ML092810407

Initial Exam 2009-301 Draft RO Written Exam (Questions 1 - 75)

ML092810407
Person / Time
Site:	Watts Bar
Issue date:	10/08/2009
From:	NRC/RGN-II
To:
References
50-390/09-301, 50-391/09-301
Download: ML092810407 (77)

Text

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 1 of 77 Question Number: 1 K/A: 000008 G2.2.20 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Knowledge of the process for managing troubleshooting activities.

Tier:

1 RO Imp:

2.6 RO Exam:

1 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

1 Source:

New Applicable 10CFR55 Section: 41.10 / 43.5 / 45.13 Learning Objective:

3-OT-SYS068C, Rev. 7, Objective 11, Describe the indication an operator has that a PORV is open or leaking through.

References:

AOI-18, "Malfunction of Pressurizer Pressure Control System", Rev. 21; ARI-88-94, Rev 19; GO-1, "Unit Startup from Cold Shutdown to Hot Standby", Rev. 60.

Question: 1 With the plant at 100% power, the following conditions are noted:

A PZR PORV LINE TEMP HI is LIT.

B PZR SAFETY LINE TEMP HI is LIT.

C PRT PRESS HI is LIT.

- Both PZR PORVs indicate CLOSED.

- PZR pressure indicates 2230 psig and is slowly lowering.

For these conditions, the Operator-at-the-Controls (OAC) will:

a.

Request permission from the US to take manual control of Pressurizer pressure and reduce pressure approximately 50 psi to attempt seating the leaking safety valve.

b.

Set up an historical trend on the ICS computer to monitor PRT pressure, temperature, and level for a rise in the associated parameter to determine which valve is leaking.

c.

Request permission from the US to close both PORV block valves and then monitor PRT parameters to see if leak is isolated.

d.

Request permission from the US to close one PORV block valve at a time, and then monitor PRT parameters to see if leak is isolated.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since similar actions are performed under plant conditions (during plant heatup) to reseat a leaking PZR safety valve, per GO-1, Unit Startup from Cold Shutdown to Hot Standby. With the plant at full power a 50 psi pressure reduction requires LCO 3.4.1 entry.

b.

Incorrect. Plausible, since trending PRT data does provide partial information required to troubleshoot the valve failure. PRT parameter changes could be indicative of other problems (letdown relief, etc.).

c.

Incorrect. Plausible, since closing both block valves does result in isolation of the leaking PORV, but is not the determining factor in diagnosing which specific valve is leaking. Also plausible if candidate believes isolating the second PORV is a conservative measure, to preclude it from spurious opening while the OTHER PORV block valve is being operated.

d.

CORRECT. These actions are taken from ARI 89-A Corrective Actions.

K/A Applicability:

Question addresses how troubleshooting is performed (in this case, it is directed by the Alarm Response Instruction), specifically a pressurizer component associated with the vapor space (a relief valve which sticks open).

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 2 of 77 Question Number: 2 K/A: 000009 EK2.03 Knowledge of the interrelations between the small break LOCA and the following:

S/Gs Tier:

1 RO Imp:

3.0 RO Exam:

2 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

2 Source:

INPO Bank Applicable 10CFR55 Section: (CFR 41.7 / 45.7)

Learning Objective:

3-OT-EOP0001, Rev. 11, Objective 12, Discuss the purpose of ES-1.2, Post LOCA Cooldown and Depressurization.

References:

ES-1.2, "Post LOCA Cooldown and Depressurization", Rev. 14. WOG Background Document ES-1.2, Rev 2.

Question: 2 Given the following plant conditions:

- The plant was initially at full power, end of life conditions.

- A small break LOCA occurred.

- The reactor tripped.

- The RCS is saturated with RCS temperature at 580°F.

Assuming NO operator actions have been taken, which one of the following components is required for establishing adequate core cooling?

a.

Accumulators.

b.

Steam generators.

c.

Safety injection pumps.

d.

Reactor coolant pumps.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the accumulators are a component designed to inject cool water into the RCS. However, candidate must determine from the conditions in the stem, that RCS pressure is higher than injection pressure capability of the cold leg accumulators.

b.

CORRECT. During the small break LOCA, the cooling from ECCS injection is not sufficient to remove decay heat. Candidate must determine from the conditions in the stem that RCS pressure, though lower than ECCS injection pressure, does not allow adequate injection flow for cooling. The steam generators become the critical components to ensure core damage does not occur.

c.

Incorrect. Plausible, since the safety injection pumps are operating, and injecting, but it is only partial injection, since SI pump shutoff head is approx. 1650 psig and current RCS pressure is approx. 1315 psig.

d.

