Instrumentation Technical Specification 3.4.1, RCS Pressure, Temperature, and Flow DNB Limits, Amendment 7.

N3-99-4003, Reactor Protection System, Revision 0021 3-OT-SYSO99A, Reactor Protection System (RCS),

Revision 9 Proposed references None to be provided:

Learning Objective: 3-OT-TIS 0304

1. Demonstrate the ability to extract specific information from the Technical Specifications and Technical Requirements, as they pertain to RCS.

3-OT-SYSO99A

1. Explain the purpose of the Reactor Protection System.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question 012 K5.01 037 modified.

Comments:

Page 704

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011

38. 012 K5.02 03$

Which ONE of the following reactor protection system trips is designed to ensure that the allowable heat generation rate (kw/ft) of the fuel is NOT exceeded?

A Over Power AT B. NIS Positive Rate C. Over Temperature AT D. Intermediate Range Hi Flux DISTRACTOR ANALYSIS:

A. Correct, the OP/iT trip backs up the Power Range Neutron flux High trip to ensure that the allowable heat generation rate (kw/ft) is not exceeded.

B. Incorrect, Plausible because the positive rate trip does provide protection against a condition being exceeded but it is for an ejected rod accident.

C. Incorrect, Plausible because the OT/iT trip does provide protection against a condition being exceeded but it is for DNB protection and is a backup to the low pressurizer pressure trip.

D. Incorrect, Plausible because the trip does provide protection against a condition being exceeded and is a backup to the Power Range Hi Flux low setpoint Trip.

Page 105

06/2011 Watts Bar SRD NRC License Exam 6/22/2011 Question Number: 38 Tier: 2 Group 1 K/A: 012 K5.02 Reactor Protection System (RPS)

Knowledge of the operational implications of the following concepts as the apply to the RPS:

Power density Importance Rating: 3.1* / 3*3*

10 CFR Part 55: 41.5 / 45.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the protection afforded (operational implication) to prevent exceeding the maximum allowed power density (kw/ft)

Technical

Reference:

N3-99-4003, Reactor Protection System, Revision 0021 3-OT-SYSO99A, Reactor Protection System, Revision 9 Proposed references None to be provided:

Learning Objective: 3-OT-5Y5099A 17 Identify the Reactor trips and give setpoints and list logic requited for the Reactor trips.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: INPO Bank question from Prairie Island used on an 2004 exam with two distractors changed for use at WBN. No changes in stem; thus not a bank modified question.

Comments: Similar question also used on a Braidwood exam in 2002.

Page 706

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

39. 013 G2.1.20 039 Given the following:

- The operating crew has entered ES-1.J, SI Termination.

- After depressing the SI RESET pushbuttons (1-HS-63-134A, SI RESET TR A and 1-HS-63-134B, SI RESET TR B) the following conditions exist on 1-XA-55-4A, BYPASS, INTERLOCK AND PERMISSIVE.

- Window 70-A SI ACTUATED is LIT.

-

- Window 70-B AUTO SI BLOCKED is LIT.

-

Which ONE of the following identifies...

(1) how the Integrated Computer System tICS) will indicate which SSPS Train has failed to reset and (2) the first method the operator sent locally to the respective Train SSPS cabinets would take to reset the failed SSPS Train locally?

A (1) The failed train will indicate ACT in RED.

(2) OPEN both 48v breakers in the SSPS Trains R panel.

B. (1) The failed train will indicate ACT in RED.

(2) Place Safeguards Test Cabinet RESET switch to RESET and RELEASE in the SSPS Trains R panel.

C. (1) The failed train will indicate ACT in CYAN.

(2) OPEN both 48v breakers in the SSPS Trains R panel.

D. (1) The failed train will indicate ACT in CYAN.

(2) Place Safeguards Test Cabinet RESET switch to RESET and RELEASE in the SSPS Trains R panel.

Page 107

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Correct, The failed SSPS train will indicate ACT in RED and the initial attempt to reset the Train locally at the SSPS cabinets will be by opening both 48v power supplies in the respective Trains R panel.

B. Incorrect, Plausible because the failed SSPS train indicating ACT in RED is correct and placing the Safeguards Test Cabinet RESET switch to RESET and RELEASE in the SSPS Trains R panel is performed if opening the 48v power supplies is not successful.

C. Incorrect, Plausible because inserted data is indicated in CYAN on the ICS screen and placing the Safeguards Test Cabinet RESET switch to RESET and opening both 48v power supplies in the respective Trains R panel is correct.

D. Incorrect, Plausible because inserted data is indicated in CYAN on the ICS screen and placing the Safeguards Test Cabinet RESET switch to RESET and RELEASE in the SSPS Trains R panel is performed if opening the 48v power supplies is not successful.

Page 108

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 39 Tier: 2 Group 1 KIA: 013 G2.1.20 Conduct of Operations Engineered Safety Features Actuation System (ESFAS)

Ability to interpret and execute procedure steps.

Importance Rating: 4.6 / 4.6 JO CFR Part 55: 41.10 /43.5/45.12 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the questions requires knowledge of the indications available to the operator to determine equipment status when executing the procedure step and how an appendix would be used to successfully complete the step if required.

Technical

Reference:

ES-1.1, SI Termination, Revision 0017 Proposed references None to be provided:

Learning Objective: 3-OT-EOPO100

28. Describe the actions in ES-1.1, SI Termination, required in the event that SI does not reset.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN question 012 G2.4.35 038 modified for use on the WBNO6I2O11 exam.

Comments: 012 G2.4.35 038 was used on the WBN 2009 spring exam which was 3 exams back.

Page 709

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

40. 013 K2.01 040 Which ONE of the following identifies the 120v AC Vital Instrument Power Board(s) that supply power to energize the input relays for the Train B SSPS?

A. 1-Il, only B. 1-lV, only C. 1-Il and 1-IV, only D 1-I, 1-Il, 1-Ill, and 1-lV DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the Train B SSPS slave relays are supplied only by 720v Vital Instrument Power Board. 7-Il.

B. Incorrect, Plausible because the 720v Vital Instrument Power Board. 1-lV is a Unit 7 Train B board that does supply the Train B SSPS system and there are parts of the S$PS system that only has one power supply.

C. Incorrect, Plausible because the 120v Vital Instrument Power Board. 1-Il and 1-lV are the only supplies to the Train B SSPS Logic Cabinet.

D. Correct, There are four bays for the input relays with each set of relays being supplied by a separate 120v Vital Instrument Power Board. (7-I through 7-lV)

Page 770

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 40 Tier: 2 Group 1 K/A: 013 K2.01 Knowledge of bus power supplies to the following:

ESFAS I safeguards equipment control Importance Rating: 3.6 / 3.8 10 CFR Part 55: 41.7 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the applicant to identify the source(s) of power to the Train B SSPS input relays.

Technical

Reference:

1-45W706-1 Rev 68 1-45W706-2 Rev 63 J-45W706-3 Rev 48 1-45W706-4 Rev 50 N3-99-4003, Reactor Protection System, Revision 0021 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO99A

2. Sketch a basic drawing of the Solid State Protection System.

8. Briefly discuss the input relays, Logic Section and Output Section of the SSPS.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/20 1 1 NRC exam Comments:

Page 711

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

41. 022 G2.4.45 041 Given the following:

- Unit 1 is Mode 3 following a shutdown.

- Containment Lower Compartment Cooler D-B is out of service for maintenance.

- Loss of an electrical board results in the loss of Containment Lower Compartment Coolers A-A and C-A.

- The following annunciators go into alarm:

Window 102-A - CRDM COOLER FLOW LO Window 104-A - LWR CNTMT CLR FLOW LO Window 104-B - LWR CNTMT TEMP HI Window 144-A - ICE COND INLET DOOR OPEN

- Lower Containment temperature is 128°F and rising.

- Lower Containment pressure is +0.18 psid and rising slowly.

As containment temperature and pressure continue to rise, which ONE of the following is the first required action in accordance with Annunciator Response Instructions for the annunuciators listed above?

A. If Lower Containment pressure exceeds the Tech Spec limit for greater than 50 minutes, then start the Air Return Fan(s).

B. If Lower Containment pressure exceeds the Tech Spec limit for greater than 50 minutes, then start the Containment Spray Pump(s).

C If Lower Containment Temperature exceeds 160°F and can NOT be reduced to less than the Tech Spec Limit within 50 minutes, then start the Air Return Fan(s).

D. If Lower Containment Temperature exceeds 160°F and can NOT be reduced to less than the Tech Spec Limit within 50 minutes, then start the Containment Spray Pump(s).

Page 772

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because starting the containment air return fans is directed but not to reduce containment pressure. They are started to reduce a high containment temperature.

B. Incorrect, Plausible because starting the containment spray pumps would reduce the containment pressure and is the next action directed in ARI 704-B to reduce containment temperature.

C. Correct, ARI 704-B directs starting the containment air return fans if containment is greater than 7 60°F temperature and cannot be reduced to less than 720°F within 60 minutes.

D. Incorrect, Plausible because starting the containment spray pumps would reduce the containment temperature and is the next action directed in ARI 704-B to reduce containment temperature.

Question Number: 41 Tier: 2 Group 1 K!A: 022 G2.4.45 Containment Cooling System (CCS)

Emergency Procedures I Plan Ability to prioritize and interpret the significance of each annunciator or alarm.

Importance Rating: 4.1 / 4.3 10 CFR Part 55: 41.10/43.5/45.3 / 45.12 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to prioritize and implement the actions in ARIs as a result of degraded conditions on the Containment Cooling System.

Technical

Reference:

ARI-102-1-8, HVAC & CVCS, Revision 0026 ARI-138-144, HVAC, 480 BDS, ICE COND, Revision 0019 N3-3ORB-4002, Reactor Building Ventilation System, Revision 0019 Proposed references None to be provided:

Page 773

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Learning Objective: 3-OT-SYSO3OC

1. State the Design Basis of the Containment Air Cooling/Containment Purge/Continuous Vent systems in accordance with FSAR section 9.4.7.

3-OT-SYSO3OD

6. Given certain plant conditions determine if manual start of air return fans would be acceptable/required.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/20 1 1 NRC exam.

Comments:

Page 714

06/2011 Watts Bar SRD NRC License Exam 6/22/2011

42. 025 A2.03 042 Given the following:

- Unit I operating at 100% power.

- The operating crew is in the process of establishing a containment purge in accordance with SOl-30.02, Containment Purge System.

Which ONE of the following identifies...

(1) the Ice Condenser doors, opened due to pressure imbalances, will result in a MCR alarm and (2) the guidance provided in SOl-30.02 to prevent the occurrence?

A. (1) Lower Inlet Doors (2) Establish a lower containment purge for 5-10 minutes, then shutdown and start an upper containment purge.

By (1) Lower Inlet Doors (2) Establish an upper containment purge for 5-10 minutes, then shutdown and start a lower containment purge.

C. (1) Intermediate Deck Doors (2) Establish a lower containment purge for 5-10 minutes, then shutdown and start an upper containment purge.

D. (1) Intermediate Deck Doors (2) Establish an upper containment purge for 5-10 minutes, then shutdown and start a lower containment purge.

Page 175

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the lower doors are the doors that cause the alarm and the sequence listed is a sequence that could be performed but is the opposite of the sequence in the procedure.

B. Correct, 1-M-6 Annunciator 744-A ICE COND INLET DOOR OPEN will alert the

-

operators to the opening of a lower Ice Condenser Inlet Door and the SO! has Precautions identifying the potential for slight pressure variances to open the doors and the need to establish an upper purge for a few minutes, then stop, prior to establishing a lower purge in order to keep the doors from opening. See below.

C. Incorrect, Plausible because the intermediate deck doors are doors on the Ice condenser and are inspected frequently for ice buildup and the sequence listed is a sequence that could be performed but is the opposite of the sequence in the procedure.

D. Incorrect, Plausible because the intermediate deck doors are doors on the Ice condenser and are inspected frequently for ice buildup and the sequence listed is a correct process to establish a purge of lower containment.

Unit I Containment Purge System SOI-30.02 Rev. 0054 3.0 PRECAUTIONS AND LIMITATIONS E. Ice Condenser Door positions should be monitored when purging Containment. A slight upper-to-lower Containment press imbalance will open the doors.

F. For lower compartment purge, the following sequence should be followed to prevent Upper-to-Lower Containment press imbalance from opening the Ice Condenser Doors:

1. Upper compartment purge should be started first and operated for 5 to 10 minutes to permit air pressures to stabilize.

2. Stop upper compartment purge, close associated dampers and stop fans.

3. Start lower compartment purge.

Question Number: 42 Tier: 2 Group 1 KIA: 025 A2.03 025 Ice Condenser System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ice condenser system; correct, control, or mitigate the consequences of those malfunctions or operations:

Opening of ice condenser doors Importance Rating: 3.0* / 3.2*

10 CFR Part 55: 41.5 / 43.5 / 45.3 / 45.13 Page 716

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 JOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the knowledge of an operation causing opening of the ice condenser doors and how the operation can be controlled to mitigate to consequences and prevent the event.

Technical

Reference:

SOl-30/02, Containment Purge System, Revision 0056 ARI-128-144, HVAC, 480 BDS, Ice Cond, Revision 0019 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO3OC

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

3-OT-SYSO6 1 A

7. Discuss which doors in the ice condenser have position indication.

Cognitive Level:

Higher Lower X Question Source:

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

Comments:

Page 117

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

43. 026 K2.02 043 Which ONE of the following identifies the power supply to 1 -FCV-72-39, CNTMT SPRAY HDR A TO CNTMT?

A Reactor MDV Board IAI-A B. C&AVent Board JAI-A C. Reactor MDV Board 1A2-A D. C & A Vent Board 1A2-A DISTRA CTOR ANALYSIS:

A. Correct, The MOV is powered from Reactor MOV Board IA I-A.

B. Incorrect, Plausible because C & A Board 7A 7-A supplies Train A Containment Spray MOV I-FCV-72-22, RWST TO CS PMP SUCTION.

C. Incorrect, Plausible because there are Train A Containment Spray MOVs supplied from Reactor MOV Boards and Reactor Vent Board 1A2-A does supply Train A safety related equipment.

D. Incorrect, Plausible because a C & A Vent Board does supply power to a Train A Containment Spray valve and C & A Vent Board IA 2-A does supply Train A safety related equipment.

Page 118

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 43 Tier: 2 Group I K/A: 026 K2.02 Containment Spray System (CSS)

Knowledge of bus power supplies to the following:

MOVs Importance Rating: 2.7* / 2.9 IOCFRPart55: 41.7 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires applicant to demonstrate knowledge of bus power supplies to MOVs the Containment Spray System.

Technical

Reference:

1 -45W760-72-4 R9 SOI-72.01 ATT 1 P, Containment Spray System Power Checklist 72.01-IP, Revision 0032, Effective Date: 09-24-2010 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO72A

09. Identify the power supplies to the major valves of the Containment Spray System.

Cognitive Level:

Higher Lower X Question Source:

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

Comments:

Page 119

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

44. 039 A1.10 044 Given the following:

- Unit 1 is at 100% power.

- A stable 8 gpd tube leak exists on SG #3.

- 1 -RM-90-1 19, Vacuum Pump Exhaust, is reading 8 Xl 02 cpm.

- 1-RM-90-423A, SG 3 DISCH, is reading 8.08E+0 mr/hr

- An automatic BOP turbine runback occurs dropping turbine load to 970 MWe.

Following the runback, which ONE of the following identifies...

(1) the monitor used as the primary indication of the steam generator tube leak rate and (2) how the indication will be affected as a result of the change in steam flow?

Li)

A 1-RM-90-119 Trends down B. 1-RM-90-119 Trends up C. 1-RM-90-423A Trends down D. 1-RM-90-423A Trends up Page 720

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANALYSIS:

A. Correct, 1-RM-90-1 19 will respond to a much smaller amount of radiation than 1-RM-9Q-423A and the monitor reading will trend down following the runback because the steam generator pressure will be higher after the runback while the RCS pressure will remain the same. With lower deltaP the leak rate will be lower and the radiation monitor will detect a lower radiation level.

B. Incorrect, Plausible because 7-RM-90-1 19 will respond to a much smaller amount of radiation than 7-RM-9Q-423A and the applicant is likely to reverse the process of the changing pressures following the reduction in steam flow.

C. Incorrect, Plausible because 1-RM-90-423A does monitor the main steam lines, but its range is much higher than 1-RM-9Q-1 19 so it will not respond first in the event of a tube leak. Also, The monitor reading will trend down following the runback because the steam generator pressure will be higher after the runback while the RCS pressure will remain the same. With lower deltaP the leak rate will be lower and the radiation monitor will detect a lower radiation level.

D. Incorrect, Plausible because 1-RM-90-423A does monitor the main steam lines, but its range is much higher than 1-RM-90-1 19 so it will not respond first in the event of a tube leak. Also because the applicant is likely to reverse the process of the changing pressures following the reduction in steam flow.

Question Number: 44 Tier: 2 Group 1 KIA: 039A1.10 Main and Reheat Steam System (MRSS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including:

Air ejector PRM Importance Rating: 2.9*! 3.0*

JOCFRPart55: 41.5/45.5 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of how the condenser vacuum pump radiation monitor (air ejector PRM) will respond if subjected to a condition where the main steam flow rate to the turbine is suddenly reduced and how that monitor responds compared to other monitors on the main steam line.

Page 127

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

ARI-i 73-1 79, U-i Radiation Detectors, Revision 0045 AOl-33, Steam Generator Tube Leak, Revision 33 Proposed references None to be provided:

Learning Objective: 3-OT-A013300

5. Describe the means of identifying a S/G Tube Leak.

8. Given a set of plant conditions, use AOI-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 Lower X Question Source:

New Modified Bank X Bank Question History: Farley Bank Question 039 Ai.i0 001 Modified.

Comments:

Page 122

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

45. 039 K3.06 045 Given the following:

- Unit 1 reactor power is stable at 90%.

- Turbine Impulse pressure transmitter 1-PT-1-72 fails LOW.

Which ONE of the following identifes the status of the white lights on 1-M-4 for (1) 1-XI-1-103D, STM DUMPS ACT D FSVS ON and (2) 1-Xl-1-IO3AJB, STM DUMPS ARMED?

1-Xl-1-103D 1-XI-1-103A/B A. DARK DARK B DARK LIT C. LIT DARK D. LIT LIT DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because 1-Xl-703D being dark is correct and 7-Xl-1-103N8 would be dark lithe instrument had failed high.

B. Correct, 1-P 1-1-72 is the pressure transmitter that is used to sense a loss of load signal to arm the steam dumps system. When the transmitter falls low the steam dumps will arm resulting in 1-Xl-1-IO3NB being energized but 1-Xl-103D will remain dark because there would not be a larger difference in temperature between Tavg and Tref created by the transmitter failure.

C. Incorrect, Plausible because 1-XI-703D would be lit and 1-XI-7-IO3NB would be dark if the failed transmitter had been the other Impulse Pressure Transmitter 7 -P 1-1-73.

D. Incorrect, Plausible because both 7-XI-703D and 7-XI-1-IO3NB would be lit during an actual rapid load reduction and the applicant could determine both are supplied form the same transmitter.

Page 123

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 45 Tier: 2 Group 1 KIA: 039 K3.06 Main and Reheat Steam System (MRSS)

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

SDS Importance Rating: 2.8*! 3.1 10 CFR Part 55: 41.7 / 45.6 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the knowledge of how the steam dumps system will be affected by a malfunction of a pressure transmitter on the Main and Reheat Steam System.

Technical

Reference:

1-47W611-1-2 R12 SQl-i .02, Steam Dump System, Rev 0013 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO1 B

25. Given a steam dump instrument and failure mode, identify how the instrument will respond and what interlock(s) or control function(s) will be affected, including effects on system/component operation.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 06/20 1 1 WBN NRC exam Comments:

Page 724

06/2011 Wafts Bar SRO NRC License Exam 6/22/2011

46. 059K1.02 046 Given the following:

- Unit 1 operating at 25% power, with rod control in manual.

A condensate header ruptures resulting in the following:

- All Main Feedwater pumps tripped.

- 30 seconds later, SG NR levels are 23% and lowering.

- Reactor power remains at 25%.

Which ONE of the following describes the current status of...

(1) the Reactor Protection System and (2) the Auxiliary Feedwater (AFW) Pumps?

A. (1) A reactor trip setpoint has NOT been exceeded (2) AFW Pumps are NOT running.

B. (1) A reactor trip setpoint has been exceeded (2) AFW Pumps are running.

C (1) A reactor trip setpoint has NOT been exceeded (2) AFW Pumps are running.

D. (1) A reactor trip setpoint has been exceeded (2)AFW Pumps are NOT running.

Page 125

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible because the reactor trip setpoint not being exceeded is correct and the same condition that trips the reactor is the condition that automatically starts the AFW pumps on low steam generator level.

B. Incorrect, Plausible because if the SG minimum level required for heat sink is mistaken as the setpoint for the reactor trip and the AFW pumps running is correct due to being started by the trip of both TD-AFW.

C. Correct, While the steam generator levels are low they have not reached the setpoint (17% NR after a TD) that will cause a reactor trip and an AFW pump start.

However the AFW pumps would be running because they also receive a start signal from both TD-MFPs being tripped. Also, while both TD-MFPs tripping will trip the turbine, the turbine trip will not trip the reactor because the power is less than 50%.

D. Incorrect, Plausible because if the SG minimum level required for heat sink is mistaken as the setpoint for the reactor trip and if the low steam generator level AFW start is correctly recalled being 17% NR.