Incorrect. Plausible, if the candidate fails to determine and understand the significance of the RCS pressure that corresponds to 580°F. The operators will manually stop the RCPs when RCS pressure is less than 1500 psig (i.e., RCS pressure is 1315 psig).

K/A Applicability:

Stem conditions are for a small break LOCA. Question tests understanding of the relationship between this event and the steam generators, in the context of adquate core cooling.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 3 of 77 Question Number: 3 K/A: 000011 EA1.05 Ability to operate and monitor the following as they apply to a Large Break LOCA:

Manual and/or automatic transfer of suction of charging pumps to borated source Tier:

1 RO Imp:

4.3 RO Exam:

3 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

3 Source:

Bank Modified Applicable 10CFR55 Section: (CFR 41.7 / 45.5 / 45.6)

Learning Objective:

3-OT-EOP0000, Obj 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.

References:

E-0, "Reactor Trip or Safety Injection", Rev. 27, Appendix A, Step 4.b. RNO.

Question: 3 With the plant initially at full power, the following conditions are given:

- A large break LOCA occurs.

FCV-62-136, Charging Pumps (CCP) Suction from RWST, is CLOSED.

Which ONE of the following describes:

(1) the current suction source to the CCPs, and (2) what action, if any, is necessary to ensure that the CCPs have an adequate suction source of borated water?

CCP Suction Source Action Required

a.

Both charging pumps are receiving suction flow only from the VCT.

Manually open 1-FCV-63-136.

b.

ONE charging pump is receiving suction flow from the RWST.

No additional action required.

c.

Both charging pumps are receiving suction flow only from the RWST.

No additional action required.

d.

ONE charging pump is receiving suction flow from the RWST.

Manually open 1-FCV-63-136.

DISTRACTOR ANALYSIS A. Incorrect. Plausible, since both charging pumps will be operating with a suction source, and since the stem lists a CLOSED valve on the suction from the RWST. However, candidate has an incorrect understanding of the layout of 1-FCV-63-136 and 1-FCV-63-135, which is the other valve in parallel with 63-136 on the suction line from the RWST. If candidate fails to recall that either valve being open provides adequate suction flow to both operating charging pumps, Column (2) is further plausible.

B. Incorrect. Plausible, since candidate may believe a valve which should have automatically opened for the SIS (and did not), results in RWST suction source to only ONE of the charging pumps. The candidate may then erroneously conclude that the other charging pump is still receiving suction from the VCT. Column (2) for any required action is plausible in this context because candidate may believe that only one charging pump taking suction from the RWST is adequate boration for given conditions.

C. CORRECT. Both CCPs will take a suction on the RWST if either of the parallel suction valves is open.

The candidate must deduce from the conditions in the stem that one valve is still open. The VCT is

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 4 of 77 isolated automatically by the SIS, provided that at least ONE of the valves from the RWST is open (and one of them IS indeed open).

D. Incorrect. Plausible because one charging pump IS receiving flow from the RWST. However, candidate may incorrectly believe a valve which should have automatically opened for the SIS (and did not) results in RWST suction source to only ONE of the charging pumps. This leads the candidate to conclude that the other charging pump is still receiving suction from the VCT. Column (2) for required action is plausible because candidate may believe that only one charging pump taking suction from the RWST is NOT adequate boration for given conditions, and that manual action must be taken to open the other valve.

K/A Applicability:

Stem conditions include a large break LOCA, and the question tests understanding of charging pump suction alignment during the event.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 5 of 77 Question Number: 4 K/A: 000015/017 AK2.10 Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: RCP indicators and controls Tier:

1 RO Imp:

2.8 RO Exam:

4 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

4 Source:

GFE Bank Applicable 10CFR55 Section: (CFR 41.7 / 45.7)

Learning Objective:

3-OT-AOI2400, Obj. 10, Given a set of plant conditions, use AOI-24 to correctly: a.

Recognize Entry Conditions; b. Identify Required Actions; c. Respond to Contingencies (RNO); d. Observe and Interpret Cautions and Notes.

References:

AOI-24, "RCP Malfunctions During Pump Operation", Rev. 28.

Question: 4 Given the following plant conditions:

- The unit is operating at 100% power.

- A reactor coolant pump (RCP) malfunction occurs.

Thirty seconds after the malfunction, which one of the following will provide positive indication for the operator to distinguish between a locked RCP rotor and a sheared RCP rotor? (Assume no operator action is taken.)

a.

Reactor trip status.

b.

Loop flow indications.

c.

RCP ammeter indications.

d.

Loop differential temperature indications.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the reactor will be tripped, and candidate may believe that since a sheared RCP rotor does not trip the RCP, and since all four RCPs are still running, the reactor does not automatically trip. Candidate then recalls that a locked rotor does result in an overcurrent trip of the RCP, and also results in a reactor trip. This distractor is from the current GFES exam bank.

b.