Question Number: 46 Tier: 2 Group 1 KIA: 059 K1.02 Main Feedwater (MFW) System Knowledge of the physical connections and/or cause effect relationships between the MFW and the following systems:

AFW system Importance Rating: 3*4* / 34 10 CFR Part 55: 41.2 to 41.9 / 45.7 to 45.8 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of how the AFW pumps are affected by a condition (all MFPS being tripped) on the MEW system prior to SG level dropping to the setpoint that will cause an AFW pump automatic start.

Technical

Reference:

1-47W611-3-3 RiO 1-47W611-3-4 R18 ARI-76-80, RX TRIP FIRST OUT, Revision 7 ARI-116-123, RPS & ESF, Revision 0010 Page 126

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO99A

17. Identify the Reactor trips and give setpoints and list logic required for the Reactor trips.

3-OT-SYSOO3B

10. Identify the A-Auto start signals of the Motor-Driven AFW 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: North Anna bank question used on their 2008 exam with changes to use at WBN.

Comments:

Page 727

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

47. 061 K5.01 047 Given the following;

- Unit 1 experienced a Reactor trip from 100% power.

- The crew completed E-0, Reactor Trip or Safety Injection, and has entered ES-0.1, Reactor Trip Response.

- Pressurizer level is 25% and slowly decreasing.

- All Steam Generator NR levels are between 21% and 25% and slowly increasing.

- All Steam Generators pressures are approximately 1060 psig and slowly decreasing.

- RCS pressure is 2020 psig and slowly decreasing.

- Tavg is 554°F and slowly decreasing.

- Steam dumps and S/G PORVs are closed.

Which ONE of the following will be the next action by the crew in accordance with ES-0.1 to address the conditions?

A. Establish Immediate Boration.

B Throttle Auxiliary Feedwater Flow.

C. Close MSIVs and bypass valves.

D. Initiate Safety Injection and Return to E-0.

Page 128

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because ES-U. I step 3 RNO directs emergency boration if coo/down drops to less than 54 7°F making this choice plausible.

B. Correct, ES-U. 1 step 3 RNO directs the throttling of AEW to address the cooldown.

C. Incorrect, Plausible because ES-U. I step 3 RNO directs the closing of MSIVs and bypass valves if the cooldown continues making this choice plausible, but this action is after the throttling of AEW flow.

D. Incorrect, Plausible because ES-U. I step I directs the initiation of SI and return to E-U but only if SI is actuated. Actuation of SI is not warranted for the stated conditions but plausible for the candidate to misuse the data trends in stem and conclude that an SI is required.

Page 729

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 47 Tier: 2 Group 1 K/A: 061 K5.01 Auxiliary/Emergency Feedwater System:

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

Relationship between AFW flow and RCS heat transfer.

Importance Rating: 3.6 I 3.9 10 CFR Part 55: 41.5 / 45.7 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the effect excess AFW flow will have on the plant following a trip by analyzing conditions stated in the question to determine that the flow needs to be reduced.

Technical

Reference:

ES-0. 1, Reactor Trip Response, Revision 0022 Proposed references None to be provided:

Learning Objective: 3-OT-EOP0000

15. Explain the purpose for and the basis of step in E-0, ES-0.0, ES-0.1, ES-0.2, ES-0.3, and ES-0.4.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN bank question 061 K5.01 047 used on the SQN 1/2008 exam. Minor wording changes and values changed to make applicable to WBN.

Comments:

Page 130

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

48. 062 A4.04 048 Given the following:

- Unit 1 is operating at 30% power.

- Following maintenance, Unit 1 480v Unit Board 1A is ready to be restored to service.

- The MCR has instructed the AUOs to energize the board using the normal supply breaker.

- Annunciator 6-D, 480V UNIT BOARD 1A UV/FAILURE/ TRANSFER, clears.

Which ONE of the following identifies indications available for the MCR operating crew to verify that the alarm cleared due to the board being energized before receiving notification from the operators in the field?

1. Voltage indicating meter on 1-M-1.

2. Normal Feed Breaker RED indicating light LIT on J-M-1.

3. GREEN light LIT on MCR handswitches for equipment fed from the board.

A. 2 ONLY B I and 2 ONLY C. 2 and 3 ONLY D. 1,2,and3 Page 131

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because the Normal Feed Breaker RED indicating light LIT on 1-M-7 is an indication that the board had been energized by closing the breaker but it is not the only one of the choices listed that will provide the indication. There is a 480v Unit board voltmeter on 1-M-1 also.

B. Correct, Voltage indicating on the 480v Unit board voltmeter on 1-M-1 and the Normal Feed Breaker RED indicating light LIT on 1-M-1 are indications that the board being energized caused the alarm to clear C. Incorrect, Plausible because, while the GREEN light LIT on MCR handswitches for equipment fed from the board will provide indication that the board has control power available, it will not indicate that the board is energized. Also plausible because the Normal Feed Breaker RED indicating light LIT on 1-M- I does provide an indication that the alarm cleared due to the board being energized.

D. Incorrect, Plausible because, while the GREEN light LIT on MCR handswitches for equipment fed from the board will provide indication that the board has control power available, it will not indicate that the board is energized. Also plausible because the other 2 of the 3 listed do provide an indication that the alarm cleared due to the board being energized.

Page 732

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 48 Tier: 2 Group 1 K/A: 062 A4.04 AC Electrical Distribution System A4 Ability to manually operate and/or monitor in the control room:

Local operation of breakers Importance Rating: 2.6 / 2.7 10 CFR Part 55: 41.7 I 45.5 /to 45.8 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the main control room indications of local breaker operations.

Technical

Reference:

1-45W760-203-1 RiO Proposed references None to be provided:

Learning Objective: 3-OT-SYSO2O3A

9. List the 480 Vsystems and describe them in terms of the following:

d. Control power

19. Locate the 480V Systems MCR Controls and Indications including:

a. Voltmeters

b. Ammeters

c. Alarms Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/20 1 1 NRC exam.

Comments:

Page 133

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

49. 063 K4.04 049 Given the following:

- Unit 1 is operating at 40% power.

- The 250v DC Turbine Building Distribution Board 1 breaker supplying the DC control powerto 6.9 kV RCP Board IA trips.

Which ONE of the following describes how the RCP #1 is affected?

A. The RCP will trip when the DC is lost resulting in an automatic reactor trip directly due to the RCP tripping.

B. The RCP will trip when the DC is lost but the reactor will not automaticaly trip directly due to the RCP tripping.

C The RCP will continue to run but the operator will not be able to trip the pump from the 1 -M-5 control switch until the control power supply is transferred to the alternate.

D. The RCP will continue to run and the operator will be able to trip the pump from the 1-M-5 control switch because the control power will automatically transfer to the alternate.

Page 134

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DI$TRA CTOR ANALYSIS:

A. Incorrect, Plausible if the normal state olf the trip coil is reversed (energized versus deenergized) or because loss of AC power would directly result in the RCP tripping and the RCP trip would lead to a reactor trip due to low flow in one loop if reactor power was greater than the P8 interlock (48%).

B. Incorrect, Plausible if the normal state olf the trip coil is reversed (energized versus deenergized) or because loss of AC power would directly result in the RCP tripping and the reactor trip due to low flow in one loop is blocked when reactor power is less than the P8 interlock (48%).

C. Correct, The RCP will continue to run because the trip coil is required to be energized in order to trip the pump. Without DC control power, the control room switch cannot energize the trip coil. Control power is required to be manually aligned to the alternate source.

D. Incorrect, Plausible because he RCP will continue to run because the trip coil is required to be energized in order to trip the pump. Without DC control power, the control room switch cannot energize the trip coil. Control power will not automatically transfer, but there are low voltage power supplies that do automatically transfer on undervoltage.

Page 135

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 49 Tier: 2 Group 1 K/A: 063 K4.04 D.C. Electrical Distribution Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following:

Trips Importance Rating: 2.6? / 2.9?

JO CFR Part 55: 41.7 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of how the DC electrical control trip circuit for a 6.9kv RCP breaker is affected by a loss of the DC control power to the breaker.

Technical

Reference:

1-45W720 R8 1-45W760-68-1 R14 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO57A

8. Identify the failure position (open, closed, or as is) of a 6.9kv or 480V Shutdown Board breaker upon loss of control power to that board.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: Diablo Canyon question from their 2007 exam modified to allow use at WBN.

Comments:

Page 136

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

50. 064 K2.03 050 Which ONE of the following correctly completes the statement below relative to the the alternate power supply to IA-A DG Battery Charger?

The alternate power supply is from a (1) which will to supply the battery charger if the normal supply is lost.

w A. Train A 480V Diesel Aux. Bd. automatically transfer B Train A 480V Diesel Aux. Bd. have to be manually aligned C. Train B 480V Diesel Aux. Bd. automatically transfer D. Train B 480V Diesel Aux. Bd. have to be manually aligned DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because it contains the correct train and there are alternate supplies on many applications that have automatic transfer schemes. But the transfer to the alternate is manual only for the DG Battery Chargers.

B. Correct, Plausible because the alternate is from the Diesel Aux Board 7A 1-A and the transfer is manual only.

C. Incorrect, Plausible because the examinee may think the alternate supply comes from the other train. The spare battery charger for the 125V DC batteries has a power supply from the opposite train( but normal is to align the supply to the spare charger to the Train it is supplying) and there are alternate supplies on many applications that have automatic transfer schemes D. Incorrect, Plausible because the examinee may think the alternate supply comes from the other train. The spare battery charger for the 125V DC batteries has a power supply from the opposite train (but normal is to align the supply to the spare charger to the Train it is supplying) and the transfer being manually aligned is correct.

Question Number: 50 Tier: 2 Group 1 Page 137

06/2011 Watts Bar SRO NRC License Exam 6/22/20 1 1 K/A: 064 K2.03 Emergency Diesel Generator (EDIG) System Knowledge of bus power supplies to the following:

Control power.

Importance Rating: 3.2* / 3.6 JOCFRPart55: 41.7 JOCFR55.43b: Not applicable K/A Match: K/A is matched because the question requires examinee to recall alternate power supply to DC lA-A battery charger (which supplies control power to DG lA-A) as well as the transfer scheme from Normal to Alternate.

Technical

Reference:

1-45W732-1 R47 1 -45W732-2 R37 SOl-82.01, Diesel Generator(DG) lA-A, Revision 0074 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO57P

1. Describe the 125v Vital, 250v, 48v and 24v battery systems in terms of the following:

U. Location and normal and alternate supplies to associated battery chargers.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question 064 K2.03 050 used on the 2006 NRC exam (written by the NRC) with minor format and wording changes Comments:

Page 738

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

51. 073 K3.01 051 Which ONE of the following Process Radiation Monitors will cause an isolation of a release path when the radiation monitor fails?

A. 1-RE-90-400, Shield Building Vent Rad Monitor B. 0-RE-90-212, Station Sump Discharge Monitor C. 0-RE-90-132, Service Building Vent Radiation Monitor D. 0-RE-90-122, Waste Disposal System Liquid Release Monitor DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible since the monitor is on a release path and actions are required lithe monitor fails.

B. Incorrect, Plausible since the monitor is on a release path and if radiation is detected a realignment of the discharge is directed; however the alignment requires manual actions.

C. Incorrect, Plausible since the monitor is on a release path and actions are required if the monitor falls.

D. Correct, This is the only radiation monitor listed which causes an automatic isolation when the monitor fails.

Question Number: 51 Tier: 2 Group 1 K/A: 073 K3.01 Process Radiation Monitoring Knowledge of the effect that a loss or malfunction of the PRM system will have on the following:

Radioactive effluent releases Importance Rating: 3.6 I 10 CFR Part 55: 41.7 I 45.6 IOCFR55.43.b: Not applicable Page 139

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: KJA is matched because the question requires knowledge of the effect an instrument malfunction (a failure) on 0-RE-90-122, Waste Disposal System Liquid Release Monitor, (a process radiation monitor) will have on an on-going release.

Technical

Reference:

ARI-1 80-1 87, Common Radiation Detectors, Revision 32 1-47W61 1-77-2, Revision 5 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO9OA

07. Determine Interlocks and/or cause-effect relationships between the Rad Monitoring Systems (ARM & Process) and the areas they monitor.

Include HVAC systems and area isolations.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question SYSO77B.10 001 with A distractor replaced with choices B, C & D re-arranged to relocate correct answer.

Comments:

Page 140

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

52. 076 A1.02 052 Given the following plant conditions:

- The plant is in MODE 4 with both trains of RHR in service.

- Spent Fuel Pool Cooling is aligned using the B Spent Fuel Pool Cooling System pump and heat exchanger.

- Four ERCW pumps, (A-A, C-A, F-B, & H-B), are in service.

- All ERCW headers are at the same pressure.

- ERCWpumpA-Atrips.

Which ONE of the following describes theimpact of the pump tripping on the listed parameters?

(Assume no operator action is taken.)

CCS temperature on Control and Station Air CCS Heat Exchanger C outlet Compressor D Oil Temperature A. Rises Rises B. Rises Remains constant C. Remains constant Rises D. Remains constant Remains constant Page 141

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because lithe ERCW pump trip had been a Train B pump which is the normal source of cooling to CCS Heat Exchanger C, the outlet would increase due to lower cooling flow. Also plausible because the D air compressor is cooled be ERCW requiring the applicant to recall the loads on the A, B and C CCS heat exchangers, and the loads on the A and B ERCW headers.

B. Incorrect, Plausible because lithe ERCW pump trip had been a Train B pump which is the normal source of cooling to CCS Heat Exchanger C, the outlet would increase due to lower cooling flow. Also, The air compressors receive water from both IA and lB ERCW supply headers with check valves in each of the flow paths.

If A-A pump tripped, the 18 ERCW Supply Header would continue to supply the compressors.

C. Incorrect, Plausible because the CCS Heat Exchanger C outlet temperature remaining constant is correct, and the Compressor D Oil Temperature rising is plausible because the D air compressor is cooled be ERCWrequiring the applicant to recall the loads on the A, B and C CCS heat exchangers, and the loads on the A and B ERCW headers.

D. Correct, CCS Heat Exchanger C outlet temperature would remain constant because it is supplied from Train B ERCW pumps and the question has only an Train A pump tripping; and the Compressor D Oil temperature also would remain constant because the air compressors receive water from both IA and 18 ERCW supply headers with check valves in each of the flow paths. If A-A pump tripped, the lB ERCW Supply Header would continue to supply the compressors.

Question Number: 52 Tier: 2 Group 1 K!A: 076 A1.02 Service Water System (SWS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including:

Reactor and turbine building closed cooling water temperatures.

Importance Rating: 2.6* / 2.6*

10 CFR Part 55: 41 .5/45.5 JOCFR55.43.b: Not applicable Page 142

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: Applicant is presented with conditions involving the loss of an Essential Raw Cooling Water pump. Understanding how parameters should respond to the pump trip demonstrates the ability to monitor system and predict the impacts on a parameter associated with a closed cooling water system used to cool components in the Auxiliary Building (Component Cooling Water) and on an additional heat load in the Turbine Building cooled by the ERCW system.

Technical

Reference:

AOI-13, Loss of Essential Raw Cooling Water, Revision 0038 1-47W845-1 R57 1 -47W845-2 R76 Proposed references None to be provided:

Learning Objective: 3-OT-SYS067A

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 Modified Bank Bank X Question History: Bank question 076 Al .02 052 used on the 05/2009 exam Comments:

Page 143

06/2071 Watts Bar SRO NRC License Exam 6/22/2011

53. 07$ A3.01 053 Plant conditions are as follows:

- The plant is operating at 100% steady state power when a toss of offsite power occurs.

- Control air pressure slowly dropped to 82 psi9 but is now recovering as a result of operator action to restart Control Air Compressors A and B.

Which ONE of the following correctly describes the condition of the Auxiliary Air Systems and the Service Air System relative to the Control Air System (CAS)?

Auxiliary Air Systems Service Air System A. Aux air compressors running Isolated from the CAS B Aux air compressors running Still being supplied from the CAS C. Aux air supplied from CAS, only Isolated from the CAS D. Aux air supplied from CAS, only Still being supplied from the CAS Page 144

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

Control air low pressure alarm at 83 psig Aux Air Compressors start at 83 psig Service air isolates at 80 psig Aux Air isolates from control air at 79.5 psig decreasing.

Control air to containment valves auto close at 70 psig A. Incorrect, Plausible because control air system pressure has dropped below the setpoint for the automatic action to start the Aux Air compressors to occur on the system but not yet low enough to result in the isolation of the service air system from the control air system.

B. Correct, the pressure has not dropped sufficiently to cause either of the systems to be isolated from the control air system but it has dropped low enough to cause the auxiliary air compressors to start.

C. Incorrect, Plausible because the auxiliary air systems has not isolated from the control system air so there can still a supply from control system available. Also, because the service air will be isolated from the control air system as the pressure continues to drop and the system pressure has dropped below the setpoint for other automatic actions on the system.

D. Incorrect. Plausible because the auxiliary air systems has not isolated from the control system air so there can still be a supply from control system available and the service air not being isolated from the control air system is correct.

Question Number: 53 Tier: 2 Group 1 KIA: 078 A3.01 Instrument Air System (lAS)

Ability to monitor automatic operation of the lAS, including:

Air pressure Importance Rating: 3.1 / 3.2 JO CFR Part 55: 41.7/45.5 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires applicant to recall both the air pressure at which aux air system and the service air system Page 145

06/2011 Wafts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

SOl-32.02, Auxiliary Air System, Revision 0020 AOl-ID, Loss of Control Air, Revision 0040 ARI-36-42, Heaters, Turb Seal & Air, Revision 0018 Proposed references None to be provided:

Learning Objective: 3-OT-AOl1000

2. Describe Auto Actions for Loss of Control Air per AOl-10 (SOER88-1, Rec2).

3-OT-SYSO32B

11. Discuss the actions that occur to auxiliary control air as the control air system pressure decreases from 106 to 60 psig.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question 078 A3.01 054 used on the 2006 exam modified with major changes to the stem conditions and choices Comments:

Page 146

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

54. 103M.01 054 Given the following plant conditions:

- The plant is in Mode 1.

- The operating crew is preparing to Place Containment Vent Air Cleanup Units (CVACU5) In Service in accordance with 501-30.03, Containment HVAC and Pressure Control.

- While the CVACUs have been out of service the conditions in containment changed as follows:

- Annulus/Vacuum AP increased from 5.7 psid to 6.1 psid.

- ContainmenUAnnulus P increased from 0.07 psid to 0.11 psid.

- Upper Containment Temperature increased from 96.3°F to 100.4°F.

- Lower Containment Temperature increased from 107.8°F to 110.2°F.

Which ONE of the following identifies the containment parameter that is required to be lowered prior to placing a CVACU in service?

A. Annulus/Vacuum AP B Containment/Annulus z\P C. Upper Containment Temperature D. Lower Containment Temperature Page 747

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible because annulus/vacuum ziP is a measured parameter that has an acceptable operating range and the annulus ziP is checked in the procedure section.

B. Correct, the containment/annulus ziP is required to be reduced to less than 7.0 psid by use of the containment purge system prior to placing the Containment Vent Air Cleanup Units (CVACUs) in service in accordance with SOI-30.03 (see below)

C. Incorrect, Plausible because the upper containment temperature is a measured parameter that has an upper limit and changes in containment air flow can affect temperatures in containment air spaces which can result in EQ concerns.

D. Incorrect, Plausible because the lower containment temperature is a measured parameter that has an upper limit and changes in containment air flow can affect temperatures in containment air spaces which can result in operational concerns and EQ concerns.

Page 148

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 54 Tier: 2 Group 1 KIA: 103A1.O1 Containment System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including:

Containment pressure, temperature, and humidity.

Importance Rating: 3.7/4.1 10 CFR Part 55: 41.5 / 45.5 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to monitor changes in containment parameters to determine acceptable conditions for operating containment system controls.

Technical

Reference:

SOI-30.03, Containment HVAC and Pressure Control, Revision 0042 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO300

11. Describe the Continuous Vent System including:

a. General description of the CVACUs

b. Inlet/Outlet flow paths

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

Cognitive Level:

Higher Lower X Question Source:

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

Page 149

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

55. 103 K4.06 055 Given the following conditions:

- Unit 1 is operating at 100% power when an automatic Safety Injection occurs.

,4c->- 2

- -Gufrent conditions are:

- Tavg 540°F and dropping.

- RCS pressure 1740 psig and dropping.

- Steam Generator pressures are 950 psig and dropping.

- Containment Pressure 2.6 psig and rising.

- All automatic actions have occurred as required.

- No manual operator actions have been taken.

Which ONE of the following identifies the current status of the three ESF actuations listed below?

Containment Phase A Isolation Containment Phase B Isolation Main Steam Line Isolation A ONLY Phase A is actuated.

B. ONLY Phase A and Phase B are actuated.

C. ONLY Phase A and Main Steam Line Isolation are actuated.

D. Phase A, Phase B, and Main Steam Line Isolation are actuated.

Page 150

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, Phase A is actuated because of the safety injection signal.

B. Incorrect, Plausible because Phase A being actuated is correct and Phase B is a containment isolation that would have occurred if the containment pressure had been higher.

C. Incorrect, Plausible because Phase A being actuated is correct and if the steam generator pressures had been lower, a Main Steam Line Isolation would have occurred.

D. Incorrect, Plausible because Phase A being actuated is correct and if the containment pressure had been higher, Phase B actuation and a Main Steam Line Isolation would have occurred.

Page 151

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 55 Tier: 2 Group 1 K/A: 103 K4.06 Containment System Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following:

Containment isolation system Importance Rating: 3.1 / 3.7 JO CFR Part 55: 41.7 JOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the conditions required to generate the different containment isolations.

Technical

Reference:

1-47W611-63-1 R13 1-47W611-88-1 R23 1-47W611-1-1 R13 Proposed references None to be provided:

Learning Objective:

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Farley bank question 103 K4.06 used on the Farley 2007 exam modified.

Comments:

Page 152

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

56. 002 K5.10 056 As Unit 1 reactor power is increased from 0% to 100%, which ONE of the following identifies how both the Reactor Coolant System (RCS) AT and the OTAT reactor trip setpoint change?

RCS 41 OTAT Reactor Trip Setpoint A. Rises by approx 29°F Increases as AT increases B. Rises by approx 29°F Decreases as AT increases C. Rises by approx 62°F Increases as AT increases D Rises by approx 62°F Decreases as AT increases DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because RCS Tavg does rise by approximately 29°F as reactor power is increased from 0% to 700% and there are reactor trip setpoints that do not change as reactor power is raised.

B. Incorrect, Plausible because RCS Tavg does rise by approximately 29°F as reactor power is increased from 0% to 100% and the OTr]T trip setpoint dropping as reactor power is raised is correct.

C. Incorrect, Plausible because RCS AT rising by approximately 62°F as reactor power is increased from 0% to 700% is correct and there are reactor trip setpoints that do not change as reactor power is raised.

D. Correct, RCS AT does rise by approximately 62°F as reactor power is increased from 0% to 100% and the OTAT reactor trip setpoint does drop as reactor power is raised.

Question Number: 56 Tier: 2 Group 2 K/A: 002 K5.10 Reactor Coolant System (RCS)

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

Relationship between reactor power and RCS differential temperature Importance Rating: 3.6 I 4.1 Page 753

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 JOCFRPart55: 41.5/45.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the relationship of reactor power to RCS differential temperature and the operational implications of how the change affects the OTAT reactor trip setpoint as power is increased.

Technical

Reference:

Tech Spec 3.3.1, page 3.3-21, Amendment 7 NOB Sheet A-i 4, AT Indicated Power vs. Actual Power, Revision 2 N3-68-4001, Reactor Coolant System, Revision 0030 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68A

12. State the normal operating values of RCS Flow, Temperature, Pressure and Pressure Drops across the Reactor Coolant System 3-OT-SYSO99A

17. Identify the Reactor trips and give setpoints and list logic requited for the Reactor trips.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/20 1 1 NRC exam.

Comments:

Page 154

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011

57. 014 K1.01 057 Given the following:

- A power ascension was in progress on Unit 1 with reactor power at 70%.

- Control Bank D (CBD) rods were at 185 steps when one CB D rod dropped into the core.

- The problem with the dropped rod has been corrected and the rod has subsequently been realigned.

- The OAC inadvertently manipulates 1-SUS, ROD CONTROL STARTUP instead of the 1-RCAR, ROD CONTROL ALARM RESET.

Which ONE of the following describes the indications or conditions that would exist due to manipulating 1-SUS, ROD CONTROL STARTUP?

A ALL control bank step counters and ALL shutdown bank would indicate 000 steps. -j i-trS B. ALL control bank step counters would indicate 000 but ALL shutdown bank step counters would indicate fully withdrawn.

C. The ROD INSERTION LOW LIMIT and the ROD INSERTION LOW LOW LIMIT alarm would be prevented from alarming.

D. The ROD INSERTION LOW LIMIT and the ROD INSERTION LOW LOW LIMIT alarm would alarm but at a LOWER value than the actual Rod Insertion Limit.

Page 155

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, Depressing the ROD CONTROL STARTUP will cause all step counters to reset to zero as well as causing several other circuits and counters to reset to zero.

B. Incorrect, Plausible because the control rods have counters that are reset from the action that the shutdown rods do not have. An example is the CEPRI circuit that emulates the P/A converter (Computer points for rod position).

C. Incorrect, Plausible because the manipulation of the switch resets several counters and circuits in the rod control and indication system but the LO and LO-LO Insertion Limit circuit is not one of them but the rod positions will be caused to indicate zero which does input to the alarm.

D. Incorrect, Plausible because the manipulation of the switch resets several counters and circuits in the rod control and indication system but the LO and LO-LO Insertion Limit circuit is not one of them but the rod positions will be caused to indicate zero which does input to the alarm.

Question Number: 57 Tier: 2 Group 2 KIA: 014 K1.01 Rod Position Indication System (RPIS)

Knowledge of the physical connections and/or cause effect relationships between the RPIS and the following systems:

CRDS Importance Rating: 3.2* I 3.6 10 CFR Part 55: 41.3 to 41.9/45.7 to 45.8 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of how an action taken in the control rod drive system affects the RPIS.

Technical

Reference:

AOl-2, Malfunction of Reactor Control System, Revision 37 SO 1-85.0 1, Control Rod Drive and Indication System, Rev. 0039 Proposed references None to be provided:

Page 756

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Learning Objective: 3-OT-SYSOO85A

20. Differentiate between the Rod Urgent Failure and Non-Urgent Failure alarms. Explain the cause and effect of the alarms and how resetting of alarms is accomplished.

22. For the rod position indicators, state the sources of signals, type of indication, and all alarms generated by each circuit.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Indian Point 3 question from the Indian Point 2007 exam.

Comments: Correct answer relocated and wording changed to allow use an WBN.

Page 157

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

58. 016 K3.02 058 Given the following:

- Unit 1 is operating steady state at 70% reactor power with rod control in manual.

Compare the effects of either one of the following RCS Loop 1 RTDs failing HIGH.

1. ThotRTD#1

2. Icold RID #1 Assuming NO operator action, which ONE of the following identifies the RCS RTD failure...

(1) having the larger effect on the pressurizer level control system and (2) how the pressurizer level would be affected?

Larciest effect Level would...

A. Thot failure rise and be controlled at a level higher than the 70% power steady state level.

B. Thot failure rise but be restored to the 70% power steady state level by the control system.

C Icold failure rise and be controlled at a level higher than the 70% power steady state level.

D. Icold failure rise but be restored to the 70% power steady state level by the control system.

Page 158

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the Thot failure would cause a change in the loop Tavg if the failed RTD was not rejected in the Tavg calculation and the change would then cause the pressurizer level control program setpoint to increase to a higher setpoint which would result in a charging flow increase to increase the pressurizer level.

B. Incorrect, Plausible because the Thot failure would cause a change in the ioop Tavg if the failed RTD was not rejected in the Tavg calculation and because it is rejected the level would return to the steady state 70% power setpoint after the failure.

C. Correct, The Tcold failure will cause the Auctioneered High RCS Tavg to signal to rise. This signal programs the pressurizer level control program setpoint resulting in an increase in the setpoint. The charging flow will increase to bring the pressurizer level up to the new setpoint.

D. Incorrect, Plausible because the Tcold failure having the largest effect is correct and because the level would return to the steady state 70% power setpoint if the failed RTD had been rejected in the Tavg calculation as is the Thot RTD failure.

Question Number: 58 Tier: 2 Group 2 KIA: 016 K3.02 Non-Nuclear Instrumentation System (NNIS)

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

PZR LCS Importance Rating: 3*4* / 35*

IOCFRPart55: 41.7/45.6 IOCFR55.43.b: Not applicable KIA Match: KJA is matched because the question requires knowledge of the effect that a loss or malfunction in the RCS temperature measuring system (NNIS) will have on the Pressurizer Level Control System.

Technical

Reference:

AOl-20, Malfunction of Pressurizer Level Control System, Revision 31 ARI-88-94, Reactor Coolant System, Revision 21 Page 159

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

AOl-20, Malfunction of Pressurizer Level Control System, Revision 31 ARI-88-94, Reactor Coolant System, Revision 21 Window 90-F N3-99-4003, Reactor Protection System, Rev. 0021 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO99B

6. Describe the process by which the Eagle 21 Process

/Protection System generates Tavg and Loop Delta T Signals.

3-OT-SYSO68C

14. Explain the basis for programming the level vs.

maintaining the level constant in the pressurizer.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 160

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

59. 017 K6O1 059 Which ONE of the following identifies...

(1) the total number of core exit thermocouples in the Unit 1 reactor and (2) how the RVLIS ICTC map page would identify a thermocouple that was greater than the critical setpoint value?

A. 58 The value would be displayed in reverse video.

B 65 The value would be displayed in reverse video.

C. 58 The value would be displayed as 9999.

D. 65 The value would be displayed as 9999.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because there are 58 thimbles for incore probes at the bottom of the reactor vessel and the thermocouple value would be displayed in reverse video when greater than the critical setpoint.

B. Correct, There are 65 core exit thermocouples and the thermocouple value would be displayed in reverse video when greater than the critical setpoint.

C. Incorrect, Plausible because there are 58 thimbles for incore probes at the bottom of the reactor vessel and there are RVLIS malfunctions that lead to 9999 being displayed, (i.e. when there are no operable core exit thermocouples in a given quadrant, RVLIS will display 9999 for the high value in that quadrant.)

D. Incorrect, Plausible because there are 65 core exit thermocouples and there are RVLIS malfunctions that lead to 9999 being displayed, (i.e. when there are no operable core exit thermocouples in a given quadrant, RVLIS will display 9999 for the high value in that quadrant.)

Page 161

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 59 Tier: 2 Group 2 KIA: 017 K6.01 In-Core Temperature Monitor System (ITM)

Knowledge of the effect of a loss or malfunction of the following ITM system components:

Sensors and detectors Importance Rating: 2.7 / 3.0 IOCFRPart55: 41.7/45.7 JOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the effect a loss of a ITM thermocouples in a core Quadrant will have on the ITM system display.

Technical

Reference:

SOl-94.01, Incore Instrumentation System, Revision 0013 TRM, B 3.3.3 Movable Incore Detectors, 09/30/95 3-OT-SYSO68F, RVLIS & ICCM, Revision 7 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO68F No objective identified but material in the lesson plan.

Cognitive Level:

Hi g her Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question SYSO94A.03 001 modified for the WBN 06/2011 NRC exam.

Comments:

Page 762

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

60. 027 K2.01 060 Which ONE of the following is the power supply to the Emergency Gas Treatment System (EGTS) Fan B?

A. C & A Vent Board 1BI-B B. C & A Vent Board 2B1-B C. Reactor Vent Board 1 B-B D. Reactor Vent Board 2B-B D1STRA CTOR ANALYSIS:

A. Correct, the EGTS Fan B is supplied with power from the C & A Vent Board 787-B.

B. Incorrect, the EGTS Fan B is supplied with power from the C & A Vent Board 787-B. Plausible because the C & A Vent Board 287-B is the power supply for similar safety related equipment including ABGTS Fan B.

C. Incorrect, the EGTS Fan B is supplied with power from the C & A Vent Board 787-B. Plausible because the Reactor Vent Board 781-B is a B train power supply for similar safety related equipment such as containment upper compartment cooling fans.

D. Incorrect, the EGTS Fan B is supplied with power from the C & A Vent Board 181-B. Plausible because the Reactor Vent Board 287-B is a B train power supply and Unit 2 boards do supply safety related equipment required for Unit loperation.

Page 163

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 60 Tier: 2 Group 2 KIA: 027 K2.01 Containment Iodine Removal System (CIRS)

Knowledge of bus power supplies to the following:

Fans Importance Rating: 3.1* / 34*

10 CFR Part 55: 41.7 JOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the knowledge of the station in-house electrical board that supplies power to one of the fans used for containment iodine removal.

Technical

Reference:

SOI-65.02, Emergency Gas Treatment System, Revision 0027 1 -45W756-5 R62 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO65A

9. Describe the EGTS fans power supplies.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: SQN bank question 027 K2.01 061 used on SQN 1/2009 exam Comments:

Page 764

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

61. 033 A1.01 061 Given the following:

- Unit 1 is operating at 100% power.

- Fuel shuffles are in progress in the Spent Fuel Pool (SFP) with the Spent Fuel Pit Gate installed.

- Due to a failure of the annunciator 128-A SEP LEVEL HI/LO, the level lowers to El 748 10 before the condition is identified.

- No leakage from the pit is identified.

- The Shift Manager has directed a makeup to the SEP in accordance with SOl-78.01, Spent Fuel Pool Cooling And Cleaning System.

Which ONE of the following completes the two statements below?

SEP level is (1) the minimum required by Technical Specifications to allow fuel movement.

Makeup to the SEP (2)

L)

A. below can be initiated immediately B. below must be coordinated with Chemistry and Reactor Engineering C above can be initiated immediately D. above must be coordinated with Chemistry and Reactor Engineering Page 765

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the level is below the indicating level on the spent fuel pool level gauge and because the make-up method not being required to be coordinated with Chemistry and Reactor Engineering is correct and makeup can be started immediately.

B. Incorrect, Plausible because the level is below the indicating level on the spent fuel pool level gauge and if the level had been 1 inches lower, SOI-78.O1, Spent Fuel Cooling System, would have required the makeup method to be coordinated with Chemistry and Reactor Engineering.

C. Correct, the stated level is only 2.5 inches below the normal operating range and there is a nominal 26 feet above the fuel at the normal operating range level. So there is still between 25 and 26 feet of water above the top of the fuel. This is more than the minimum of 23 feet required by Tech Spec. Also since the level is above 748 feet and 9 inches, the make-up method is not required to be coordinated with Chemistry and Reactor Engineering and can be started immediately.

D. Incorrect, Plausible because the level being above the minimum of 23 feet required by Tech Spec is correct and if the level had been 1 inches lower, SOl-78.Q1, Spent Fuel Cooling System, would required the makeup method to be coordinated with Chemistry and Reactor Engineering.

Question Number: 61 Tier: 2 Group 2 KIA: 033A1.01 Spent Fuel Pool Cooling System (SFPCS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including:

Spent fuel pool water level Importance Rating: 2.7 / 3.3 IOCFRPart55: 41.5/45.5 JOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires predicting if Tech Spec limits have been exceeded following level change due to a loss of water from the pit and the plant coordination required to restore the level to normal.

Page 766

06/2011 Watts Bar SRO NRC License Exam 6/22/20 1 1 Technical

Reference:

SOI-78.01, Spent Fuel Pool Cooling And Cleaning System, Revision 0061 Tech Spec 3.7.13 Fuel Storage Pool Water Level FSAR 9.1.2.2 47W200-8 R8 Proposed references None to be provided:

Learning Objective: 3-OT-T/S 0307

3. Given plant conditions and parameters, correctly determine the OPERABILITY of components associated with different Plant Systems in Section 7 of Technical Specifications.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Bank question 033A1 .01 from Vogtle 2009 exam (31) modified for use at WBN. Changed stem conditions, second part of the questions, made different answer correct and changed the format of the question.

Comments:

Page 767

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

62. 034 A2.01 062 Given the following:

- Unit 1 is in Mode 6 with core off-load in progress.

- 1 -HS-90-41 0-A [back of I -R-73J and I -HS-90-41 5-B [back of 1 -R-78J are in the REFUEL position.

- A fuel assembly being withdrawn into the Refueling Mast is dropped on to the lower core plate.

- The following annunciator goes into alarm:

Window 174A CNTMT PURGE EXH 1-RM-130/131 RAD HI

-

- Both MASTER ISOL SIGNAL STATUS PNLs, 1-XX-55-6C and 1-XX-55-6C, indicate CVI actuation has occurred.

- AOl-29, Dropped Or Damaged Fuel or Refueling Cavity Seal Failure, is entered.

Which ONE of the following identifies...

(1) how the Fuel Handling Area Exhaust fans are affected and (2) the action requited in AOl-29 relative to the upper and lower containment air locks?

A. (1) Fuel Handling Area Exhaust fans will be automatically stopped.

(2) Both doors must be closed in both Upper and Lower containment air locks.

B (1) Fuel Handling Area Exhaust fans will be automatically stopped.

(2) At least one door must be closed in both the Upper and Lower containment air locks.

C. (1) Fuel Handling Area Exhaust fans will NOT automatically stop but will be required to be manually shutdown.

(2) Both doors must be closed in both Upper and Lower containment air locks.

D. (1) Fuel Handling Area Exhaust fans will NOT automatically stop but will be required to be manualty shutdown.

(2) At least one door must be closed in both the Upper and Lower containment air locks.

Page 768

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because the Fuel Handling Area Exhaust fans being automatically stopped is correct and normally both doors are required to be closed when containment closure is established.

B. Correct, With 7-HS-90-4 70-A and 7-HS-90-4 75-B in the REFUEL position a CVI signal will also initiate a signal to isolate the Auxiliary Building which causes the Fuel Handling Area Exhaust fans to be tripped (with the switches not in the refuel position, the CVI would not isolate the Aux Building). AOI-29 section 3.3 has a caution prior to the first step stating Maintenance must be notified to IMMEDIATELY ensure at least one door is closed on both upper and lower personnel airlocks,...

C. Incorrect, Plausible because the Fuel Handling Area Exhaust fans are not normally stopped by a CVI signal (only stopped when 7-HS-90-4 10-A and 7-HS-90-4 75-B are in the REFUEL position) and normally both doors are required to be closed when containment closure is established.

D. Incorrect, Plausible because the Fuel Handling Area Exhaust fans are not normally stopped by a CVI signal (only stopped when 7-HS-90-4 70-A and 7-HS-90-4 75-B are in the REFUEL position) and at least one door being required to be closed is correct.

Question Number: 62 Tier: 2 Group 2 K/A: 034 A2.01 Fuel Handling Equipment System (FHES)

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

Dropped fuel element Importance Rating: 3.6 / 4.4 10 CFR Part 55: 41.5 / 43.5/45.3 / 45.13 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires predicting the impacts to the Fuel Handling Area Exhaust fans due to conditions existing as a result of a dropped fuel assembly and then actions required when using the procedure to control the event.

Page 169

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

AOl-29, Dropped Or Damaged Fuel or Refueling Cavity Seal Failure, Revision 0021 Proposed references None to be provided:

Learning Objective: 3-CT-AD 12900

4. Identify possible Auto Actions from:

a. Dropped fuel assembly in SFP area.

b. Dropped fuel assembly in Cntmt.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: SQN Bank question 103 A2.04 054, used on SQN Feb 2010 exam, modified for use on the WBN 06/2011 exam.

Comments:

Page 770

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

63. 035 A4.06 063 Given the following:

- A safety valve fails open on S/G #1 resulting in a reactor trip and safety injection on Unit 1.

- When the MSIV5 handswitches are placed to closed, the #1 MSIV failed to close.

Which ONE of the following identifies...

(1) the earliest an isolation of AFW flow to S/G #1 is allowed in accordance with 11-12.04, Users Guide For Abnormal And Emergency Operating Instructions, and (2) the status of the SG #1 MSIV indicating lights on 1-M-4 as the AUOs perform E-2,Faulted Steam Generator Isolation, Attachment I in attempt to close S/G #1 MSIV?

A. (1) After the immediate operator actions of E-0 are complete.

(2) Lights will be available to indicate status both after the MSIV transfer control switch is placed to AUX and after the 1 25v control power fuses are removed.

B. (1) After the immediate operator actions of E-0 are complete.

(2) Lights will be available to indicate status after the MSIV transfer control switch is placed to AUX but will be lost when the 125v control power fuses are removed.

C (1) After the steps of E-0 are complete through the verification of heat sink and minimum heat sink is ensured for the unaffected S/Gs.

(2) Lights will be available to indicate status both after the MSIV transfer control switch is placed to AUX and after the I 25v control power fuses are removed.

D. (1) After the steps of E-0 are complete through the verification of heat sink and the minimum heat sink is ensured for the unaffected S/Gs.

(2) Lights will be available to indicate status after the MSIV transfer control switch is placed to AUX but will be lost when the 125v control power fuses are removed.

Page 771

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the requirements for immediate action steps are different then the requirements for the remainder of the steps in a procedure and because the indicating lights being available after the fuses are removed is correct.

B. Incorrect, Plausible because the requirements for immediate action steps are different then the requirements for the remainder of the steps in a procedure and because in most circuits the indicating lights are in the same circuit as the control power fuses.

C. Correct, Tl-72.04 states the early isolation of AEW is allowed to a faulted S/G providing the steps of E-O are complete through verification of heat sink and the heat sink minimum requirements are ensured to be met for the unaffected SIGs. (See below) The position indicating lights for SG #1 will remain available to provide indication of position after both the control transfer switch is placed to AUX and the 725v control fuses are pulled in accordance with the E-2 attachment. There is a separate 125v control circuit for the indicating lights.

D. Incorrect, Plausible because the isolation of AFW being allowed after the steps of E-Q are complete through verification of heat sink and the heat sink minimum requirements are ensured to be met for the unaffected S/Gs is correct. Also, plausible that the indicating lights would not be available after the control fuses were removed because in most circuits the indicating lights are in the same circuit as the control power fuses.

TI-i 2.04 2.7 Prudent Operator Actions page 33

9. Early isolation of auxiliary feedwater (AFW) should be performed after positive identification of a ruptured, faulted, or ruptured and faulted S/G, providing the following guidelines are met:

a. Prior to isolation, the steps of E-0 are complete through verification of heat sink.

b. For a ruptured S/G, ensure that narrow range (NR) level on the affected S/G is

>32% and heat sink minimum requirements are met for the unaffected S/Gs.

c. For a faulted or ruptured and faulted S/G, ensure heat sink minimum requirements are met for the unaffected SIGs.