Incorrect. Plausible, since loop flow indications will be affected. Candidate incorrectly believes that the difference in these indications are significant enough to distinguish between a locked rotor or sheared shaft. This distractor is from the current GFES exam bank.

c.

CORRECT. A locked rotor does result in high amps initially, but when the breaker trips on overcurrent the ammeter indicates 0 amps. A sheared shaft results in low amps until operation action is taken to trip the pump.

d.

Incorrect. Plausible, since loop differential temperatures in the affected loop are affected, however not to the extent required for diagnosing between a locked rotor and a sheared shaft. This distractor is from the current GFES exam bank.

K/A Applicability:

Question stem presents a loss of RC flow condition caused by an RCP malfunction, and candidate is tested on which indication is used to diagnose the type of malfunction. The correct answer involves an RCP indication.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 6 of 77 Question Number: 5 K/A: 000025 AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System:

Shift to alternate flowpath Tier:

1 RO Imp:

3.1 RO Exam:

5 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

5 Source:

INPO Bank Applicable 10CFR55 Section: (CFR 41.5,41.10 / 45.6 / 45.13)

Learning Objective:

3-OT-AOI1400, Obj. 5, Explain Alternate RHR Cooling methods.

References:

AOI-14, "Loss of RHR Shutdown Cooling", Rev. 34 Question: 5 Given the following plant conditions:

- The plant is in Mode 5.

- RCS temperature is 195°F and stable.

- RCS pressure is 325 psig and stable.

- Train "A" RHR is in service.

- Train "B" RHR is out of service for repairs to the B RHR pump seals.

- A loss of Train "A" RHR shutdown cooling occurs.

For these conditions, which ONE of the following is the preferred method for heat removal in accordance with AOI-14, "Loss of RHR Shutdown Cooling"?

a.

RWST gravity feed to RCS with spill through the PZR PORVS.

b.

SI Pump Hot Leg Injection with spill through a 2-inch vent.

c.

Natural circulation or forced RCS flow while steaming intact S/Gs.

d.

Reflux cooling to any S/G with level equal to or greater than 29% NR level.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since gravity feed is a method for cooling the core. The conditions described in the stem do not support gravity feed, since RCS pressure is at 325 psig.

b.

Incorrect. Plausible, since the SI pump could inject into the RCS with the conditions given in the stem.

However, for the given conditions, Cold Overpressure Protection is in service, and the SI pumps are therefore disabled.

c.

CORRECT. Since the RCS is intact and full, the use of the S/Gs as a heat sink with either forced or natural circulation flow is the preferred method for cooling.

d.

Incorrect. Plausible, since reflux cooling is a potential method for removing heat from the core. The conditions described in the stem do not support reflux boiling as a heat removal process.

K/A Applicability:

Stem includes a loss of RHR event, and the question tests knowledge of which alternate method (alternate method of flow through the core and cooling) will provide adequate heat removal.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 7 of 77 Question Number: 6 K/A: 000026 AK3.04 Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water:

Effect on the CCW flow header of a loss of CCW Tier:

1 RO Imp:

3.5 RO Exam:

6 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

6 Source:

New Applicable 10CFR55 Section: (CFR 41.5,41.10 / 45.6 / 45.13)

Learning Objective:

3-OT-AOI1500, Obj. 7, Determine Action for Loss of an ESF Equipment header.

References:

AOI-15, "Loss of Component Cooling Water (CCS)", Rev. 31; ARI 226-E ERCW/CCS MOTOR TRIPOUT; 1-47W611-70-1, Logic Diagram Component Cooling System, Rev. 9.

Question: 6 A leak has developed on the #3 RCP thermal barrier. After performance of AOI-15, Loss of Component Cooling Water (CCS) Section 3.2, CCS flow will _____(1)_____ through the Miscellaneous Equipment Header, since ______(2)_______.

(1)

(2)

a.

Decrease only the flow to all RCPs' thermal barriers has been isolated.

b.

Decrease the flow to all thermal barriers and RCP oil coolers has been isolated.

c.

Increase only the flow to all RCPs thermal barriers has been isolated.

d.

Increase the flow to all thermal barriers and RCP oil coolers has been isolated.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the isolation of the leak results in a complete isolation of the thermal barrier flowpath, and requires stopping the thermal barrier booster pumps. This causes the overall flow through the Reactor Building Header to drop and provides more flow to the Miscellaneous Equipment Header.

b.

Incorrect. Plausible, since the isolation of the leak results in a complete isolation of the thermal barrier flowpath, and requires stopping the thermal barrier booster pumps. RCP oil coolers will not be isolated.