Question Number: 63 Tier: 2 Group 2 KIA: 035 A4.06 Steam Generator System (S/GS)

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

S/G isolation on steam leak or tube rupture/leak Importance Rating: 4.5 / 4.6 Page 772

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 10 CFR Part 55: 41.7 / 45.5 to 45.8 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires ability to monitor a faulted S/G condition to know when components can be manually isolated and how actions taken while isolating the S/G will affect the ability to monitor the S/G status.

Technical

Reference:

E-2, Faulted Steam Generator Isolation, Revision 0012 TI-i 2.04, Users Guide For Abnormal And Emergency Operating Instructions, Revision 0009 i-45W600-i-5 R8 i-47W600-i-6 R4 Proposed references None to be provided:

Learning Objective: 3-OT-EOPO200

5. Given a set of plant conditions, use procedure E-2 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions.

3-OT-TI 1204

13. Apply the rules of usage which relate to performing steps of an EOP in a specified sequence to determine when steps may/must be performed.

3-OT-SYSOO 1 A

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

a. MSIVs and bypasses.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 173

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

64. 071 K4.04 064 Which ONE of the choices below completes the following statement relating to operation of the Waste Gas Decay Tanks?

The in-service Waste Gas Decay Tank will automatically isolate when its pressure rises to (1) and the tank aligned for standby will (2)

Cl)

A. 100 psig be placed in service manually B. 135 psig be placed in service manually C 100 psig be automatically placed in service D. 135 psig be automatically placed in service DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the in-service WGDT does automatically isolate on increasing pressure at 100 psig and standby WGDT is placed in service, but it will occur automatically and not require manual action. However, manual action is required to select another tank for standby operation.

B. Incorrect, Plausible because the alarm for tank high pressure occurs at 135 psig and manual operator action is required to select a new standby tank but not to place the standby tank in service.

C. Correct, The in-service tank WGDT does automatically isolate on increasing pressure at 100 psig and standby tank WGDT is placed in service automatically when the in service WGDT isolates.

D. Incorrect, Plausible because the alarm for tank high pressure occurs at 135 psig and the standby tank being place in service automatically is correct.

Page 174

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 64 Tier: 2 Group 2 KIA: 071 K4.04 Waste Gas Disposal System (WGDS)

Knowledge of design feature(s) and/or interlock(s) which provide for the following:

Isolation of waste gas release tanks Importance Rating: 2.9 / 3.4 JOCFRPart55: 41.7 I OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of an interlock that will automatically isolate the in-service Waste Gas Decay Tank (Waste Gas Release Tank) and a design feature with associated with the tank aligned for stand-by operation.

Technical

Reference:

ARI-0-L-2C, Waste Gas Panel, Revision 0005 AOl-77.02, Waste Gas System, Revision 0035 Proposed references None to be provided:

Learning Objective: 3-OT-SYS 077B

7. Describe the gas decay tanks, include number and maximum pressure.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN Bank question SYSO77B.07 002 modified for use on the 06/2011 NRC exam.

Comments:

Page 175

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

65. 079 G2.1.30 065 Given the following:

- Unit I is in Mode 3 following a manual reactor trip required due to a control air line break in the turbine building.

- The operating crew performed the applicable emergency instructions and has stabilized the plant.

- The crew has implemented AOl-JO, Loss of Control Air, to address the loss of air.

Assuming all systems respond as designed, which ONE of the following identifies a local action that could be required in Auxiliary Building during the performance of AOl-i 0 due to the loss of air?

A. Control Steam Generator PORVs using nitrogen bottles stationed on El 729 and El 737.

B. Close Control-to-Aux Air header isolation valves using local controls on El 757.

C. Stop the air pumps for containment radiation monitors using breakers on the front of the monitors on El 737.

D Adjust Reactor Coolant Pump seal injection flow using in-service seal water filter outlet valve on El 713.

Page 176

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRA CTOR ANAL YSIS:

A. Incorrect, Plausible because controlling the SG PORV locally would be required if the loss had been a loss of Essential Control Air (Auxiliary Air) instead of a loss of control air in the turbine building. See AOl-b, Section 3.2 B. Incorrect, Plausible because the AOl does direct local action to close the valves, but this action would only be required if the automatic closure failed and the question states that all plant equipment functions as designed. See A 01-70, Section 3.3 C. Incorrect, Plausible because stopping the air pumps locally at the radiation monitors would be required if the loss had been a loss of Essential Control Air (Auxiliary Air) instead of a loss of control air in the turbine building. See AOl-b, Section 3.2.

D. Correct, The AOl directs the in-service seal water filter outlet valve 7-IS V-62-549 or 7-IS V-62-550 to be adjusted to supply seal water at greater than 6 gpm to each RCP. See AOl-b, Section 3.3 Page 177

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 65 Tier: 2 Group 2 KIA: 079 G2.1.30 Station Air System (SAS)

Ability to locate and operate components, including local controls.

Importance Rating: 4 .4/ 4.0 IOCFRPart55: 41.7/45.7 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the knowledge of the components requiring operation from local controls and the ability to locate the controls.

Technical

Reference:

AOl-b, Loss of Control Air, Revision 0040 SQl-i .01, Main Steam System, Revision 0040 Proposed references None to be provided:

Learning Objective: 3-OT-AOlb000

5. Explain the actions to control PZR level if Non-Essential Air is lost in Mode 3.

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

a. Recognize Entry Conditions

b. Identify Required Actions

c. Respond to Contingencies (RNO)

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/20 1 1 NRC exam.

Comments:

Page 778

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

66. G2.1.13 066 In accordance with OPDP-1, Conduct of Operations, which ONE of the following identifies...

(1) the individual(s), other than shift operations personnel, that are allowed access to the MCR without first requesting permission and (2) an individual who controls access for entry into the horseshoe area on Unit 1?

Li)

ANRConly OAC B. NRC only Shift Manager C. NRCandQA OAC D. NRC and QA Shift Manager Page 179

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Correct, As stated in OPDP-7, Conduct of Operations, identifies (in Section 4.5 A & D, see below) that With the exception of NRC and shift Operations personnel, all others are to request permission to enter the main control room and that Access to the horseshoe area will be strictly controlled by the US or OA IC...

B. Incorrect, Plausible because NRC being able to enter without requesting permission is correct and the Shift Manger is identified as 7 of the 2 individuals that control access to the Control room (but not I of the 2 individuals controlling access to the horseshoe area)

C. Incorrect, Plausible because NRC being able to enter without requesting permission and while QA is an oversight organization that will be granted permission when requesting, they are not excepted from requesting permission before entering. Also, because the OAC being I of the 2 individuals identified as controlling access to the horseshoe area is correct.

D. Incorrect, Plausible because NRC being able to enter without requesting permission and while QA is an oversight organization that will be granted permission when requesting, they are not excepted from requesting permission before entering. Also, the Shift Manger is identified as I of the 2 individuals that control access to the Control room (but not I of the 2 individuals controlling access to the horseshoe area)

OPDP-7 A. The SM and US control access to the Control Room and they have the authority to clear the Control Room of unnecessary personnel not supporting operations. With the exception of NRC and shift Operations personnel, all others are to request permission to enter the main control room.

D. The number of non-shift personnel in the control room surveillance area (past the swinging gate) will be limited to only those necessary to perform authorized tasks or communicate with the operating crew. Access to the horseshoe area will be strictly controlled by the US or OATC and entry made only by individuals directly involved in the conduct or monitoring of a work activity in the area. Pre-job briefs shall not be conducted in the horseshoe area. Reactivity management briefs, emergency/

abnormal operations briefs from the US or crew updates from crew members, are permitted.

Page 180

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 66 Tier: 3 Group n/a KIA: G2.1.13 Conduct of Operation Knowledge of facility requirements for controlling vital/controlled access.

Importance Rating: 2.5 / 3.2 10 CFRPart55: 41.10/43.5/45.9/45.10 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge facility requirements to gain access to the main control room(MCR) and the exceptions to the requirement as well as the operations positions assigned responsibility for managing and maintaining control of specific controlled areas in the MCR. This question expands the knowledge beyond GET level (and into the operational level) pertaining to the requirements for controlling access to vital/controlled areas.

Technical

Reference:

OPDP-1, Conduct of Operations, Revision 0019 Proposed references None to be provided:

Learning Objective: 3-OT-SPP1000

3. Describe Unit Supervisor responsibilities.

5. Describe Unit Operator responsibilities.

Cognitive Level:

Higher Lower X Question Source:

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

Comments:

Page 781

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

67. G2.1.3$ 067 Given the following:

- While performing E-3, Steam Generator Tube Rupture, the Unit Supervisor (US) determines that a transition to ECA-3.1, SGTR and LOCA Subcooled Recovery, is required.

-

- The US gets the crews attention and announces the procedure number and title.

Which ONE of the following identifies...

(1) the action required in accordance with TI-i 2.04, Users Guide for Abnormal and Emergency Operating Instructions, that must be met prior to initiating steps in the procedure and (2) the OAC5 minimum requirement in accordance with OPDP-1, Conduct of Operations, when the US reads a step directing the manipulation of a control switch?

A. (1) The US must conduct a crew briefing prior to implementing steps.

(2) Provide repeat back prior to taking the action, and report back after the operation is completed.

B. (1) The US must conduct a crew briefing prior to implementing steps.

(2) Provide repeat back and operate only after confirmation is received from the sender.

C. (1) The US must read the purpose for the procedure to the crew.

(2) Provide repeat back prior to taking the action, and report back after the operation is completed.

D (1) The US must read the purpose for the procedure to the crew.

(2) Provide repeat back and operate only after confirmation is received from the sender.

Page 182

06/2011 Watts Bat SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because conducting a Crew Brief at procedure transition is a good practice but not the required action and the communication as listed is a three way communication but not the required three-way process.

B. Incorrect, Plausible because conducting a Crew Brief at procedure transition is a good practice but not the required action and the communication as listed is a correct three way communication as required by OPDP-7.

C. Incorrect, Plausible because reading the purpose of the ECA being required is correct and the communication as listed is a three way communication but not the required three-way process.

D. Correct, When transition is made into Emergency Contingency Actions (ECA),

TI-12. 04 requires the Procedure Reader to get the attention of the crew and read the purpose of that procedure. The operators will NOT take actions at this time.

This ensures crew understanding of where they are headed. OPDP-1 requires the operator to receive confirmation of the direction prior to taking an action.

Question Number: 67 Tier: 3 Group n/a KIA: G2.i.38 Conduct of Operations Knowledge of the stations requirements for verbal communications when implementing procedures.

Importance Rating: 37* / 3.8 IOCFRPart55: 41.10/45.13 IOCFR55.43.b: Not applicable KIA Match: Question requires knowledge of the stations requirements for verbal communications when implementing procedures.

Technical

Reference:

TI-i 2.04, Users Guide for Abnormal and Emergency Operating Instructions, Revision 0009 OPDP-i, Conduct of Operations, Revision 0019 Proposed references None to be provided:

Learning Objective: 3-OT-SPP1 000 Page 783

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Learning Objective: 3-OT-SPP1000

20. Describe the responsibilities of operations personnel as it pertains to Back To Basics fundamentals, including:

h. Communications 3-OT-TI 1204

6. Identify the special requirement of the procedure reader when a transition is made to an ECA.

Cognitive Level:

Higher Lower X Question Source:

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

Comments:

Page 784

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

68. G2.1.44 06$

Given the following:

- Unit 1 is in Mode 6 with the lift of the upper internals package in progress.

- Annunciator 81 B - SOURCE RANGE HI FLUX AT SHUTDOWN alarms.

- OAC observes count rates rising on both source range monitors.

Which ONE of the following identifies the response of the containment evacuation alarm and required action in accordance with the ARI?

The containment evacuation alarm will A. NOT be automatically actuated but an announcement to evacuate containment is required.

B. NOT be automatically actuated and an announcement to evacuate containment is only required if both of the SRM count rates double.

C be automatically actuated and action is required to ensure containment is evacuated.

D. be automatically actuated but if neither SRM count rate has doubled, the alarm is to be reset and announced as expected.

DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the containment evacuation alarm can be blocked during different sets of conditions on the plant and a required action is to ensure containment is evacuated. Making a plant announcement on the PA system is an effective action to ensure individuals are aware of the need for the evacuation.

B. Incorrect, Plausible because the containment evacuation alarm can be blocked during different sets of conditions on the plant and the SRM count rate doubling is a condition during refueling that requires action.

C. Correct, The containment evacuation alarm will be automatically actuated and action is required to be taken to ensure containment is evacuated.

D. Incorrect, Plausible because the containment evacuation alarm being automatically actuated is correct and the SRM count rate doubling is a condition during refueling that requires action.

Question Number: 68 Page 785

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Tier: 3 Group n/a KIA: G2.1.44 Conduct of Operations Knowledge of RD duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Importance Rating: 3.9 / 3.8 IOCFRPart55: 41.10/43.7/45.12 IOCFR55.43.b: Not applicable KIA Match: KA is matched because the question requires knowledge of alarms generated from refueling activities as how the instrumentation affects system alarms and also the actions and communiciations required by the ARI due to the alarm.

Technical

Reference:

ARI-81-87, NIS & Rod Controls, Revision 0033 GO-7, Refueling Operations, Revision 0032 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO92A

26. Explain the operation of the Gamma-Metrics Shutdown Monitor including the 1/M circuit/display and the High Flux at Shutdown alarm circuit.

3-OT-G00700

1. Identify the reason for each prerequisite and precaution as discussed in this lesson or provided in GO-7.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 786

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Page 187

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

69. G2.2.13 069 Given the following:

- MDAFWP 1 B-B is tagged for work on the pump discharge check valve, 1-CKV-3-821.

- 1-PCV-3-132, MDAFWP lB-B Discharge Pressure Control Valve, is to be used as a boundary isolation valve and has been closed.

In addition to the valve handwheel, which ONE of the following identifies the minimum local positioning and tag placement requirements for 1-PCV-3-132 in accordance with NPG-SPP-1 0.2, Clearance Procedure to Safely Control Energy?

A. The air supply valve verified to be open and then tagged in the open position.

B The air supply valve closed and tagged, and air regulator depressurized.

C. A jacking device installed on the valve and tagged, along with the air supply valve tagged in the open position.

D. A jacking device installed on the valve and tagged, along with the air regulator depressurized, and the air supply valve closed and tagged.

DIS TRA CTOR ANALYSIS:

A. Incorrect, Plausible because if the valve had been a fall open valve the air supply would hold the valve closed.

B. Correct, The applicant must have knowledge of which way the valve falls and because it falls closed, SPP-1O.2, Clearance Procedure to Safely Control Energy, Appendix E requires that the valves air supply be isolated, depressurized and tagged.

C. Incorrect, Plausible because if the valve was determined to fall open, the requirement would be to install a blocking device and the air supply bring open would hold the valve closed.

D. Plausible because lithe valve was determined to fall open,the minimum requirement would be to install a blocking device and tag the air supply valve in the closed position.

Page 788

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 69 Tier: 3 Group n/a KIA: G2.2.13 Equipment Control Knowledge of tagging and clearance procedures.

Importance Rating: 4.1 / 4.3 IOCFRPart55: 41.10/45.13 IOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires knowledge of the clearance procedure and tagging associated with an air operated valve used as a boundary isolation valve.

Technical

Reference:

NPG-SPP-10.2, SPP-10.2, Clearance Procedure to Safely Control Energy, Revision 0001 1-47W611-3-3 RiO Proposed references None to be provided:

Learning Objective: 00059195

16. Identify special requirements for plant mechanical equipment clearances Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question SPP1002.16 004 modified.

Comments:

Page 189

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

70. G 2.2.4 1 070 Given the following:

- Unit I is in Mode 5.

- To support an alignment in the plant, the operating crew restored power to 1-FCV-74-9-B, RHR SYSTEM ISOL BYPASS, and places 1-HS-74-9 to CLOSE.

- 1 second after the handswitch was placed to close, one of the fuses in the MOVs control power circuit blew.

Using the print provided, which ONE of the choices below completes the two following statements?

The valve travel movement would be immediately stopped because the (1)

Both of the valve position indicating lights on the 1-M-6 would be (2)

REFERENCE PROVIDED A. (1) breaker on the MDV board trips.

(2) LIT B. (1) breaker on the MDV board trips.

(2) DARK C (1) contactor inside the breaker compartment opens.

(2) LIT D. (1) contactor inside the breaker compartment opens.

(2) DARK Page 790

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the breaker does have a fuse protection device that will trip the breaker open, but the device is not there to protect the control power fuses, it is there to protect the line side fuses and the indicating light being LIT is correct.

B. Incorrect, Plausible because the breaker does have a fuse protection device that will trip the breaker open, but the device is not there to protect the control power fuses, it is there to protect the line side fuses and in most circuits the indicating light position is powered from the same control power circuit as the contactor.

C. Correct, the contactor is not a latching contactor. If the control power is lost the contactor will de-energize resulting in the contactor opening and the power supply to the indicating lights is from a separate control power supply, so the light or lights would be LIT.

D. Incorrect, Plausible because the control power being lost due to the fuse failure will result in the contactor being de-energized and in most circuits the indicating light position is powered from the same control power circuit as the contactor.

Page 197

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 70 Tier: 3 Group n/a KIA: G2.2.41 Equipment Control Ability to obtain and interpret station electrical and mechanical drawings.

Importance Rating: 3.5 I 3.9 IOCFRPart55: 41.10/45.12/45.13 JOCFR55.43.b: Not applicable KIA Match: K/A is matched because the question requires the ability to interpret station electrical drawings as to how a blown fuse would affect the travel of an MOV and the MCR indications for the valve position Technical

Reference:

1 -45W760-74-4, Ri 3 Proposed references 1 -45W760-74-4 to be provided:

Learning Objective: No objective identified Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 192

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

71. G2.3.12 071 Which ONE of the following identifies the incore flux detector placement and tagging requirements listed in Tl-12.07A, Containment Access Modes I 4, for an entry into

lower containment or the annulus?

Required Incore Flux Detector Placement Taqqed with a...

A Storage position or inserted Hold Order to within 10 feet of the core B. Storage position only Hold Order C Storage position or inserted CautiUn Order to within 10 feet of the core D. Storage position only Caution Order Page 193

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Correct, In accordance with Tl-72.07, the incore flux detectors must be verified to be in the storage position or inserted to within 70 feet of the core and tagged out. A Hold Order is used to maintain the configuration control. See below B. Incorrect, Storage is not the only position allowed for the incore flux detectors, they can also be inserted to within 10 feet of the core. A Hold Order is used to maintain the configuration control. Plausible because storage is one of the two approved positions and tagging with a Hold Order is correct.

C. Incorrect, the incore flux detectors must be verified to be in the storage position or inserted to within 10 feet of the core and tagged out. A Caution Order cannot be used to maintain the configuration control. Plausible because the two locations are approved positions and a Caution Order is one of the types of clearances used for tagging equipment.

D. Incorrect, Storage is not the only position allowed for the incore flux detectors, they can also be inserted to within 70 feet of the core. A Caution Order cannot be used to maintain the configuration control. Plausible because storage is one of the two approved positions and a Caution Order is one of the types of clearances used for tagging equipment.

TI-I 2.07 3.0 PRECAUTIONS AND LIMITATIONS 3.1 General B. Prior to entry into Upper or Lower Containment or the Annulus the incore flux detectors shall be verified to be in the storage position or inserted to within ten feet of the core and the system TAGGED. The RP Superintendent! Designee and the Shift Manager may approve an exception to this requirement provided that adequate controls are established to prevent personnel overexposure.

3.2.3 Operations Shift Manager (SM)/SRO A. Ensure incore flux detectors in storage position or inserted to within ten (10) feet of the reactor core. A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore flux detector drive motors; otherwise, SM must evaluate necessity of entry and coordinate entry with RADIATION PROTECTION and applicable maintenance section.5,6 Page 194

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 71 Tier: 3 Group n/a KIA: G2.3.12 Radiation Control Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance Rating: 3.2 / 3.7 10 CFR Part 55: 41.12 /45.9 / 45.10 IOCFR55.42.b: Not applicable K/A Match: K/A is matched because the question requires knowledge that the Incore flux detectors pose radiation hazards and must be properly positioned and controlled to protect personnel inside lower containment.

Technical

Reference:

TI-i 2.07A, Containment Access Modes 1 - 4, Revision 0003 Proposed references None to be provided:

Learning Objective: 3-OT-T11207

12. Discuss the precaution associated specifically to an entry into the Annulus and Lower Containment.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question T11207 006 with choices in different order.

Comments:

Page 195

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

72. G 2.3.7 072 Given the following:

- While releasing a clearance, a clearance card on a HPFP system valve inside the Auxiliary Building is to be removed and the valve is to be opened.

- Valve is located 8 feet above floor level requiring a portable ladder to access.

Which ONE of the following identifies...

(1) if any additional Radiation Protection support is needed to access the valve and (2) the required verification technique in accordance with NPG-SPP-J0.3, Verification Program?

Li)

A. No additional support required Concurrent Verification B. No additional support required Independent Verification C. Additional support required Concurrent Verification D Additional support required Independent Verification Page 796

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible if applicant determines the area can be accessed without RadPro support, however any area over 6 ft requires RadPro support. Concurrent Verification is not the correct method of verifying the position of a throttled valve.

B. Incorrect, Plausible if applicant determines the area can be accessed without RadPro support, however any area over 6 ft requires RadPro support. Also Independent Verification is the correct method of verifying the position of throttled valves.

C. Incorrect, Plausible since RadPro support is required for access to the area but Concurrent Verification is not the correct method of verifying the position of throttled valves.

D. Correct, RadPro support is required for access to elevations of 6 ft and greater.

Independent Verification is the correct method of verifying the position of a throttled valve.

Question Number: 72 Tier: 3 Group n/a K/A: G 2.3.7 Radiation Control Ability to comply with radiation work permit requirements during normal or abnormal conditions.

Importance Rating: 3.5 I 3.6 IOCFRPart55: 41.12/45.10 I OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to comply with the height limitation on standing RWPs and is operationally valid because it requires knowledge of the verification requirements when positioning plant equipment.

Technical

Reference:

NPG-SPP-10.3, Verification Program, Revision 0000 RCI-152, Radiological Postings, 0006 Proposed references None to be provided:

Learning Objective: 3-OT-RAD0003 Page 197

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Learning Objective: 3-OT-RAD0003

8. Identify the responsibilities of the following concerning the ALARA program:

c. Employee 3-OT-SPP1 003

6. Explain the verification requirements for those systems listed in appendix A of SPP-10.3 Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: SQN bank question G 2.3.7 072 used of the SQN 1/2009 RETAKE exam modified by changing conditions in the stem to make a different answer correct.

Comments:

Page 198

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

73. G 2.4.28 073 Given the following:

- Unit 1 is operating at 100% power.

- AOl-42.01, Security Events, Section 3.1.3, INFORMATION of Potential Threat of Aircraft Attack, is entered.

In accordance with AOI-42.01, which ONE of the choices below completes both of the following statements?

AOl-42.01 will direct the MCR to be staffed by (1)

When AOl-42.01 directs the Diesel Fire Pump to be started, (2)

A. (1)one RO and one SRO (2) an AUO must be sent to locally start the pump B. (1) two ROs, two SROs, and the STA (2) an AUO must be sent to locally start the pump C. (1) two ROs, two SROs, and the STA (2) the pump can be started from the Main Control Room D (1)one RO and one SRO (2) the pump can be started from the Main Control Room DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the staffing of one RO and one SRO is corrrect, and the Diesel Fire Pump is normally started by an AUO locally.

B. Incorrect, Plausible because the staffing listed is the normal staffing which would be on hand during an emergency and the Diesel Fire Pump is normally started by an AUO locally.

C. Incorrect, Plausible because the staffing listed is the normal staffing which would be on hand during an emergency and the pump is started from the control room during performance of AOl-42. 01.

D. Correct, The staffing of one RO and one SRO is correct per AOl-42. 01 and the Diesel Fire Pump is started from the control room during performance of AOI-42.01.

Page 199

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 73 Tier: 3 Group n/a KIA: G 2.4.28 Emergency Procedures /Plan Knowledge of procedures relating to a security event (non-safeguards information).

Importance Rating: 3.2 / 4.1 IOCFRPart55: 41.10/43.5/45.13 JOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the applicant to to have knowledge of procedural requirements for control room staffing and plant operation, specifically fire pump operation, during a security event.

Technical

Reference:

AOl-42.01, Security Events, Revision 0017 Proposed references None to be provided:

Learning Objective: 3-OT-A014200

4. Discuss the contingencies associated with an Imminent Aircraft Attack.

Cognitive Level:

Higher Lower X Question Source:

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

Comments:

Page 200

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

74. G 2.4.42 074 Which ONE of the following identifies...

(1) the Emergency Response Organization responsible for dispatching Emergency Response Teams during a plant emergency and (2) the lowest level emergency that will result in the emergency center being required to be staffed?

A. Technical Support Center Alert B. Technical Support Center NOUE C. Operations Support Center Alert D. Operations Support Center NOUE DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because the TSC performs many of the actions normally performed by Main Control Room staff during emergency operations and Alert is correct for the lowest level emergency that requires staffing of the emergency centers.

B. Incorrect, Plausible because the TSC performs many of the actions normally performed by Main Control Room staff during emergency operations and an NOUE is the lowest emergency classification level.

C. Correct, The OSC is the organization responsible for dispatching emergency response teams during an emergency and Alert is correct for the lowest level emergency that requires staffing of the emergency centers.

D. Incorrect, Plausible because the OSC is the organization responsible for dispatching emergency response teams during an emergency and an NOUE is the lowest emergency classification level.

Question Number: 74 Tier: 3 Group n/a KIA: G2.4.42 Emergency Procedures IPIan Page 207

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 K/A: G2.4.42 Emergency Procedures !Plan Knowledge of emergency response facilities.

Importance Rating: 2.6 I 3.8 JO CFR Part 55: 41.10 / 45.11 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the questions requires knowledge of the functions of the different emergency response centers and when the centers are required to be staffed.

Technical

Reference:

EPIP-7, Activation and Operation of the Operations Support Center (OS C), Revision 0030 EPIP-3, Alert, Revision 0033 Proposed references None to be provided:

Learning Objective: 3-OT-PDC-048C

3. Identify the functions of the onsite emergency response facilities.

8. Recognize how AUOs are dispatched and controlled during radiological emergencies.

16. Recognize conditions which constitute activation of the emergency response facilities regardless of the time of day when an emergency has been declared.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question PCDO48C.03 002 modified by using only 2 of the choices and adding a second part to the question.

Comments:

Page 202

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

75. G 2.449 075 Given the following:

- An ATWS has occurred on Unit 1.

- When the operator places 1-HS-47-24, TURBINE TRIP, to the TRIP position the turbine trip buses fail to actuate.

Which ONE of the following identifies...

(1) the indicating light that will be LIT on 1-HS-47-24 and (2) the first action the operator is directed to take due to the turbine trip failure in accordance with FR-S.1, Nuclear Power Generation! ATWS?

A. RED Close MSIVs and Bypasses B RED Manually runback turbine C. GREEN Close MSIVs and Bypasses D. GREEN Manually runback turbine Page 203

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the RED indicating light being LIT is correct and closing the MSIVs and bypass valves is the next action to be taken lithe manual turbine runback is unsuccessful. (See ER-S. I step 2 below)

B. Correct, The RED indicating light will be LIT lithe turbine falls to trip following actuation of the handswitch and the action directed in the RNO states IF turbine will NOT trip THEN Manually runback turbine. (See FR-S. I step 2 below)

C. Incorrect, Plausible because of the light configuration on the handswitch. (See picture) The handswitch is turned to the right to trip and the RED light is on the right but the GREEN indicating light would be LIT lithe turbine did trip and closing the MSIVs and bypass valves is the next action to be taken if the manual turbine runback is unsuccessful. (See FR-S. I step 2 below)

D. Incorrect, Plausible because of the light configuration on the handswitch. (See picture) The handswitch is turned to the right to trip and the RED light is on the right but the GREEN indicating light would be LIT lithe turbine did trip and action to manually runback the turbine is correct. (See FR-S. 1 step 2 below)

FR-S.1 Step 2

2. ENSURE Turbine Trip: Manually TRIP turbine.

A11 turbine stop valves CLOSED. IF turbine will NOT trip, THEN Manually runback turbine.

IF turbine can NOT be run back, THEN CLOSE MSIVs and bypasses Page 204

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 75 Tier: 3 Group n/a KIA: G 2.4.49 Emergency Procedures I Plan Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Importance Rating: 4.6 I 4.4 IOCFRPart55: 41.10/43.2/45.6 IOCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to perform without reference to procedures the immediate actions of required for a failure of the turbine to trip during an ATWS event.

Technical

Reference:

FR-S.1, Nuclear Power GenerationlATWS, Revision 0020 Proposed references None to be provided:

Learning Objective: 3-OT-FRS000 1

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

Cognitive Level:

Higher Lower X Question Source:

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

Comments:

Page 205

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

76. 015 AG2.4.$ 076 Given the following conditions:

- Unit I is operating at 8% power.

- Annunciator window 100-D, RCP SEAL LEAK OFF FLOW HI, alarms.

- The DAC reports RCP #4 seal leakoff flow is off-scale high and the RCP lower bearing temperature is rising.

- In accordance with AOI-24, RCP Malfunction During Pump Operation, Section 3.2, RCP Tripped or Shutdown Required, the reactor is tripped.

- The crew enters E-0, Reactor Trip or Safety Injection.

Which ONE of the following identifies...

(1) the required implementation of AOI-24 after the EOPs are entered in accordance with Tl-12.04, Users Guide for Abnormal and Emergency Operating Instructions, and (2) the maximum time allowed by AOI-24 to close RCP #4 Seal Return Valve, 1-FCV-62-22 after the pump is stopped?

A. (1) Continued during the performance of E-0.

(2) 3 minutes B. (1) Continued during the performance of E-0.

(2) 5 minutes C. (1) Continued ONLY after a transition is made to ES-0.1, Reactor Trip Response.

(2) 3 minutes Db (1) Continued ONLY after a transition is made to ES-0.1, Reactor Trip Response.

(2) 5 minutes Page 206

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because A 01-24 being continued concurrently E-O would support cornpletion of the action required in a short time period (3-5 minutes) and 3 minutes does appear in the procedure associated with isolating the seal leakoff but it is the minimum time allowed, not the maximum time allowed.

B. Incorrect, Plausible because A 01-24 being continued concurrently E-0 would support cornpletion of the action required in a short time period (3-5 minutes) and 5 minutes being the maximum time allowed to isolate the seal return valve is correct.

C. Incorrect, Plausible because A 01-24 being continued concurrently with EOP performance after transition is made from E-0 is correct and 3 minutes does appear in the procedure associated with isolating the seal leakoff but it is the minimum time allowed, not the maximum time allowed.

D. Correct, In accordance with 11-72.04, when an AOl in effect directs a Reactor Trip, then the performance of the AOl should continue immediately following transition to ES-0. 7. In this question the transition from E-0 is to ES-0. 1, so continuing A 01-24 concurrently with the EOPs would be after the transition from E-0 was made. Also A 01-24 direct the seal leakoff to be isolated with 3 to 5 minutes after the RCP is stopped to protect the pump, so 5 minutes is the maximum time.

Question Number: 76 Tier: 1 Group 1 KIA: 015 AG2.4.8 Reactor Coolant Pump Malfunctions Emergency Procedures / Plan Knowledge of how abnormal operating procedures are used in conjunction with EQ Ps.

Importance Rating: 3.8 I 4.5 JOCFRPart55: 41.10/43.5/45.13 JOCFR55.43.b: 5 K/A Match: KJA is matched because the question requires knowledge of how abnormal operating procedures are used in conjunction with EOPs and is SRO because it requires knowledge of an administrative procedure to specifies hierarchy, implementation and coordination of plant abnormal and emergency procedures, as well as the content of the procedure (max time allowed) versus the overall mitigating strategy.

Page 207

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

AOI-24, RCP Malfunction During Pump Operation, Revision 0029 Tl-12.04, Users Guide for Abnormal and Emergency Operating Instructions, Revision 0008 Proposed references None to be provided:

Learning Objective: 3-OT-TI 1204

00. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with this procedure that are rated 2.5 during Initial License training and 3.0 during License Operator Requalification training for the appropriate license position as identified in Appendix A.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the WBN 06/20 1 1 NRC exam Comments:

Page 208

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011

77. 029 EA2.05 077 Given the following:

- Unit 1 is operating at 100% power when a Safety Injection occurs.

- The reactor did NOT trip and can NOT be tripped manually from the MCR.

- The crew has implemented FR-S.1, Nuclear Power Generation/ATWS, and has completed performance of step 4 to borate the RCS.

Which ONE of the following identifies:

(1) the expected indication on 1-M-5 for 1-Xl-62-1228A & B, VCT VENT OUTLET ISOL VALVE and (2) when E-0, Reactor Trip or Safety Injection, Appendix A, Equipment Verifications, will be first performed?

A. (1) RED indicating light LIT.

(2) ONLY after the transition is made to E-0.

B. (1) RED indicating light LIT.

(2) Time permitting, during the performance of FR-S.1.

C. (1) GREEN indicating light LIT.

(2) ONLY after the transition is made to E-0.

D (1) GREEN indicating light LIT.

(2) Time permitting, during the performance of FR-S.1.

Page 209

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the VCT VENT OUTLET ISOL VALVE normally has the RED light lit and the valve is not positioned directly during the performance of FR-S. 1 but does reposition due to the change in charging pump suction alignment. Also plausible because normally the initiation of the Appendix is performed when directed by E-O.

B. Incorrect, Plausible because the VCT VENT OUTLET ISOL VALVE normally has the RED light lit and the valve is not positioned directly during the performance of FR-S. I but does reposition due to the change in charging pump suction alignment. Also the initiation of Appendix A being implemented while performing the actions of FR-S. 7 is correct.

C. Incorrect, Plausible because for the conditions stated, the VCT VENT OUTLET ISOL VALVE indication would be GREEN as a result of the SI causing the charging pump suction valves to reposition automatically. Also plausible because normally the initiation of the Appendix is performed when directed by E-Q.

D. Correct, For the conditions stated the SI would cause the suction of the charging pumps to transfer to the VCT. When this transfer occurs a resulting action is for the VCT VENT OUTLET ISOL VALVE to reposition from open to close. Thus the GREEN light would be lit. Due to the auto SI (because of low RCS pressure), as directed by step 8 of FR-S. 1, if SI is actuated then E-O steps 1-6 (which includes the performance of E-O, Appendix A) are to be implemented while performing the actions of FR-S. 1.

Question Number: 77 Tier: 1 Group 1 KIA: 029 EA2.05 Anticipated Transient Without Scram (ATWS)

Ability to determine or interpret the following as they apply to a ATWS:

System component valve position indications.

Importance Rating: 3.4 / 3.4 10 CFR Part 55: 43.5 / 45.13 IOCFR55.43.b: 5 KIA Match: This question matches the K/A by having the candidate determine the correct indicated position of valves in the CVCS following completion of the RCS boration step in the ATWS procedure and is SRO because it requires by knowledge of when to implement attachments including how to coordinate the attached with procedure Page 210

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 K/A Match: This question matches the K/A by having the candidate determine the correct indicated position of valves in the CVCS following completion of the RCS boration step in the ATWS procedure and is SRO because it requires by knowledge of when to implement attachments including how to coordinate the attached with procedure steps (determining when E-0 Appendix A, Equipment Verifications, would be performed while performing steps in FR-Si)

Technical

Reference:

FR-S.1, Nuclear Power Generation/ATWS, Revision 0020 E-0, Reactor Trip or Safety Injection, Revision 29 Proposed references None to be provided:

Learning Objective: 3-OT-FRS0001

1. Apply the rules of usage (TI-i 2.04) and analyze plant conditions to identify any required procedure transitions in FR-S.i and FR-S.2.

9. Given a set of plant conditions, use FR-Si, FR-S.2 and the Critical Safety Function Status Trees to correctly diagnose and implement: Action Steps, RNOs, Notes and Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: SQN bank question 029 EA2.06 076 modified forWBN 06-201i NRC exam.

Comments:

Page 211

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

78. 038 EG2.4.46 072 Given the following:

- A SGTR has occurred on Unit 1. The crew has initiated an RCS cooldown to the target temperature.

- The following annunciators are in alarm:

90-B - PZR PRESS LO-DEVN BACKUP HTRS ON 92-A - PZR LEVEL HI/LO 92-C - PZR LEVEL LO-HTRS OFF & LTDN CLOSED

- Pressurizer level is 13% and rapidly dropping.

- Pressurizer pressure is 1470 psig and dropping.

- RCS subcooling is 54°F and rising.

Which ONE of the following describes (1) the procedure required, and (2) the status of the Reactor Coolant Pumps?

A (1) Remain in E-3, Steam Generator Tube Rupture.

(2) RCPs will continue to run.

B. (1) Remain in E-3, Steam Generator Tube Rupture.

(2) RCPs will be secured.

C. (1) Transition to ECA-3.1, SGTR and LOCA Subcooled Recovery.

-

(2) RCPs will continue to run.

D. (1) Transition to ECA-3.1, SGTR and LOCA Subcooled Recovery.

-

(2) RCPs will be secured.

Page 272

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANALYSIS:

A. Correct, the rapid RCS cooldown will result in the pressurizer level dropping. The conditions and other alarms are consistent with the effects of the cooldown. E-3 provides direction that the 7500 psig RCP trip criteria is not applicable during or after the cooldown to target incore temperature.

B. Incorrect, Plausible because the alarms and conditions being consistent with the rapid RCS cooldown in E-3 is correct and the RCS pressure is less than the 7500 psig RCP trip criteria that is applicable in E-3 prior to the initiation of the RCS cooldown.

C. Incorrect, Plausible because the transition to ECA-3. 7 is on the foldout page and would be required with the pressurizer level less than 75% after SI Termination.

Also, the 7500 psig RCP trip criteria is not applicable in E-3 during or after the cooldown to target incore temperature.

D. Incorrect, Plausible because the transition to ECA-3. I is on the foldout page and would be required with the pressurizer level less than 15% after SI Termination.

Also, the RCS pressure is less than the 7500 psig RCP trip criteria that is applicable in E-3 prior to the initiation of the RCS cooldown.

Question Number: 78 Tier: 1 Group 1 KIA: 038 EG2.4.46 Steam Generator Tube Rupture Emergency Procedures / Plan Ability to verify that the alarms are consistent with the plant conditions.

Importance Rating: 4.2 / 4.2 10 CFR Part 55: 41.10/43.5/45.3/45.12 IOCFR55.43.b: 5 KIA Match: K/A is matched and the question is SRO because the question requires the ability select the actions to take based on alarms providedand the procedure with which to proceed from conditions and the decision points in EOPs that involve transitions to event specfic emergency contingency procedure. (assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.)

Page 213

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

E-3, Steam Generator Tube Rupture, Revision 23 ARI-88-94, Reactor Coolant System, Revision 19 Proposed references None to be provided:

Learning Objective: 3-OT-EOPO300

5. Given a set of plant conditions, use E-3, ES-3. 1, ES-3.2, and ES-3.3 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 EOPO300 001 modified for WBN 06/2011 NRC exam.

Comments:

Page 274

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

79. 055 EG2.4.1$ 079 Given the following:

- ECA-0.0, Loss of All AC Power, is in effect with the depressurization of the intact S/Gs in progress.

- An RCS cooldown was established at a cooldown rate of 120°F/hr.

- During performance of the depressurization, Shutdown Board IA-A is restored.

- When determining the applicable recovery instruction to implement the DAC reports the following parameters exist:

- Containment pressure has risen to 0.4 psid.

- RCS subcooling is 68°F.

- Pressurizer level is off scale low.

- RCS pressure is 470 psig and stable.

Which ONE of the following identifies...

(1) the basis for allowing the cooldown to exceed the normal maximum rate during the depressurization step and (2) the required recovery procedure to be implemented when transitioning from ECA-0.0?

A (1) Protect the RCP seals.

(2) ECA-0.2, Recovery from Loss of Shutdown Power With SI Required B. (1) Protect the RCP seals.

(2) ECA-0.1, Recovery from Loss of Shutdown Power Without SI Required C. (1) Conserve secondary inventory.

(2) ECA-0.2, Recovery from Loss of Shutdown Power With SI Required D. (1) Conserve secondary inventory.

(2) ECA-0.1, Recovery from Loss of Shutdown Power Without SI Required Page 215

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, in accordance with the WOG ECA-0.0 background document protecting the RCP seals to reduce the RCS inventory loss is a primary concern. Reducing or delaying seal damage is accomplished by reducing the differential pressure and the temperature to which the seals are subjected by depressurizing and cooling the RCS by secondary side depressurization and cooling. ECA-0. 0 states that the 700°F/hr cooldown rate limit does not apply and if pressurizer level is less than the required 15% when ready to transition, the required recovery procedure is ECA-0.2.

B. Incorrect, Plausible because protecting the RCP seals is correct and if the pressurizer level had been greater than 15% when ready to transition, ECA-0. I would have been correct because the other conditions including minimum subcooling are met. Plausible to conclude pressurizer level is not required because the procedure identifies that the level could go off scale low and should be anticipated. If so the procedure directs the action to depressurize to be continued.

C. Incorrect, Plausible because cooling faster would conserve secondary inventory and without power there is an inability to makeup to the CSTs in order to restore the inventory. Also, when ready to transition, ECA-0.2 being the correct procedure to implement is correct because pressurizer level is not equal to or above the required 15% minimum.

D. Plausible because cooling faster would conserve secondary inventory and, without power, there is an inability to makeup to the CSTs in order to restore the inventory.

Also, if the pressurizer level had been greater than 15% when ready to transition, ECA-0. I would have been correct because the other conditions including minimum subcooling are met. Plausible to conclude pressurizer level is not required because the procedure identifies that the level could go off scale low and should be anticipated. If so the procedure directs the action to depressurize to be continued.

Question Number: 79 Tier: 1 Group 1 KIA: 055 EG2.4.18 Loss of Offsite and Onsite Power (Station Blackout)

Emergency Procedures I Plan Knowledge of the specific bases for EOPs.

Importance Rating: 3.3 / 4.0 IOCFR Part 55: 41.10/43.1 /45.13 IOCFR55.43.b: 5 K/A Match: K/A is matched because the question requires knowledge of the Page 216

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: K/A is matched because the question requires knowledge of the basis for actions in ECA-0.0 and is SRO because it requires an assessment of plant conditions to allow the selection of the procedure with which to proceed to mitigate and recover the plant.(Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.)