This causes the overall flow through the Reactor Building Header to drop and provide more flow to the Miscellaneous Equipment Header.

c.

CORRECT. The isolation of the flow to all RCP thermal barriers results in an overall drop in Reactor Building header flow, and a corresponding rise in Miscellaneous Equipment header flow.

d.

Incorrect. Plausible, since the isolation of the flow to all RCP thermal barriers results in an overall drop in Reactor Building header flow, and a corresponding rise in Miscellaneous Equipment header flow.

RCP oil cooler flow is not isolated for the thermal barrier leak.

K/A Applicability:

Stem conditions include a loss of component cooling through a thermal barrier on an RCP. Question tests understanding of the effect of this loss on other parts (Misc. Equipment Header) of the component cooling system.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 8 of 77 Question Number: 7 K/A: 000027 AK1.03 Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions:

Latent heat of vaporization/condensation.

Tier:

1 RO Imp:

2.6 RO Exam:

7 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

7 Source:

Indian Point 2007 Applicable 10CFR55 Section: (CFR 41.8 / 41.10 / 45.3)

Learning Objective:

3-OT-SYS068C, Obj. 8, Describe the operation of the master pressure controller.

References:

System Description, N3-68-4001, Reactor Coolant System, Rev. 25 Question: 7 With the plant initially at 100% power, 1-A MFP trips. The Pressurizer master pressure controller output signal fails AS IS at the same time as the MFP trip.

Which one of the following occurs in the Pressurizer (PZR) to help limit the magnitude of the initial pressure transient on the Reactor Coolant System (RCS)?

a.

An outsurge cools the PZR. This allows some steam to condense to water and limits the resulting pressure increase in the RCS.

b.

An outsurge causes the steam space to expand in the PZR. This allows some liquid to flash to steam and limits the resulting pressure drop in the RCS.

c.

An insurge of hotter water heats the PZR. More liquid then flashes to steam which helps limit the resulting pressure drop in the RCS.

d.

An insurge of cooler water compresses the steam space in the PZR. Steam is condensed to water which helps limit the overall pressure increase in the RCS.

DISTRACTOR ANALYSIS

a.

Incorrect. An insurge is expected. During an outsurge, water flash, not condense. Plausible, since the second part of the answer is correct for an insurge.

b.

Incorrect. An insurge is expected. Plausible, since the second part of the answer is correct for an outsurge.

c.

Incorrect. An insurge cools the PZR causing the liquid to be subcooled. Plausible since the second part of the answer is correct for an outsurge.

d.

CORRECT. RCS temperature rises and causes an insurge into the PZR and an increase in pressure.

Since the pressure control system is failed as is, spray will not respond, leaving only condensation of steam in the vapor space which has been compressed.

K/A Applicability:

Question tests understanding of the operational implications and effects of an expected Pressurizer insurge following a transient.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 9 of 77 Question Number: 8 K/A: 000029 EA1.12 Ability to operate and monitor the following as they apply to a ATWS: M/G set power supply and reactor trip breakers.

Tier:

1 RO Imp:

4.1 RO Exam:

8 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

8 Source:

New Applicable 10CFR55 Section: (CFR 41.7 / 45.5 / 45.6)

Learning Objective:

3-OT-SYS085B, Obj 9, Explain how to locally trip the reactor in the event of an ATWS.

References:

FR-S.1, Nuclear Power Generation/ATWS, Rev. 19; SR 3.1.5.3, Rod Group Alignment Limits.

Question: 8 If the 480V Unit Board breakers to the MG sets are the ONLY breakers that can be opened following an ATWS, which one of the following describes:

(1) approximately how long it takes for the control rods to be completely inserted, AND (2) the effect on related reactor trip functions?

Approximate Time Effect on Reactor Trip Functions

a.

2 seconds.

Reactor trip functions will occur normally.

b.

5 - 6 seconds.

Reactor trip functions will occur normally.

c.

2 seconds.

P-4 functions will be inhibited until additional manual actions are performed.

d.

5 - 6 seconds.

P-4 functions will be inhibited until additional manual actions are performed.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the candidate may believe that opening the MG set supply breakers causes rods to insert immediately. Opening the MG set supply breakers causes the MG set speed to decrease, and rods will insert after the voltage decays to a point where rod holding coils deenergize.

b.

Incorrect. Plausible, since opening the MG set supply breakers causes the MG set speed to decrease, and rods insert after the voltage decays to a point where rod holding coils deenergize.

c.

Incorrect. Plausible, since the candidate may believe opening the MG set supply breakers causes rods to insert immediately. The P-4 contacts will not reposition and therefore the reactor trip functions will not occur without additional manual actions.

d.

CORRECT. Opening the MG set supply breakers causes the MG set speed to decrease, and rods will insert after the voltage decays to a point where rod holding coils deenergize. The P-4 contacts will not reposition and therefore the reactor trip functions will not occur without additional manual actions.

K/A Applicability:

The question addresses the difference between opening rod drive MG set breakers and reactor trip breakers after an ATWS condition. It specifically tests the candidates understanding of how deenergizing the MG set differs in response to simply opening the reactor trip breakers during an ATWS event. The question also tests the impact of P-4 contacts NOT being repositioned, since the reactor trip breakers are still in the CLOSED position.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 10 of 77 Question Number: 9 K/A: 000038 EK1.03 Knowledge of the operational implications of the following concepts as they apply to the SGTR:

Natural circulation.

Tier:

1 RO Imp:

3.9 RO Exam:

9 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

9 Source:

INPO Bank Applicable 10CFR55 Section: (CFR 41.8 / 41.10 / 45.3)

Learning Objective:

3-OT-EOP0300, Obj. 8, Given a set of plant conditions, evaluate the conditions to determine if natural circulation exists and take appropriate action to initiate, restore, or maintain natural circulation.

References:

E-3, Steam Generator Tube Rupture, Rev. 22; WOG Background Document E-3, Rev. 2.

Question: 9 Why does the loss of reactor coolant pumps during a steam generator tube rupture increase the risk of voiding during the subsequent cooldown and depressurization?

a.

The upper head region becomes inactive and the fluid temperature in that region significantly lags the temperatures in the RCS loop.

b.

More ECCS flow is injected into the ruptured loop cold leg due to the reduced pressure, resulting in less flow to the core and less heat removal.

c.

The RCS hot legs reach saturation temperature during the rapid depressurization from the tube rupture, causing the RCS to flash.

d.

The isolation of the steam generator in the affected loop causes that loop to stagnate; therefore, insufficient heat removal capacity is available to cool the RCS.

DISTRACTOR ANALYSIS

a.

CORRECT. Under natural circulation conditions, the upper head region receives limited cooling flow. If the plant is depressurized rapidly, the risk of voiding is increased.

b.

Incorrect. Plausible, since SGTR is a type of small break LOCA. ECCS design, specifically the flow balancing valves in the injection lines, limits the amount of flow out of the break.

c.

Incorrect. Plausible, but the bulk RCS fluid is not susceptible to flashing; only the upper head region is susceptible.

d.

Incorrect. Plausible, since isolation of the S/G does cause a stagnant loop condition. This results in an increase in the potential for a PTS event, not upper head voiding.

K/A Applicability:

In the context of a SGTR, candidate is tested on how mitigation of that event (risk of voiding) is affected by having to use natural circulation.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 11 of 77 Question Number: 10 K/A: 000054 G2.2.22 Knowledge of limiting conditions for operations and safety limits.

Tier:

1 RO Imp:

4.0 RO Exam:

10 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

10 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.5 / 43.2 / 45.2)

Learning Objective:

3-OT-T/S0200, Obj. 3, Given a set of plant conditions, determine if a Safety Limit has been exceeded.

References:

Technical Specifications, Section 2.0 Safety Limits. Figure 2.1.1-1, Reactor Core Safety Limits.

Question: 10 During the performance of TI-127, "Reactor/Turbine Trip Report, Event Critique, Root Cause Analysis", data collected indicates the following SIMULTANEOUS readings occurred during an ATWS from an initial power level of 50%. A manual reactor trip was successful from the MCR.

- RCS pressure - 2520 psig.

- Reactor power - 52%.

- RCS T-hot - 695°F.

- RCS T-cold - 660°F.

- All RCPs are running.

Which (if any) Technical Specification Safety Limits were exceeded?

a.

NO safety limits were exceeded.

b.

ONLY the Reactor Core Safety Limit was exceeded.

c.

ONLY the RCS Pressure Safety Limit was exceeded.

d.

BOTH the Reactor Core and RCS Pressure Safety Limits were exceeded.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the given RCS pressure does not violate a Safety Limit. The Reactor Core Safety Limit was, however, exceeded.

b.

CORRECT. The combination of parameters places the plant in an UNACCEPTABLE OPERATION REGION on Figure 2.1.1-1 of Technical Specifications.

c.

Incorrect. Plausible if the candidate confuses design pressure of 2485 psig with the safety limit value of 2735 psig. Conditions in the stem do not indicate that RCS pressure exceeded the RCS Pressure Safety Limit.

d.

Incorrect. Plausible if the candidate confuses design pressure for the Safety Limit value. Conditions in the stem do not indicate that RCS pressure exceeded the 2735 psig RCS Pressure safety Limit. The Reactor Core Safety Limit HAS been exceeded.