Technical

Reference:

ECA-0.0, Loss Of Shutdown Power, Revision 21 WOG ECA-0.0 Background, Revision 2 Proposed references None to be provided:

Learning Objective: 3-OT-ECA0000

03. Explain why the intact SIG are depressurized to 300 psig during performance of ECA-0.0.

08. Given a set of plant conditions, use ECA-0.0, ECA-0.1 and ECA-0.2 to correctly diagnose and implement: Action Steps, RNOs, Notes and Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN question ECA0000.08 002 modified Comments:

Page 217

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

80. 077 AA2.04 080 Unit 1 Generator is operating at 90% reactor power steady state conditions due to maintenance on the generator hydrogen seal oil system:

- Megawatts 1100 MWe

- Generator Voltage 23.6 Ky

- Hydrogen Pressure 45 psig

- All 500kv switchyard lines are in service.

- Reactive load 0 Mvars

- System Frequency 60.00 Hertz A disturbance on the 500kV electrical system results in the following:

- Megawatts 1100 MWe

- Generator Voltage 23.6 Ky

- Hydrogen Pressure 45 psig

- Generator voltage regulator tripped to MANUAL.

- An additional +240 Mvars of reactive load is being applied to the generator.

- System Frequency 60.10 Hertz Attempts to contact the System Load Coordinator have failed.

Which ONE of the following identifies...

(1) the status of the Generator Capability Curve Limits and (2) the procedure required to be implemented?

REFERENCE PROVIDED A. Capability Curve Limits are being violated.

GO-4, Normal Power Operation, Section 5.5, Frequency Variation

Response

B Capability Curve Limits are being violated.

1-Pl-OPS-1-MCR, Main Control Room, Section 5.3, Voltage Control Monitoring C. Capability Curve Limits are NOT being violated but administrative MVAR limits are being violated.

GO-4, Normal Power Operation, Section 5.5, Frequency Variation

Response

D. Capability Curve Limits are NOT being violated but administrative MVAR limits are being violated.

1-Pl-OPS-1-MCR, Main Control Room, Section 5.3, Voltage Control Monitoring Page 218

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the generator operating outside the limits of the Generator Capability Curve is correct and if frequncy had been greater than 61. 15, hertz then dropping load in accordance with GO-4, Normal Power Operation, Section 5.5, Frequency Variation Response would have been correct.

B. Correct, the conditions result in an outgoing reactive load of 240 MVARs. With 7700 MWe load at 45 psig hydrogen pressure the generator is operating outside the limits of the Generator Capability Curve and the procedure required to be implemetned is 7-Pl-OPS-1-MCR, Main Control Room, Section 5.3, Voltage Control Monitoring to reduce voltage thus reducing megavar loading to get the unit back within the limits of the capability curve.

C. Incorrect, Plausible because if the reactive loading change is applied in the opposite (incoming) direction the generator would be within the limits of the Generator Capability Curve but in excess of the administrative limit of lOOM VARs maximum incoming and if frequncy had been greater than 67.75, hertz then dropping load in accordance with GO-4, Normal Power Operation, Section 5.5, Frequency Variation Response would have been correct.

D. Incorrect, Plausible because if the reactive loading change is applied in the opposite (incoming) direction, the generator would be within the limits of the Generator Capability Curve but in excess of administrative limits of 700 MVARs incoming. Also, because 1-PI-OPS-7-MCR, Main Control Room, Section 5.3, Voltage Control Monitoring to reduce voltage thus reducing megavar loading to get the unit back within the limits of the capability curve is the correct procedrue implementation.

Question Number: 80 Tier: 1 Group 1 KIA: 077 AA2.04 Generator Voltage and Electric Grid Disturbances Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

VARs outside the capability curve Importance Rating: 3.6 / 3.6 10 CFR Part 55: 41.5 and 43.5 / 45.5, 45.7, and 45.8 IOCFR55.43.b: 7 KIA Match: K/A is matched because the question requires the ability to apply Page 219

06/2011 Watts Bar SRD NRC License Exam 6/22/2011 K/A Match: K/A is matched because the question requires the ability to apply conditions to the generator capability curve to determine if VARs are outside the limits and is SRO because it requires assessing plant conditions and determining the correct procedure and section required to be implemented due to existing plant conditions.

Technical

Reference:

1-Pl-OPS-1-MCR, Main Control Room, Revision 0054 SO 1-47.02, Turbo-Generator Startup Operation, Revision 0069 GO-4, Normal Power Operation, Revision 0060 Proposed references SO 1-47.02, Turbo-Generator Startup Operation, to be provided: Appendix E, Generator Capability Curve Learning Objective: 3-OT-GOO400

5. Identify plant parameters to be monitored periodically during either a load increase from 30% to 100%

Reactor power or load decrease from 100% to 30%

Reactor power. (SER 91-024)

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for WBN 06/201 1 NRC exam.

Comments:

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06/2011 Watts Bar SRO NRC License Exam 6/22/2011

81. wiii EA2.2 081 Given the following:

- A LOCA has occurred on Unit 1.

- ECA-1 .1, Loss of RHR Sump Recirculation, is in effect.

- Containment pressure rises and the Containment Critical Safety Function turns ORANGE.

Which ONE of the following identifies the proper procedure usage and operation of the Containment Spray Pumps?

A. Remain in ECA-1 .1, and direct the operator to operate the Containment Spray Pumps as described in FR-Z.1, High Containment Pressure.

B. Remain in ECA-1 .1, and direct the operator to operate the Containment Spray Pumps as described by ECA-1 .1.

C. Transition to FR-Z.1, High Containment Pressure, and direct the operators to operate Containment Spray Pumps as described by FR-Z.1.

D Transition to FR-Z.1, High Containment Pressure, but direct the operators to operate Containment Spray Pumps as described by ECA-1 .1.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because there is another instruction (ES-1.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from ES-7.3) and FR-Z. I does provide for the direct operation of the Containment Spray Pumps.

B. Incorrect, Plausible because there is another instruction (ES-I.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from ES-I.3) and the Containment Spray Pumps are required to be operated in accordance with ECA-I. I as identified in both ECA-1. I and FR-Z. 7.

C. Incorrect, Plausible because the transition to FR-Z. us required due to the ORANGE path, and FR-Z. I does provide for the direct operation of the Containment Spray Pumps.

D. Correct, The transition to FR-Z. I is required due to the ORANGE path, but the Containment Spray Pumps are required to be operated in accordance with ECA-7. I as identified in both ECA-l. I and FR-Z. I.

Page 221

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 81 Tier: 1 Group 2 K/A: W/E1 1 EA2.2 Loss of Emergency Coolant Recirculation Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Importance Rating: 3.4 / 4.2 JO CFR Part 55: 43.5 / 45.1 IOCFR55.43.b: 5 K/A Match: K/A is matched because the question requires interpreting the conditions and adhering to the appropriate conditions within the emergency procedures which are required by the facilitys license.

SRO because the question requires knowledge of the content of the procedures versus knowledge of the overall mitigative strategy or purpose as well as the assessment of plant conditions, then selecting the procedure with which to proceed.

Technical

Reference:

ECA-1 .1, Loss of RHR Sump Recirculation, Revision 0012 FR-Z.1, High Containment Pressure, Revision 0011 Proposed references None to be provided:

Learning Objective: 3-OT-FRZ000 1

2. Discuss the reasons that ECA-1.1, Loss of RHR Sump Recirculation, is given priority over FR-Z.1, High Containment Pressure for directing Containment Spray Operation.

3-OT-ECAO1 01

08. Given a set of plant conditions, use procedures ECA-1 .1 and 1.2 to identify any required procedure transition.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Page 222

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question History: Diablo Canyon bank question used on their 2009 exam with wording changes to make applicable to WBN. One distractor replaced.

Comments:

Page 223

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

82. 005 AA2.04 082 Given the following:

- Unit 1 is operating at 42% power.

- Reactor core power conditions result in the operating crew suspecting a rod is partially dropped into the core.

In accordance with TR 3.3.9, Power Distribution Monitoring System (PDMS),

which ONE of the following identifies...

(1) the minimum number of incore thermocouple required for the PDMS to be operable and (2) ii the PDMS is required to be operable when being used to determine the position of a dropped rod?

A. (1) 8 total with 2 thermocouples in each quadrant (2) is required to be operable B. (1) 8 total with 2 thermocouples in each quadrant (2) is NOT required to be operable C (1) 17 total with 2 thermocouples in each quadrant (2) is required to be operable D. (1) 17 total with 2 thermocouples in each quadrant (2) is NOT required to be operable Page 224

06/2011 Watts Bar SRD NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because a total of 8 thermocouples with a minimum of 2 in each quadrant is required by LCO 3.3.3, Post Accident Monitoring and TR 3.3.9 requiring the PDMS to be operable when being used to determine the position of a dropped rod is correct.

B. Plausible because a total of 8 thermocouples with a minimum of 2 in each quadrant is required by LCO 3.3.3, Post accident Monitoring and the TR for the Moveable Incore Detectors (TR 3.3.3) does not contain the requirement for operability when using the detectors to determine the position of a rod as TR 3.3.9 does. Both of these systems can be used to determine the rod postion. Also, the PDMS is required to be operable when being used to determine several other core conditions related to I/S limits on reactor core as is TR 3.3.3 for the Moveable Incore Detectors.

C. Correct, The minimum number of thermocouples required to be operable in accordance with SR 3.3.3.2 is 17 total with a minimum of 2 in each quadrant and TR 3.3.9 requires the PDMS to be operable when being used to determine the position of a dropped rod.

D. Incorrect, Plausible because a total of 17 thermocouple with a minimum of 2 in each quadrant being required is correct and the TR for the Moveable lncore Detectors (TR 3.3.3) does not contain the requirement for operability when using the detectors to determine the position of a rod as TR 3.3.9 does. Both of these systems can be used to determine the rod postion. Also, the PDMS is required to be operable when being used to determine several other core conditions related to T/S limits on reactor core as is TR 3.3.3 for the Moveable Incore Detectors.

Question Number: 82 Tier: 1 Group 2 K/A: 005 AA2.04 Inoperable/Stuck Control Rod Ability to determine and interpret the following as they apply to the Inoperable / Stuck Control Rod:

Interpretation of computer in-core TC map for dropped rod location Importance Rating: 2.3* / 3.4 JO CFR Part 55: 43.5 / 45.13 IOCFR55.43.b: 2 K/A Match: K/A is matched because the question requires the knowledge of in-core TC status required for the PDMS (computer incore model) to Page 225

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: K/A is matched because the question requires the knowledge of in-core TC status required for the PDMS (computer incore model) to be operable to allow its use to determine dropped rod status and is SRO because it requires knowledge of the information contained below the line in Tech Specs (minimum requirements for TCs from table)

Technical

Reference:

Tech Requirement 3.3.9, Power Distribution Monitoring System (PDMS), Revision 46 09/20/2010 Tech Spec 3.3.3 Post Accident Monitoring Tech Requirement 3.3.9, Moveable Incore Detectors, 9/30/1995 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO85A

26. Discuss applicable Technical Specifications, Technical Requirements, and Bases.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History:

Comments:

Page 226

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

83. 032 AG2.2.25 083 Given the following:

- A reactor startup is in progress with the reactor critical at 2.0 X105%

power.

- Source Range Monitor, 1-Nl-92-131, fails LOW.

Which of the following identifies...

(1) how the failure affects the plant start-up and (2) a Technical Specification basis for the source range instruments relative to the current mode of operation?

A. (1) Startup can NOT continue until the SRM is restored to an operable status.

(2) To ensure a high flux at shutdown alarm in reponse to a rod ejection event.

B. (1) Startup can continue provided power is raised to greater than P-6 within 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />.

(2) To ensure a high flux at shutdown alarm in reponse to a rod ejection event.

C (1) Startup can NOT continue until the SRM is restored to an operable status.

(2) To ensure a high flux trip in response to boron dilution events.

D. (1) Startup can continue provided power is raised to greater than P-6 within 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />.

(2) To ensure a high flux trip in response to boron dilution events.

Page 227

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because the startup not being able to be continued is correct and while the SRMs are the monitors that are the source of the high flux at shutdown alarm, the applicable T/S basis is not to ensure the alarm in response to an ejected rod event.

B. Incorrect, Plausible because there is a condition with an IRM failure associated with P-6 and a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time limit but the action is not associated with an SRM failure and while the SRMs are the monitors that are the source of the high flux at shutdown alarm, the applicable T/S basis is not to ensure the alarm in response to an ejected rod event.

C. Correct, In accordance with A 01-4, Nuclear Instrumentation Malfunctions, and Tech Spec 3.3. 1, with the plant in Mode 2, the startup cannot continue until both SR Nis become operable. Also, in accordance with Tech Spec basis, the source range high flux trip is designed to protect the plant from control rod withdrawal from subcritical, boron dilution and control rod ejection events.

D. Incorrect, Plausible because there is a condition with an IRM failure associated with P-6 and a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time limit but the action is not associated with an SRM failure and a basis for the SRMs being to protect the plant from a boron dilution event is correct Question Number: 83 Tier: 1 Group 2 K/A: 032 AG2.2.25 Loss of Source Range Nuclear Instrumentation Equipment Control Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Importance Rating: 3.2 / 4.2 JO CFR Part 55: 41.5 / 41.7 / 43.2 IOCFR55.43.b: 2 K/A Match: K/A is matched because the question requires knowledge of the bases in Technical Specifications for limiting conditions for operations if a Source Range Instrument fails during a reactor/plant startup as well as the action required due to the failureand while the Tech spec action is an immediate action, the question is SRO based on requiring the knowledge of Tech Spec Bases. (Knowledge of TS bases that are required to analyze TS required actions and Page 228

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

Tech Spec 3.3.1, RTS Instrumentation, Amendment 52 Tech Spec 3.3.1, RTS Instrumentation Basis Tech Spec 3.9.3, Nuclear Instrumentation Bases AOl-4, Nuclear Instrumentation Malfunctions, Revision 0029 Proposed references None to be provided:

Learning Objective: 3-OT-AO 10400

3. Describe Operator Actions for an SRM failure.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 229

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

84. W/E08 EG2.1.25 084 Given the following:

- During the RCS rapid cooldown in accordance with E-3, Steam Generator Tube Rupture, the following conditions exist.

- RCPs have been stopped.

- Ruptured loop Tcold is 21 5°F.

- Lowest Intact Loop Tcold is 403°F.

- RCS pressure is 1250 psig.

Based on the above conditions, which ONE of the following identifies the status of the Pressurized Thermal Shock and the required procedure action?

REFERENCE PROVIDED A. A RED path exists. Transition to FR-P.1, Pressurized Thermal Shock, immediately.

B. A RED path exists. Remain in E-3 until the note prior to the step for determining if Containment Spray can be stopped.

C. An ORANGE path exists. Transition to FR-P.J, Pressurized Thermal Shock, immediately.

D An ORANGE path exists. Remain in E-3 until the note prior to the step for determining if Containment Spray can be stopped.

Page 230

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANALYSIS:

A. Incorrect, Plausible because lithe Limit A Curve is misread or misapplied, a RED path to FR-P. 1 could be identified and normally an immediate transition is required due to a RED path being identified.

B. Incorrect, Plausible because if the Limit A Curve is misread or misapplied, a RED path to FR-P. 1 could be identified and remaining in E-3 until the note prior to the step for determining if containment spray pumps can be stopped without making the transition even though the RED path is present is correct.

C. Incorrect, Plausible because the Limit A Curve identifying an ORANGE path to FR-P. us correct and normally an immediate transition is required due to an ORANGE path being identified.

D. Correct, The condition indicates an Orange path exists on the ruptured loop to FR-P. 1 because the temperature is to the right of Limit A, but as identified in the E-3 Caution preceding the cooldown step the transition to FR-P. 1 is not to be made and E-3 would be continued until the note prior to the step for determining if containment spray pumps can be stopped.

Question Number: 84 Tier: 1 Group 2 K/A: W/E08 EG2.1.25 Pressurized Thermal Shock Conduct of Operations Ability to interpret reference materials, such as graphs, curves, tables, etc.

Importance Rating: 3.9! 4.2 IOCFRPart55: 41.10/43.5/45.12 IOCFR55.43.b: 5 KIA Match: K/A is matched because the question request the ability to use a graph (Limit A Curve) to determine status of the safety function associated with Pressurized Thermal Shock and then determine the correct procedure with which to mitigate, recover or with which to proceed required by the conditions. This question requires knowledge of the provisions for not following the standard rules of usage pertaining to the implementation of FR Red and Orange paths.

Technical

Reference:

FR-C, Status Trees, Revision 14 Page 231

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

FR-0, Status Trees, Revision 14 E-3, Steam Generator Tube Rupture, Revision 23 Proposed references FR-0, Status Trees, Limit A Curve with Acceptable to be provided: Region and Limit A wording deleted from inside the curve.

Learning Objective: 3-OT-FRP0001

1. Given a set of plant conditions, use the FR-P, Pressurized Thermal Shock Status Tree to identify and implement the appropriate Function Restoration Procedure (FR-P.1 or P2).5.

3-OT-EOPO300 Given a set of plant conditions, use E-3, ES-3.1, ES-3.2, and ES-3.3 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 EOPO300.05 013 modified to include use of a chart for the WBN 06/2011 NRC exam.

Changes made to stem and all 4 choices.

Comments:

Page 232

06/20 1 1 Watts Bar SRO NRC License Exam 6/22/2011

85. W/E13 EA2.2 085 Given the following:

- A reactor trip occurred on Unit I as a result of the loss of control air which resulted in the MSIVs closing.

- The crew is currently performing FR-H.2, Steam Generator Overpressure, with the following conditions existing.

- RCS temperature is 570°F.

SG#1 SG#2 SG#3 SG#4

- Pressure 1200 psig 1190 psig 1230 psig 1210 psig

- Level(NR) 43% 56% 90% 61%

Which ONE of the following is the proper procedure implementation?

A. Do NOT open SG#3 PORV and continue with FR-H.2.

B Manually open SG#3 PORV and continue with FR-H.2.

C. Do NOT open SG#3 PORV and transition to FR-H.3, Steam Generator High Level.

D. Manually open SG#3 PORV then transition to FR-H.3, Steam Generator High Level.

DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because if SG#3 narrow range level had been higher, the PORV would not be opened and continuing with FH.2 is correct for the conditions.

B. Correct, Conditions require SG#3 PORV to be opened to reduce the pressure in the steam generator and the FR-H. 2 to be continued.

C. Incorrect, Plausible because if SG#3 narrow range level had been higher, the PORV would not be opened and a transition to FR.3 would be correct.

D. Incorrect, Plausible because opening the PORV is correct for the stated conditions and a transition to FR. 3 would be correct if SG#3 narrow range level had been higher.

Question Number: 85 Page 233

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Tier: 1 Group 2 KIA: W/E13EA2.2 Steam Generator Overpressure Ability to determine and interpret the following as they apply to the (Steam Generator Overpressure)

Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

Importance Rating: 3.0 / 3.4 10 CFR Part 55: 43.5 / 45.13 IOCFR55.43.b: 5 K/A Match: K/A is matched because the question requires applying the conditions to procedure requirements and is SRO because the conditions must be applied to the procedrue steps to determine if proceeding with the current procedure or transition to another procedure is the proper operational decision. (Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.)

Technical

Reference:

FR-H.2, Steam Generator Overpressure, Revision 6 Proposed references None to be provided:

Learning Objective: 3-OT-FRH000 1

18. Identify the 2 major action categories of procedure FR-H.2 and explain the basis for each major action step.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question FRH0001.12 003 modified. Stem format, wording of stem & choices, correct answer location and data to make different answer correct included in the modification.

Comments:

Page 234

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Page 235

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

86. 012 A2.05 086 Given the following:

- Unit 1 is operating at 12% power when a component failure causes the following three annunicator windows to alarm:

93-F - EAGLE PROC PROT CH-IV RTD FAILURE 110-F - PROTSETTROUBLE 94-A - TAVG-TREF DEVIATION

- Status Panel window 79, PROT SET IV TROUBLE, is LIT.

- No other abnormal annunciators are LIT.

Which ONE of the following identifies...

(1) an impact of the failure that caused the alarms and (2) the procedure that will be implemented to mitigate and control the consequences of the failure?

A. (1) S/G LO-LO level reactor trip time delay interval will be changed.

(2) AOl-44, Eagle 21 Malfunctions B (1) S/G LO-LO level reactor trip time delay interval will be changed.

(2) AOl-2, Malfunction of Reactor Control System C. (1) The Rack Test Sequence Processor (TSP) circuit has failed.

(2) AOl-44, Eagle 21 Malfunctions D. (1) The Rack Test Sequence Processor (TSP) circuit has failed.

(2) AD 1-2, Malfunction of Reactor Control System Page 236

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the SG LO-LO level reactor trip time delays will be changed and using A 01-44 would be correct if an Eagle Rack had failed.

B. Correct, the alarms indicate a failure of either I Tcold RTD or 2 Thot RIDs. These RTDs are used for Tavg and Loop zITs. The loop ziTs for ioop 2, 3 and 4 provide the power level input to the $G LO-LO level reactor trip time delays and the delay is based on power level. If indicated power level is changing, then the time delay is changing. A 01-2 has a diagnostic for instrument failure directing the use of Subsection 3.2 which will defeat the failed Loop zlT, address T/S, and initiate repairs.

C. Incorrect, Plausible because an Eagle TSP failure would cause the alarms but would also cause additional alarms and using A 01-44 would be correct if an Eagle Rack had failed.