K/A Applicability:

This question matches the K/A since it requires recall of and application of the Reactor Core and RCS Pressure Safety Limits.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 12 of 77 Question Number: 11 K/A: 000056 AK1.04 Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power:

Definition of saturation conditions, implication for the systems.

Tier:

1 RO Imp:

3.1 RO Exam:

11 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

11 Source:

Bank Mod.

Applicable 10CFR55 Section: CFR 41.8 / 41.10 / 45.3)

Learning Objective:

3-OT-EOP0000, Obj. 10, Given a set of plant conditions, determine if natural circulation is occurring in the RCS and identify actions required to establish natural circulation per ES-0.2.

References:

ES-0.2, "Natural Circulation Cooldown", Rev. 20, WOG Background Doc., ES-0.2, Rev 2.

Question: 11 A loss of 161kV offsite power has occurred and the reactor has tripped. A decision has been made to cooldown the plant by natural circulation. The highest available temperature indication for the RCS is 455° F.

To confirm that natural circulation is occurring, which one of the following is the MINIMUM RCS pressure allowed in accordance with ES-0.2, "Natural Circulation Cooldown"?

a.

1515 psig.

b.

948 psig.

c.

798 psig.

d.

435 psig.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since a cooldown to 550°F is accomplished followed by a depressurization. Later in ES-0.2, there is guidance to support the further depressurization; i.e., the subcooling margin is increased to 144°F from the initial value of 65°F. To calculate the required temperature/pressure, add 144°F + 455°F = 599°F. This equates to a saturation pressure of approximately 1530 psia, per the Steam Tables.

b.

Incorrect. Plausible since candidate may incorrectly use required subcooling value as 85° (the value if containment conditions were adverse).

c.

CORRECT. ES-0.2 requires at least 65° subcooling. Saturation pressure for 455°F = 444.33 psia. To obtain the minimum required subcooling (65°) add 65° + 455° = 520°. This equates to a saturation pressure of 812.53 psia per Steam Tables. The conversion from psia to psig (i.e., 812.53 - 14.7 = 798 psig) results in the correct answer.

d.

Incorrect. Plausible if the value used is slightly above saturation for the given RCS temperature.

K/A Applicability:

Candidate is tested on interpreting saturation conditions, including calculating a pressure from a given temperature, applying which amount of subcooling applies, and then determining the implications of this; i.e., the minimum pressure that the Reactor Coolant System can be reduced to.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 13 of 77 Question Number: 12 K/A: 000057 AA2.19 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus Tier:

1 RO Imp:

4.0 RO Exam:

12 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

12 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 43.5 / 45.13)

Learning Objective: 3-OT-AOI2500, Obj. 1, Demonstrate ability to recognize a loss of any 120V AC Vital Power Board, including effects on equipment and controls (SOER 81-02).

References:

AOI-25.01, "Loss of 120V AC Vital Instrument Power Boards 1-I and 2-I", Rev. 27.

Question: 12 Given the following plant conditions:

- The Unit is at 50% power.

- Pressurizer level channel 1-LT-68-335 (Channel II) was removed from service and the associated bistables have been tripped due to a problem during the previous shift.

- 120V AC Vital Instrument Power Board 1-I has just been lost.

Which ONE of the following is a consequence of the loss of 120V AC Vital Instrument Power Board 1-I?

a.

Automatic control of #2 SG Main Feed Reg. Valve is lost. Manual control of #2 SG Main Feed Reg.

Valve is needed in order to stabilize SG level.

b.

Automatic rod insertion occurs due to loss of the T-ref signal. Rod withdrawal in manual rod control will be used to stabilize reactor power.

c.

An automatic reactor trip is generated at the time of the failure. The TDAFW Pump is maintaining levels in SGs 1 and 2; 1B-B AFW Pump is maintaining SGs 3 and 4.

d.

An automatic reactor trip is generated at the time of the failure. Steam dumps receive an arming signal and open.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the failure does affect one of the Main Feed Reg. Valve controllers, but it does not affect #2, it affects #1.

b.

Incorrect. Plausible, since the Tref signal will be lost and rods will insert, but rod withdrawal is blocked by the failure of NIS Channel 1 functions.

c.

CORRECT. With the Channel II bistables already tripped, the board loss results in the Channel I PZR level trip and satisfies the coincidence for an automatic reactor trip. The board loss renders 1A AFW train inoperable (1-PDIC-3-122A, AFW pump A-A Discharge PCV is inoperable).

d.

Incorrect. Plausible, since the steam dumps normally operate on a reactor trip. However, the steam dump arming circuit is disabled and NO arming signal occurs. This renders the steam dumps inoperable in both Tavg and Steam Pressure modes.

K/A Applicability:

Stem conditions involve a loss of a vital ac electrical instrument bus, and then tests candidate's ability to determine the effects of the loss, and then interpret the significance from an operational perspective.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 14 of 77 Question Number: 13 K/A: 000058 AA1.02 Ability to operate and / or monitor the following as they apply to the Loss of DC Power:

Static inverter dc input breaker, frequency meter, ac output breaker, and ground fault detector.

Tier:

1 RO Imp:

3.1 RO Exam:

13 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

13 Source:

New Applicable 10CFR55 Section: (CFR 41.7 / 45.5 / 45.6)

Learning Objective:

3-OT-SYS235A Obj. 3, Describe the Low Voltage System in terms of: a. Purpose (Feeds), b. Description (Number of boards, their location, and their power supplies),c. Board or feed protection.

References:

Drawing 1-45W700-1, rev. 24; ARI 15-21.

Question: 13 The plant is at 100% power with all systems in normal alignment, when a failure of the fuse between 125V Vital Battery Board I and 120 VAC Vital Inverter 1-I occurs, causing the following alarm:

ARI-17-B, 125 DC VITAL BATT BD I ABNORMAL CKTS ISOLATED Which one of the following describes the effect on Vital Inverter 1-I, including operation of the static transfer switch, and whether entry into LCO 3.8.7, "Inverters-Operating", is required?

Static Transfer Switch Operation LCO 3.8.7 Entry

a.

Automatically bypasses the inverter in order to maintain a source of power to 120V AC Vital Instrument Power Board I.

NOT required

b.

Does NOT operate, and a source of power to 120V AC Vital Instrument Power Board will be maintained.

NOT required

c.

Automatically bypasses the inverter in order to maintain a source of power to 120V AC Vital Instrument Power Board I.

Required

d.

Does NOT operate, and a source of power to 120V AC Vital Instrument Power Board will be maintained.

Required DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the loss of DC is an abnormal condition, and candidate may believe that operation of the static transfer switch occurs to maintain a source of inverter output. This incorrect conclusion makes LCO entry NOT required plausible, since inverter output has been maintained, and candidate may then believe that the inverter is operable.

b.

Incorrect. Plausible because it is true that the static transfer switch does not operate, and is not required to operate to maintain power to the board; however, candidate fails to recognize that losing the DC power input to a vital inverter does require entry into LCO 3.8.7.

c.

Incorrect. Plausible, since the loss of DC is an abnormal condition, and candidate may believe that operation of the static transfer switch occurs to maintain a source of inverter output. Further plausibility is added by the fact that LCO 3.8.7 is required for the loss of DC input.

d.

CORRECT. Loss of DC input to the inverter is not a condition that causes actuation of the static transfer switch, since the normal AC input (480V Shutdown Board) to the inverter remains intact.

However, losing any power source input to a vital inverter renders the inverter inoperable, and requires LCO 3.8.7 entry to ensure the condition is corrected.

K/A Applicability:

Stem involves a loss of DC power, and question tests candidate's understanding of what this means in the context of which component is either expected to actuate (to compensate for the loss), or NOT actuate. In this case, since power is maintained to the board, no automatic actuation occurs. Further testing of the "monitor" aspect of this K/A is in the form of testing knowledge of HOW this affects the component; i.e., is it operable, and therefore is an LCO entry required.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 15 of 77 Question Number: 14 K/A: 000062 AA2.01 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:

Location of a leak in the SWS Tier:

1 RO Imp:

2.9 RO Exam:

14 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

14 Source:

Bank Applicable 10CFR55 Section: (CFR: 43.5 / 45.13)

Learning Objective:

3-OT-AOI1300, Obj. 3, Determine the general location of a rupture, given a supply header Hi flow and TURB/AUX/RX BLDG FLOODED alarm.

References:

AOI-13, "Loss Of Essential Raw Cooling Water", Section 3.1, Rev. 36.

Question: 14 Given the following plant conditions:

- The Unit is at 100% power.

- ERCW is in normal alignment.

- ERCW header 2A is indicating high flow.

The following MCR alarms are LIT:

- M-15B Window 167-D, "TURB/AUX/RX BDLG FLOODED".

- M-27A Window 223-A, "ERCW HDR A SUP PRESS LO".

- M-27A Window 223-B, "ERCW PMP A-A Discharge Pressure Low".

- M-27A Window 226-B, "ERCW PMP D-A Discharge Pressure Low".

- NO OTHER alarms are lit associated with the ERCW system.

Which ONE of the following ERCW conditions accounts for the above indications?

a.

A supply header has ruptured in the Auxiliary Building.

b.

A discharge header has ruptured in the Auxiliary Building.

c.

A supply header has ruptured upstream of the 2A strainer.

d.

A supply header has ruptured between the IPS and Auxiliary Bldg.