D. Incorrect, Plausible because an Eagle TSP failure would cause the alarms but would also cause additional alarms and using A 01-2 is correct in responding to the conditions Question Number: 86 Tier: 2 Group 1 KIA: 012 A2.05 Reactor Protection System (RPS)

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

Faulty or erratic operation of detectors and function generators Importance Rating: 3.1* / 3.2*

10 CFR Part 55: 41.5 / 43.5 I 45.3 / 45.5 IOCFR55.43.b: 5 K/A Match: K/A is matched because question requires knowledge of the impact of a malfunction (RCS RTD failure(s)) of a Reactor Protection system sensor on the RPS and the procedure (AOl) that is required to be used to correct, control, or mitigate the consequences of the malfunction. SRO because the question requires the assessment of plant conditions and then selecting between two Abnormal Operating Instructions (both having entry conditions in the alarms) to mitigate the malfunction. (Assessing plant conditions (normal, abnormal, or Page 237

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: K/A is matched because question requires knowledge of the impact of a malfunction (RCS RTD failure(s)) of a Reactor Protection system sensor on the RPS and the procedure (AOl) that is required to be used to correct, control, or mitigate the consequences of the malfunction. SRO because the question requires the assessment of plant conditions and then selecting between two Abnormal Operating Instructions (both having entry conditions in the alarms) to mitigate the malfunction. (Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.)

Technical

Reference:

AOl-2, Malfunction of Reactor Control System, Revision 37 AOI-44, Eagle 21 Malfunctions, Revision 0002 ARI-116-123, RPS & ESF, Revision 0010 ARI-88-94, Reactor Coolant System, Revision 0021 ARI-1 09-115, CVCS & RHR RPS & ESF,

-

Revision 0018 Proposed references None to be provided:

Learning Objective: 3-OT-SYSOO9OB

11. Describe the Trip Time Delay feature of the Eagle 21 system with regard to:

a. Inputs

b. Alarms

c. Outputs

12. Relative to Hot Leg and Cold Leg RTD inputs, describe the Eagle 21 control and alarm functions affected when the inputs fail and are subsequently disabled.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 238

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

87. 013 A2.04 087 Given the following:

- Unit 1 is in Mode 3 with the RCS at normal operating temperature and pressure.

- A loss of 12OvAC Vital Instrument Power Board 1-Ill occurs.

- AOl-25.03, Loss of 120V AC Vital Instrument Power Boards 1-Ill or 2-Ill, directs use of SOI-235.03, 120V AC Vital Power System 1-Ill, to restore the board.

Which ONE of the following identifies...

(1) how the Unit 1 SSPS Train A ESF relays are affected by the loss of the vital board and (2) the SOI-235.03 procedure section that if implemented to re-energize the board will allow Tech Spec LCO 3.8.9, Distribution Systems Operating to

-

be exited?

A. (1) ONLY the master relays could be energized.

(2) Section 8.3, Transfer 120V AC Vital Instrument Power Board 1-Ill to Spare 120V AC Vital Inverter 0-Ill.

B. (1) ONLY the master relays could be energized.

(2) Section 8.1, Transferring 480V AC Vital Transfer Switch Ill to Alternate 480V Power Supply.

C (1) Both the master relays and the slave relays could be energized.

(2) Section 8.3, Transfer 120V AC Vital Instrument Power Board 1-Ill to Spare I2OVAC Vital Inverter 0-Ill.

D. (1) Both the master relays and the slave relays could be energized.

(2) Section 8.1, Transferring 480V AC Vital Transfer Switch Ill to Alternate 480V Power Supply.

Page 239

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because if the loss had been Channel I instead of Channel Ill then only the master relays would be able to be energized and implementing Section 8.3 is correct.

B. Incorrect, Plausible because if the loss had been Channel I instead of Channel Ill then only the master relays would be able to be energized and implementing Section 8. 7 would restore the board but it would also place the board on the opposite Train power supply. The diesel generator battery charger normal and alternate feeds are both from the correct power train.

C. Correct, SSPS Train A is supplied by both 720V AC Vital Instrument Power Boards 7-I and 1-Ill. A loss of Channel Ill would result in the master relays still being energized because the master relays are powered via an auctioneered circuit from either Channel I or Channel Ill and the slave relays are powered from Channel I only, so they would still have the capability to be energized. Section 8.3 would restore the vital instrument power board to an operable status.

D. Incorrect, Plausible because both the master relays and the slave relays being able to be energized is correct and implementing Section 8. 7 would restore the board but it would also place the board on the opposite Train power supply. The diesel generator battery charger normal and alternate feeds are both from the correct power train.

Question Number: 87 Tier: 2 Group 1 KIA: 013 A2.04 Engineered Safety Features Actuation System (ESFAS)

Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Loss of instrument bus Importance Rating: 3.6 I 4.2 JO CFR Part 55: 41.5/43.5 / 45.3 / 45.13 JOCFR55.43.b: 2, 5 K/A Match: The K/A is matched because the question requires predicting the impact of the loss if an instrument bus on the ESFAS and is SRO because it requires the selection of a section of a procedure to mitigate and recover the loss of the instrument bus. (Assessing plant Page 240

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 K/A Match: The K/A is matched because the question requires predicting the impact of the loss if an instrument bus on the ESFAS and is SRO because it requires the selection of a section of a procedure to mitigate and recover the loss of the instrument bus. (Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.)(Knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.)

Technical

Reference:

AOl-25.03, Loss of 120V AC Vital Instrument Power Boards 1-Ill or 2-Ill, Revision 0025 SOl-235.03, 120V AC Vital Power System 1111, Revision 0020 Proposed references None to be provided:

Learning Objective: 3-OT-A0l2500

1) Demonstrate ability to recognize a loss of any 1 20V AC Vital Power Bd, including effects on equipment and controls (SOER 81-02).

3-OT-SYSO99A

2. Sketch a basic drawing of the Solid State Protection System.

8. Briefly discuss the input relays, Logic Section and Output Section of the SS PS.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 247

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

88. 061 G2.2.12 088 Given the following:

- Unit I is in Mode 4 with RCS heat up in progress following a refueling outage.

- 1-51-3-902, Turbine Driven Auxiliary Feedwater Pump lA-S Quarterly Performance Test, is determined to be out of frequency and past the maximum late date.

Which ONE of the following identifies if the AFW condition allows entry into Mode 3 and the associated Tech Spec action required if heatup is to be continued?

Mode 3 entry is...

A allowed and 1-Sl-3-902 must be satisfactorily completed within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> after SG pressure is> 1092 psig.

B. allowed and l-S1-3-902 must be satisfactorily completed within 48 hours2 days <br />0.286 weeks <br />0.0658 months <br /> afterSG pressure is> 1092 psig.

C. NOT allowed and 1-Sl-3-902 must be satisfactorily performed within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> or the TDAFW pump declared inoperable.

D. NOT allowed and 1-SI-3-902 must be satisfactorily performed within 48 hours2 days <br />0.286 weeks <br />0.0658 months <br /> or the TDAFW pump declared inoperable.

Page 242

06/2011 Watts Bar SRD NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Correct, Tech Specs allow the Mode 3 entry with the SI out of frequency but require the SI to be satisfactorily completed within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> after SG pressure is> 7092 psig. This is identified in a Note preceding SR 3.7.5.

B. Incorrect, Plausible because Tech Specs allowing the Mode 3 entry is correct but while 48 hours2 days <br />0.286 weeks <br />0.0658 months <br /> is a time associated with Tech Spec requirements for AEW (suction valves), the surveillance being required to be performed within 48 hours2 days <br />0.286 weeks <br />0.0658 months <br /> of the steam generator pressure reaching 1092 psig is not correct. The maximum tine allowed is 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> after the pressure reaches the setpoint.

C. Incorrect, Plausible because normally Mode changes cannot be made if an LCO with a limited required action completion time is in effect and 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> is a time allowed to perform a missed surveillance (SR 3.0.3) before a component is required to be declared inoperable.

D. Incorrect, Plausible because normally Mode changes cannot be made if an LCO with a limited required action completion time is in effect and with the pump SI out of frequency, SR 3.0.3 allows only 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> to complete the SI without a risk evaluation being required before the pump is required to be declared inoperable.

Question Number: 88 Tier: 2 Group 1 KIA: 061 G2.2.12 Auxiliary / Emergency Feedwater (AFW) System Equipment Control Knowledge of surveillance procedures.

Importance Rating: 3.7/4.1 JO CFR Part 55: 41.10/45.13 JOCFR55.43.b: 2 K/A Match: K/A is matched because the question requires the knowledge of surveillance procedures associated with the Auxiliary Feedwater system as to when the surveillances are required. SRO because the question requires knowledge of the requirements contains in the LCOs and SRs in Section 3.0 and 4.0 of Tech Specs.

Technical

Reference:

Tech Spec 3.7.5 Auxiliary Feedwater (AFW) System, Amendent 55 Tech Spec 3.0, LCO Applicability, Amendment 55 Page 243

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

Tech Spec 3.7.5 Auxiliary Feedwater (AFW) System, Amendent 55 Tech Spec 3.0, LCO Applicability, Amendment 55 Tech Spec 3.0, SR Applicability, Amendment 55 Proposed references None to be provided:

Learning Objective: 3-OT-T1S0700

4. Given plant conditions and parameters, correctly determine the Conditions for Operation or Technical Requirements for various components listed in Section 7 of Tech. Specs.

3-OT-T/S0700

7. Demonstrate the ability to determine the operablility of equipment based on required surveillances.

9. Given the surveillance performance history, determine if a change in Mode can occur.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 244

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011

89. 062 A2.16 089 Given the following:

- Unit 1 is operating at 100% power.

- Transmission System Grid conditions result in the 161 kV voltage slowly dropping.

- Annunciator window 501-B, CSST C ABNORMAL alarms.

- The crew determines the Train A 6.9kV Shutdown Board voltages to be:

IA-A 2A-A 6790v 7020v

- The SRO places the required CSST C tap changers to MANUAL and returns Shutdown board voltages to normal.

Which ONE of the following identifies...

(1) the status of 6.9kV Shutdown Board lA-A prior to manual tap changer adjustment and (2) how placing the tap changer in MANUAL affects the status of offsite power?

A. (1) Operable (2) Operability is NOT maintained B. (1) Inoperable (2) Operability is NOT maintained C. (1) Operable (2) Operability is maintained D(1) Inoperable (2) Operability is maintained Page 245

0612011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the minimum voltage to be operable could be mistaken to be the lower degraded voltage value and with the tap changer in manual the note in the procedure could be applied.

B. Incorrect, Plausible because the voltage is below the minimum allowed for the board and with the tap changer in manual the off site power supply remaining operable is correct.

C. Incorrect, Plausible because the minimum voltage to be operable could be mistaken to be the lower degraded voltage value and with the tap changer in manual the note in the procedure could be applied.

D. Correct, With the voltage less than 6800 volts the board voltage is below the Tech Spec minimum. The note in S01-200.04 identifies the tap changer being in manaul is an alternate alignment but is allowed at SRO discretion and placing in manual does not make the offsite system inoperable as long as limitations imposed by the WBN grid operating guide are met.

Question Number: 89 Tier: 2 Group 1 K/A: 062 A2.16 062 AC Electrical Distribution System Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Degraded system voltages Importance Rating: 2.5 2.9 10 CFR Part 55: 41.5/43.5/45.3/45.13 IOCFR55.43.b: 2, 5 K/A Match: K/A is matched because the question requires knowledge of the impact of a degraded system voltage on Tech Spec and is SRO because it requires detailed knowledge of how actions affect the operabilty of the offsite 1 E electrical power supplies.

Technical

Reference:

SOl-200.04, CSST C & D And Supply Breakers To 6.9kV Shutdown Boards, Revision 0014 1-PI-OPS-1-MCR, Main Control Room, Revision 0051 Tech Spec 3.8.9, Distribution Systems Operating

-

Page 246

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Technical

Reference:

501-200.04, CSST C & D And Supply Breakers To 6.9kV Shutdown Boards, Revision 0014 1-PI-OPS-1-MCR, Main Control Room, Revision 0051 Tech Spec 3.8.9, Distribution Systems Operating

-

ARI-501-508, Common Station Service, Revision 0020 0-Sl-0-3, Weekly Log, Revision 0036 1-75W530-1 R3 1-75W530-3 R2 Proposed references None to be provided:

Learning Objective: 3-OT-T/S0308

3. Given plant conditions and parameters correctly determine the OPERABILITY of components associated with the Electrical Power Systems.

3-OT-SYS2O 1 A

9. Discuss how voltage is maintained on the 6.9kV Shutdown Boards.

13. Identify the 6.9kV Shutdown Board parameters governed by Tech specs.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for WBN 06/20 1 1 NRC exam.

Comments:

Page 247

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

90. 076 G2.4.30 090 Given the following:

0700 - Unit 1 is operating at 100% power.

0800 - ERCW Header I B-B ruptures requiring the header to be crosstied to the ERCW 2A-A header.

0900 - A plant shutdown to Mode 3 is initiated.

Which ONE of the following identifies...

(1)the latest time allowed to make the initial required §50.72 report in accordance with NPG-SPP-03.5, Regulatory Reporting Requirements, and (2)the reason the report is required?

A. (1)1200 (2)The plant is in an unanalyzed condition that degrades plant safety.

B (1)1300 (2)A shutdown was initiated that was required by plant Tech Specs.

C. (1)1600 (2)The plant is in an unanalyzed condition that degrades plant safety.

D. (1)1700 (2)A shutdown was initiated that was required by plant Tech Specs.

Page 248

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible lithe plant being in an unanalyzed condition is assumed to be a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report and is the initiator for the required report. However, the report would be an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report.

B. Correct, In accordance with NPG-SPP-03. 5, Regulatory Reporting Requirements, the initiation of a Tech Spec Required shutdown is reportable within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />. With the shutdown started at 0900, the report is required by 1300.

C. Incorrect, Plausible because 1600 would be the required time for the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report due to the unanalyzed condition for the headers being crosstied.

D. Incorrect, Plausible because the initiation of a shutdown being required by Tech Spec is the correct initiator for the first required report and 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> is the maximum time allowed.

Question Number: 90 Tier: 2 Group 1 KIA: 076 G2.4.30 Service Water System Emergency Procedures I Plan Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Importance Rating: 2.7 / 4.1 10 CFR Part 55: 41.10 / 43.5 / 45.11 IOCFR55.43.b: 2, 7 K/A Match: The K/A is matched and is at the SRO level because the questions requires knowledge of Tech Spec requirements (including LCO 3.0, Applicability) on the Service Water System and of the NRC reporting requirements associated with a Tech Spec requited shutdown.

Technical

Reference:

NPG-SPP-03.5, Regulatory Reporting Requirements, Revision 0002 Tech Spec 3.7.8, Essential Raw Cooling Water (ERCW)

System, Amendment 69 Proposed references None to be provided:

Page 249

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011 Learning Objective: 3-OT-SYSO67A

24. 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 X Lower Question Source:

New X Modified Bank Bank Question History: New question for WBN 06/20]] NRC exam.

Comments:

Page 250

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

91. 001 G2.1.7 091 Given the following:

- Unit 1 operating at 100% power with Power Range Monitor Nl-41 out of service.

- Power is reduced to 96% due to secondary equipment problems, the following conditions exist.

- Control Bank D step counters indicate 194/1 93 steps.

- Control Bank D CERPI indications are:

Group 1 Group 2 M4 D4 M12 D12 H4 M8 H8 D8 H12 185 192 196 193 187 192 193 198 192

- QPTR is determined to be 1.04.

Which ONE of the following identifies if...

(1) LCO 3.1.5 Rod Group Alignment Limits is required to be entered and (2) the power level the unit must be lowered to due to QPTR?

A(1) No (2) 88%

B. (1) No (2) 84%

C. (1) Yes (2) 88%

D. (1) Yes (2) 84%

Page 251

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Correct, LCO 3.1.5 states that all individual rods shall have indicated rod positions within 12 steps of their group step counter demand position, which all of the given rod positions are within 12 steps. LCO 3.2.4 states that QPTR shall be less than or equal to 1.02. Condition A requires that thermal power be reduced by greater than or equal to 3% for each 1% of QPTR> 1.00. With QPTR 7.04, power would have to be reduced to 88%.

B. Incorrect, Plausible because LCO 3. 1.5 states that all individual rods shall have indicated rod positions within 12 steps of their group step counter demand position, which all of the given rod positions are within 72 steps and is correct. Also if the reduction had been 4% instead of 3% for each % the QPTR limit was exceeded, then reducing power to 84% would be correct.

C. Incorrect, Plausible because LCO 3.1.8, Rod Position Indication, requires the most withdrawn rod and the least withdraw rod to be within 72 steps of each other when a group demand step counter is inoperable. Rods M4 and D8 are not within 72 steps but they are in different groups and there is no step counter inoperable. Also plausible because LCO 3.2.4 states that QPTR shall be less than or equal to 1.02.

Condition A requires that thermal power be reduced by greater than or equal to 3%

for each 1% of QPTR> 1.00. With QPTR = 1.04, power would have to be reduced to 88%.

D. Incorrect, Plausible because LCO 3. 7.8, Rod Position Indication, requires the most withdrawn rod and the least withdraw rod to be within 72 steps of each other when a group demand step counter is inoperable. Rods M4 and D8 are not within 72 steps but they are in different groups and there is no step counter inoperable. Also if the reduction had been 4% instead of 3% for each % the QPTR limit was exceeded, then reducing power to 84% would be correct.

Question Number: 91 Tier: 2 Group 2 K/A: 001 G2.1.7 Control Rod Drive System Conduct of Operations 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: 41.5/43.5/45.12/45.13 Page 252

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 K/A Match: K/A is matched because the question requires evaluating Tech Spec requirements (operational judgements) based on operating characteristic (misaligned rod) and the required instrument interpretations to determine the reactor behavior (QPTR) with a misaligned rod. SRO because the question requires knowledge of actions and requirement below the line in 2 separate Tech Specs.

Technical

Reference:

1-SI-0-21, Excore QPTR & Axial Flux Difference, Revision 0017 Tech Spec LCD 3.1.5, Rod Group Alignment Limits Tech Spec LCD 3.1.8, Rod Position Indication Tech Spec LCD 3.2.4, Quadrant Power Tilt Ratio, Amendment 82 Proposed references None to be provided:

Learning Objective: 3-DT-T/S0301

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

3-OT-T/S0302

4. Given plant conditions, determine what action(s) must be taken and the bases for those actions.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

Page 253

06/2011 Waifs Bar SRO NRC License Exam 6/22/2011

92. 016 A2.02 092 Given the following:

- Reactor power is currently at 90% during a power reduction when the following occurs.

- Several annunciators alarm, simultaneously.

- Bank D control rods begin inserting and stop when the 0AC places rod control in manual.

- Steam generator levels start dropping but are stabilized by the manual actions of the CRC.

- The CR0 determines that one of the 120V AC Vital Instrument Power Boards has de-energized.

- OAC reports that Tavg is stabilized but is 4°F below Tref.

Which ONE of the following identifies...

(1) the procedure the Unit Supervisor would implement and (2) the required time allowed to restore the lost electrical board to operable status before being required to be in Mode 3 within the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />?

A. (1) AOI-25.02, Loss of 120V AC Vital Instrument Power Boards 1-Il or 2-Il (2) 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> B. (1) A0l-25.02, Loss oil 20V AC Vital Instrument Power Boards 1-Il or 2-Il (2) 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> C. (1) A0I-25.04, Loss of 120V AC Vital Instrument Power Boards 1-IV or 2-IV (2) 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> D (1) A0I-25.04, Loss of 1 20V AC Vital Instrument Power Boards I -IV or 2-IV (2) 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> Page 254

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because while the loss of 120v AC Vital Instrument Power Board 1-Il would result in some of the same initial conditions, but the plant could not be stabilized at power. A0I-25.02 would be correct if board 7-Il had been lost and Tech Spec 3.8.9 (which is the correct T/S) does allow 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> to restore an AC subsystem board that is inoperable, but in this case it is a Vital Board that is inoperable.

B. Incorrect, Plausible because while the loss of 720v AC Vital Instrument Power Board 1-Il would result in some of the same initial conditions, but the plant could not be stabilized at power. A 01-25.02 would be correct if board 7-Il had been lost and Tech Spec 3.8.9 (which is the correct T/S) allowing 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> to restore a AC board that is inoperable before requiring the unit be placed in Mode 3 within six hours is correct.

C. Incorrect, Plausible because implementing AOl-25.04 is correct and Tech Spec 3.8.9 (which is the correct T/S) does allow 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> to restore an A C subsystem board that is inoperable, but in this case it is a Vital Board that is inoperable.

D. Correct, The conditions would exist if the 120v AC Vital Instrument Power Board 1-IV were lost. The loss of the board requires the manual control of SG #4 feed water regulating valve due to loss of instrumentation. The condition is addressed in A 01-25.04 as well as the need to bypass the high power rod block to allow Tavg to be restored. T/S requires the board to be restored within 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> and if not then placing the unit in Mode 3 is required with 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />.

Question Number: 92 Tier: 2 Group 2 KIA: 016 A2.02 NonNuclear Instrumentation System Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of power supply Importance Rating: 2.9* / 3.2*

10 CFR Part 55: 41.5 / 43.5 I 45.3 / 45.5 IOCFR55.43.b: 5, 2 K/A Match: K/A is matched because the question requires the assessment of facility conditions and selection of appropriate procedure during an Page 255

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 K/A Match: KJA is matched because the question requires the assessment of facility conditions and selection of appropriate procedure during an abnormal situation. SRO because the question requires selecting between AOls to be entered and because it requires the knowledge of Tech Spec required actions (impact and consequences of the loss of power) which are below the line.

Technical

Reference:

AOl-25.01, Loss of 120v AC Vital Instrument Power Boards 1-I or 2-I, Revision 0031 AOI-25.04, Loss of 120v AC Vital Instrument Power Boards 1-IV or 2-IV, Revision 0027 Tech Spec 3.8.9, Distribution Systems Operating Proposed references None to be provided:

Learning Objective: 3-OT-A012500

4) Demonstrate ability/knowledge of AOl, by:

a. Recognizing Entry conditions

b. Responding to Actions

c. Responding to Contingencies (RNO)

d. Responding to Notes/Cautions Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: Calloway exam question modified Comments:

Page 256

06/201 1 Watts Bar SRO NRC License Exam 6/22/2011

93. 071 A2.01 093 Given the following:

- Unit 1 is at 100% power.

- At the Wednesday day shift turnover meeting, Chemistry reports that the Daily Monitoring Worksheet for Waste Gas Analyzer, 0-XIC-43-450, had not been completed during performance of 0-51-77-3, WDS Waste Gas Oxygen Determination, for Monday and Tuesday.

- During the shift, inadequate cover gas results in oxygen intrusion into HUT B as water is transferred to the Monitor Tank.

Which ONE of the following identifies...

(1) if the monitor can be considered operable at the beginning of the shift and (2) the lowest oxygen level detected that will result in implementation of SOI-77.02, Waste Gas Disposal System, Section 8.7, Response to Oxygen Intrusion?

A. (1) No (2) 2%

B. (1) No (2) 4%

C(1) Yes (2) 2%

D. (1) Yes (2) 4%

Page 257

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because the surveillance is beyond the 1.25 times the required frequency which is the maximum time allowed to perform the surveillance. Also plausible because the 2% oxygen level is correct for the lowest level of oxygen that would result in implementation of 501-77.02, section 8.7.

B. Incorrect, Plausible because the surveillance is beyond the 1.25 times the required frequency which is the maximum time allowed to perform the surveillance. Also plausible because at 4% oxygen the ARI for AUTO GAS ANALYZER UNITS I & 2 on panel 0-L-2 will require additional actions by the operating crew and 4% is the setpoint for the Oxygen Hi-Hi input to that annunciator.

C. Correct, SR 3.0.3 states that for a surveillance not performed within its specified frequency, entry into the LCO may be delayed up to 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> or the limit of the frequency, whichever is greater. Additionally, 2% oxygen level is correct for the lowest level of oxygen that would result in implementation of S0I-77.02, section 8.7.

D. Incorrect, SR 3.0.3 states that for a surveillance not performed within its specified frequency, entry into the LCO may be delayed up to 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> or the limit of the frequency, whichever is greater. Also plausible because at 4% oxygen the ARI for AUTO GAS ANALYZER UNITS 7 & 2 on panel 0-L-2 will require additional actions by the operating crew and 4% is the setpoint for the Oxygen Hi-Hi input to that annunciator.

Question Number: 93 Tier: 2 Group 2 K/A: 071 A2.01 Waste Gas Disposal System (WGDS)

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

Use of WGDS to prevent entry of oxygen into holdup tanks during liquid transfers importance Rating: 2.3? / 2.8?

10 CFR Part 55: 41.5/43.5/45.3/45.

IOCFR55.43.b: 2, 4 Page 258

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 KIA Match: K/A is matched because the question requires knowldege of the limit on oxygen intrusion into a holdup tank that requires implementation of a procedure to reduce the oxygen content in the tank and is SRO because the question requires the knowledge of the provision of SR 3.0 Surveillance Requirements and the applicablity of SR 3.0.3 to TS Section 5.7, Procedures, Programs, and Manuals.

Technical

Reference:

Tech Spec 5.7.2.15, Explosive Gas and Storage Tank Radioactivity Monitoring Program SOl-77.02, Waste Gas Disposal System, Revision 0035 ARI-0-L-2C, Waste Gas Panel, Revision 0005 1 -47W809-3 R22 1-47W61 1-77-5 R4 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO77A

19. Discuss how processed water is released.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the WBN 02/20 1 1 NRC exam Comments:

Page 259

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

94. G2.1.34 094 Given the following:

- Unit I is operating at 100% power.

- Chemistry reports the RCS Dissolved Oxygen concentration is above the TR-3.4.4, Reactor Coolant System, Chemistry Steady State Limit.

Which ONE of the following identifies...

(1) the completion time allowed to restore the Dissolved Oxygen to within limits without further action being required and (2) the relationship of the Transient Limits to the Steady State Limits in TR-3.4.4?

A. (1) 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> (2) The Transient Limits are 5 times the Steady State Limits.

B. (1) 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> (2) The Transient Limits are 10 times the Steady State Limits.

C. (1) 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> (2) The Transient Limits are 5 times the Steady State Limits.

D(1) 24hours (2) The Transient Limits are 10 times the Steady State Limits.

Page 260

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because TR-3.4.4 Chemistry does have actions with a completion time of 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to restore the concentration to within Transient Limits and the Transient Limits are 10 times (not 5 times) the Steady State Limits for dissolved oxygen parameters in the TR. Five times the RCS target value (lppb) is where an action level is entered in accordance with the CM-3.07.

B. Incorrect, Plausible because TR-3.4.4 Chemistry does have actions with a completion time of 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to restore the concentration to within Transient Limits and the Transient Limits being 10 times the Steady State Limits for each of the three parameters in the TR is correct.

C. Incorrect, Plausible because TR-3.4.4 Chemistry allowing 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> to restore the concentration to within the Steady State limit is correct but the Transient Limits are 10 times (not 5 times) the Steady State Limits for dissolved oxygen parameters in TR-3. 4.4, Chemistry. Five times the RCS target value (lppb) is where an action level is entered in accordance with the CM-3.01.

D. Correct, TR-3.4.4 Chemistry allows 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> to restore the concentration to within the Steady State limit and the Transient Limits are 10 times the Steady State Limits for each of the three parameters in the TR.

Page 267

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 94 Tier: 3 Group n/a KIA: G2.1.34 Conduct of Operations Knowledge of primary and secondary plant chemistry limits.

Importance Rating: 2.7 / 3.5 IOCFRPart55: 41.10/43.5/45.12 IOCFR55.43.b: 2 K/A Match: K/A is matched because the question requires knowledge of primary plant chemistry limits and is SRO because it requires knowledge of the Technical Requirement actions below the line.

Technical

Reference:

Tech Requirement 3.4.4, Chemistry, dated 9/30/95 and the associated bases.

Proposed references None to be provided:

Learning Objective: 3-OT-T-S0304

4. Given plant conditions and parameters correctly determine the applicable Limiting Conditions for Operations or Technical Requirements for the various components of the RCS.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History:

Comments:

Page 262

06/2011 Watts Bar SRO NRC License Exam 6/22/2011

95. G2.1.6 095 Given the following:

- Unit I is operating at 100% power.

- The Shift Manager is temporarily filling the Unit 1 SRO position while the Unit SRO is performing a work observation in the Auxiliary Building and is three minutes away.

- The CR0 has left the Main Control Room to assist an AUO in the Turbine Building.

- Other personnel in the Main Control Room are:

- the Operator at the Controls (OAC),

- a Work Control SRO with an inactive license, and

- an extra Unit Operator with an RO license.

- A rapid drop in SG #4 level results in a Reactor Trip.

- After the transition to ES-0.1, Reactor Trip Response, is made the Procedure Reader announces that due to conflicting indications that a transition to ES-0.0, Rediagnosis, is needed.

In accordance with Tl-12.04, Users Guide for Abnormal and Emergency Operating Instructions, which ONE of the following identifies...

(1) the preferred individual the Shift Manager will assign to be the procedure reader until the Unit 1 SRO returns and (2) if the transition to ES-0.0, Red iagnosis is allowed?

A. (1) The extra UnitOperator (2) Transition to ES-0.0 is allowed.

By (1) The extra Unit Operator (2) Transition is NOT allowed, remain in ES-0.1.

C. (1) The Work Control SRO (2) Transition to ES-0.0 is allowed.

D. (1) The Work Control SRO (2) Transition is NOT allowed, remain in ES-0.1.

Page 263

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because an extra UO with an RO License is correct and while transitioning to ES-O. 0 for rediagnosis is appropriate if the crew is uncertain, only one of the two required conditions exist (Transition from E-0.) The other being That an SI has been actuated does not exist, so the required condition for using ES-0. 0 do not exist.

B. Correct, The Procedure Reader selected by the Shift Manager would be the extra Unit Operator with an RO license in accordance with the recommended order of priority indentified in Tl-12.04 and the transition to ES-0.0 is not allowed to be made because an SI has not been actuated.

C. Incorrect, Plausible because the work control SRO would be the correct reader if the SRO had an active license and while transitioning to ES-0. 0 for rediagnosis is appropriate if the crew is uncertain, only one of the two required conditions exist (Transition from E-0.) The other being that an SI has been actuated does not exist, so the required condition for using ES-0. 0 do not exist.

D. Incorrect, Plausible because the work control SRO would be the correct reader if the SRO had an active license and the transition to ES-0. 0 is not being allowed to be made is correct because an SI has not been actuated.

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06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 95 Tier: 3 Group n/a K/A: G2.1.6 Conduct of Operations Ability to manage the control room crew during plant transients.

Importance Rating: 3.8* / 4.8 10 CFR Part 55: 41.10 /43.5/45.12/45.13 10CFR55.43.b: 5 KIA Match: K/A is matched and is SRO because the question requires knowledge of administrative directions the Shift Manager should follow in assigning task to the crew during reduced staffing and the ability to assess of plant conditions to select a procedure to mitigate, recover or with which to proceed.

Technical

Reference:

TI-i 2.04, Users Guide for Abnormal snd Emergency Operating Instructions, Revision 0009 Proposed references None to be provided:

Learning Objective: 3-OT-TI 1204

5. Identify the Operating Team member who should NOT serve as the procedure reader during any plant transient.

30. List 2 limitations which apply to the use of procedure ES-0.0, Rediagnosis.

Cognitive Level:

Higher X Lower Question Source:

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

Comments:

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06/2011 Watts Bar SRO NRC License Exam 6/22/2011

96. G 2.2.36 096 Given the following:

- Unit 1 is operating at 100% power with RHR pump 1 B-B out of service and tagged.

0750 - DG IA-A is determined to be inoperable.

0845 - 0-S 1-82-2, 8 Hour Diesel Generator AC Power Source Operability Verification, is completed.

Which ONE of the following identifies the latest time allowed by Tech Specs...

(1) to declare RHR pump IA-A inoperable.

and (2) for the next completion of 0-S 1-82-2?

A. (1) 1150 (2) 1645 B(1) 1150 (2) 1845 C. (1) 0750 the following day (2) 1645 D. (1) 0750 the following day (2) 1845 Page 266

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because maximum allowable time to declare the RHR pump IA-A inoperable is 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> with the DG inoperable but 1645 is not the latest time allowed for the performance of the SI because it does not include the allowed extension time (applying the 1.25 extension time is correct) after the first required performance of the surveillance instruction.

B. Correct, A Tech Spec LCO 3.8.1 required action is to Declare required feature(s) with no DG available inoperable when its redundant required feature(s) is inoperable within the required Completion time of 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> from discovery of no DG power to one train concurrent with inoperability of redundant required feature(s) .

Also, Tech Spec SR 3.0.2 states The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. and If a Completion Time requires periodic performance on a once per. . . basis, the above Frequency extension applies to each performance after the initial performance.

C. Incorrect, Plausible because maximum allowable time to declare the RHR pump IA-A inoperable would be 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> (0750 the following day) if there was no offsite power circuit available to the Train B board but 1645 is not the latest time allowed for the performance of the SI because it does not include the allowed extension time (applying the 1.25 extension time is correct) after the first required performance of the surveillance instruction.

D. Incorrect, Plausible because maximum allowable time to declare the RHR pump IA-A inoperable would be 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> (0750 the following day) if there was no offsite power circuit available to the Train B board and 1845 is correct because the 1.25 extension time can be applied to the interval specified in the Frequency for each performance after the initial performance.

Question Number: 96 Tier: 3 Group n/a KIA: G2.2.36 Equipment Control Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Importance Rating: 3.1 / 4.2 10 CFR Part 55: 41.10 I 43.2 / 45.13 IOCFR55.43.b: 2 Page 267

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 K/A Match: KA is matched because the question requires knowledge of how maintenance activities and degraded power sources affect LCO status and required actions.

Technical

Reference:

0-61-82-2, 8 Hour Diesel Generator AC Power Source Operability Verification, Revision 0012 Tech Spec LCO 3.8.1, AC Sources Operating,

-

Amendment 55 Tech Spec 3.0, Surveillance Requiremetns (SR)

Applicability, SR 3.0.2, Amendment 42 Proposed references None to be provided:

Learning Objective: 3-OT-T/S0300

6. Given plant conditions where LCOs and/or TRs are not met, determine what actions must be performed on associated supported equipment

8. Given the surveillance performance history, determine the limit for the next performance of a surveillance.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History:

Comments:

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06/2011 Watts Bar SRO NRC License Exam 6/22/2011

97. G 2.2.6 097 Given the following:

- Unit I is in MODE 5 during a refueling outage.

- It is determined that SOI-63.01, Safety Injection System, needs a Minor/Editorial Change revision prior to performance of Section 5.1, Fill & Vent SI Pumps and Piping from RWST.

In accordance with NPG-SPP-01 .2, Administration of Site Procedures, Which ONE of the following statements identifies the requirements for a 50.59 Screening Review and/or an Independent Qualified Reviewer (IQR) for the proposed revision?

A. Both are required.

B. Neither is required.

C Only the IQR is required.

D. Only the 50.59 Screening Review is required DIS TRACTOR ANALYSIS:

A. Plausible because both the 50.59 review and an IQR is requited for other types of revisions.

B. Plausible because 50.59 reviews not being requited is correct and the requirements for reviews do change depending on the type of revision.

C. Correct, NPG-SPP-07.2 (Section 3.7 A) states an IQR review for minor editorial changes for Quality Related procedures is required and also states that 70 CFR 50.59 reviews are not required.

D. Incorrect, Plausible because 50.59 reviews are addressed in the procedure and the 50.59 screening review is required for other type of revisions.

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06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 97 Tier: 3 Group n/a K/A: G 2.2.6 Equipment Control Knowledge of the process for making changes to procedures.

Importance Rating: 3.0 I 3.6 10 CFR Part 55: 41.10 / 43.3 / 45.13 10CFR55.42.b: 3 K/A Match: K/A is matched because the question requires knowledge of the reviews required (process) for making changes to procedures. SRO only because the question requires the knowledge of the processes for changing plant procedures.

Technical

Reference:

NPG-SPP-01 .2, Administration of Site Technical Procedures, Revision 0001 Proposed references None to be provided:

Learning Objective: 3-OT-NPGSPPO1O2

5. Describe the procedure revision process.

6. Define the difference between a minor/editorial change and a revision.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: WBN bank question G 2.2.6 096 modified to reflect procedure change. Second part of question & all 4 choices modified, minor changes to wording in the stem, and relocated position of correct answer.

Comments:

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06/2011 Watts Bar SRO NRC License Exam 6/22/2011

98. G2.3.14 098 Which one of the following choices completes the statement below?

With Unit I in Mode 1, compliance with Technical Specification 3.4.16, RCS Specific Activity, ensures that the 2-hour dose at the site boundary will not exceed a small fraction of (1) limits following a (2) or a Main Steam Line Break.

Li)

A 1OCFRIOO Steam Generator Tube Rupture B. IOCFR100 Loss of Coolant Accident C. 100FR2O Steam Generator Tube Rupture D. 1OCFR2O Loss of Coolant Accident DIS TRA CTOR ANAL YSIS:

A. Correct, the limit being based on not exceeding a fraction of the Part 700 limits during a SGTR or Main Steam Line Break accident are included in the Bases for Tech Spec 3.4.16, RCS Specific Activity.

B. Incorrect, Plausible because not exceeding a fraction of the Part 100 limits is correct and a LOCA is a design bases accident where elevated dose rates could be encountered.

C. Incorrect, Plausible because Part 20 identifies radiological limits and the bases being during a SGTR or Main Steam Line Break accident is correct.

D. Incorrect, Plausible because Part 20 identifies radiological limits and a LOCA is a design bases accident where elevated dose rates could be encountered.

Question Number: 98 Tier: 3 Group n/a KIA: G2.3.14 Radiological Controls Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Importance Rating: 3.4 / 3.8 JOCFRPart55: 41.12/43.4/45.10 Page 277

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 1 OCFR55.43.b: 2,4 K/A Match: K/A is matched because the question requires the knowledge of radiation hazards that may arise during plant conditions and is SRO because of both the knowledge of the radiation hazards that may arise during normal, abnormal and emergency conditions and the Tech Spec bases for limits on RCS specific activity.

Technical

Reference:

Tech Spec Basis 3.4.16, RCS Specific Activity, Revision 52 (Amendment 41)

Proposed references None to be provided:

Learning Objective: 3-OT-T/S-0304

00. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with the Reactor Vessel that are rated >2.5 during Initial License Training and >3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank X Bank Question History: SQN bank question G 2.3.14 099 modified by changing second part of the question in both the stem and the choices. Correct answer relocated.

Comments: SQN bank question G 2.3.14 099 used on the SQN 1/2009 retake exam.

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06/2011 Watts Bar SRO NRC License Exam 6/22/2011

99. G 2.3.6 099 Which ONE of the following identifies...

(1) the minimum time required to allow the contents of a Gas Decay tank to decay prior to release, and (2) who can waive the minimum time in accordance with SOI-77.02, Waste Gas Disposal System?

Decay Time Who can waive A. 60 days Chemistry Duty Manager B. 60 days Radiation Protection Manager C. 40 days Chemistry Duty Manager D. 40 days Radiation Protection Manager DIS TRACTOR ANAL YSIS:

A. Correct, The procedure requires a 60 day decay time and does provide for waiving of the time by the Chemistry Duty Manager.

B. Incorrect, The decay time required is 60 days but the Radiation Protection Manager is not the position that can waive the requirement if earlier release is required.

Plausible because the required time is for radioactive decay which could be addressed by RadCon.

C. Incorrect, The decay time required is 60 days, not 40 days, but the waiving of the requirement by the Chemistry Duty Manager is allowed by the procedure.

Plausible because 40 days would be five times the 8 day half live identified in the ODCM for Gaseous Effluents radionuclides that set dose rate limits at and beyond the Unrestricted Area Boundary and because the Chemistry Duty Manager is correct.

D. Incorrect, The decay time required is 60 days, not 40 days and the waiving of the requirement by the Radiation Protection Manager is not allowed by the procedure.

Plausible because 40 days would be five times the 8 day half live identified in the ODCM for Gaseous Effluents radionuclides that set dose rate limits at and beyond the Unrestricted Area Boundary and because the required time is for radioactive decay which could be addressed by RadCon.

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06/2011 Watts Bar SRD NRC License Exam 6/22/2011 Question Number: 99 Tier: 3 Group n/a K/A: G2.3.6 Radiation Control Ability to approve release permits.

Importance Rating: 2.9 / 2.9 10 CFR Part 55: 41.13 I 43.4 I 45.10 IOCFR55.43.b: 4 K/A Match: K/A is matched and the question is SRO because it requires knowledge of limits that may be encountered (decay time) to prevent a radiation hazard from arising during a normal operation and the approvals needed for waiving the normal decay time when approving a release.

Technical

Reference:

501-77.02, Waste Gas Disposal System, Rev 0035.

Proposed references None to be provided:

Learning Objective: 3-OT-SYSO77B

10. Describe the general procedure to make a gaseous release.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question G 2.3.6 097 that was used on the WBN 5/2008 exam.

Comments:

Page 274

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 100. G2.4.38 100 Given the following plant conditions:

- A Site Area Emergency has been declared.

- Emergency Centers have NOT been activated.

Which of the following identifies the limitations, if any, on the delegation of the Site Emergency Director responsibilities in accordance with the Radiological Emergency Plan Implementing Procedures?

Emergency Classification Determination of escalation Protective Action Recommendations A Can NOT be delegated. Can NOT be delegated.

B. Can NOT be delegated. Can be delegated.

C. Can be delegated. Can NOT be delegated.

D. Can be delegated. Can be delegated.

DISTRA CTOR ANAL YSIS:

A. Correct, The classification level cannot be delegated and with the CECC not staffed and active, the PAR could not be delegated by the SED.

B. Incorrect, Plausible because the classification level not being delegated is correct and the determination of Protective Actions Recommendations (PAR) could be delegated to the CECC director if the CECC were staffed and activated.

C. Incorrect, Plausible because the SED can delegate other responsibilities and with the CECC not being staffed, the PAR not being delegated is correct.

D. Incorrect, Plausible because the SED can delegate other responsibilities and if the CECC were staffed and activated the PAR could be delegated.

Page 275

06/2011 Watts Bar SRO NRC License Exam 6/22/2011 Question Number: 100 Tier: 3 Group n/a KIA: G2.4.38 Emergency Procedures/Plan Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Importance Rating: 2.4 / 4.4 JO CFR Part 55: 41.10 / 43.5 / 45.11 IOCFR55.43.b: 4, 5 KIA Match: K/A is matched and is SRO only because the question requires the ability to take action as the SED (emergency coordinator) and know the requirements and limitations duties of the position, including the recall of specific facts regarding SED duties and their execution (what cannot be delegated).

Technical

Reference:

EPIP-1, Emergency Plan Classifiction Logic, Revision 0034; EPIP-6, Activation and Operation of the Technical support Center (TSC), Revision 038; Proposed references None to be provided:

Learning Objective: 3-OT-PCD-048C

7. Identify Operations responsibilities for:

Site Emergency Director (who is initially the SM)

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question G 2.4.38 100 used on the 05/2009 exam Comments:

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