DISTRACTOR ANALYSIS

a.

CORRECT. The diagnostic section (Section 3.1 ) of AOI-13, "Loss of Essential Raw Cooling Water",

uses the annunciators and indications listed in the stem to indicate that a supply header has ruptured in the Auxiliary Building.

b.

Incorrect. Plausible, since the flooding alarm also indicates a discharge header rupture; however, the diagnostics section of AOI-13 indicates that ERCW flow will be normal for a discharge header rupture.

c.

Incorrect. Plausible since low header pressure is a symptom of a supply header rupture at this location, but the IPS, not the Auxiliary Building, will be flooded.

d.

Incorrect. Plausible since low header pressure is a symptom of a supply header rupture at this location, however, the strainer D/P alarm will be lit and the building flooded alarm is not expected. The building flooded alarm is possible, if the leak caused flow from the break to enter the auxiliary building through the concrete conduit.

K/A Applicability:

Stem conditions are given for a leak in the Emergency Raw Cooling Water System, and candidate is then tested on interpreting the indications to determine the location of the leak.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 16 of 77 Question Number: 15 K/A: 000065 AK3.03 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air:

Knowing effects on plant operation of isolating certain equipment from instrument air Tier:

1 RO Imp:

2.9 RO Exam:

15 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

15 Source:

New Applicable 10CFR55 Section: (CFR 41.5, 41.10 / 45.6 / 45.13)

Learning Objective:

3-OT-SYS002A, Obj. 19, State the purpose of the Short Cycle Valve (FCV-2-35) and describe how it is controlled.

References:

AOI-10, Loss of Control Air, Rev. 38, Appendix A Page 3 of 6, Non-Essential Air Components, System 2, Question: 15 The Unit is operating at 100% power when an air leak develops in the Turbine Building. The Turbine Building AUO reports that the leak is immediately upstream of the air operator for 1-FCV-2-35, Short Cycle Valve.

When the AUO isolates air to stop the leak 1-FCV-2-35 will ______(1)______, causing main feedwater pump speed to _________(2)________.

(1)

(2)

a.

OPEN RISE

b.

OPEN DROP

c.

CLOSE RISE

d.

CLOSE DROP DISTRACTOR ANALYSIS

a.

CORRECT. 1-FCV-2-35 fails open on loss of air. As 1-FCV-2-35 opens, less flow will be supplied to the condensate and feedwater systems. Overall system flow requirements remain unchanged, so the main feedwater pump speed must increase to compensate for the valve repositioning.

b.

Incorrect. Plausible, since 1-FCV-2-35 does fail open on loss of air. Candidate confuses the effect of the response on feed pump speed.

c.

Incorrect. Plausible, since the response of the main feedwater pump is correct. Candidate may confuse valve failure mode with somehow reducing feed flow, thereby requiring the feed pump speed to compensate by rising.

d.

Incorrect. Plausible, since the failure mode for many valves in the plant is CLOSED. If the candidate believes the failure response is correct, then the response of the main feedwater pump is also correct.

K/A Applicability:

Question stem lists conditions involving a loss of instrument air to a valve which affects feedwater.

Candidate is tested on why and how that loss affects plant operation.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 17 of 77 Question Number: 16 K/A: E04 EK2.1 Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

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

Tier:

1 RO Imp:

3.5 RO Exam:

16 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

16 Source:

New Applicable 10CFR55 Section: (CFR: 41.7 / 45.7)

Learning Objective:

3-OT-ECA0101, Obj. 1, Explain the major actions of procedures ECA-1.1 and 1.2.

References:

ECA-1.2, "LOCA Outside Containment", Rev. 4 Question: 16 Given the following plant conditions:

- Unit 1 is at 100% power.

- The Auxiliary Building AUO reports that the piping upstream of 1-FCV-63-93, RHR TO Cold Leg 2& 3 Injection Isolation, is hot to the touch.

- Pressurizer Relief Tank (PRT) temperature and pressure are rising.

Which one of the following accounts for the above conditions, including the upstream piping temperature and the PRT parameters?

a.

Cold Leg check valve leakage is occurring, resulting in lifting of the RHR Suction Relief valve 74-505.

b.

Cold Leg check valve leakage is occurring, resulting in lifting of the RHR Discharge Relief 63-626.

c.

1-FCV-63-93 is open, resulting in lifting of the RHR Suction Relief valve 74-505.

d.

1-FCV-63-93 is open, resulting in lifting of the RHR Discharge Relief 63-626.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since Cold Leg check valve leakage is occurring. Further, there are multiple check valves between the check valve leakage point and the RHR Suction Relief valve, making this selection incorrect, but plausible since the RHR Suction Relief valve does relieve to the PRT.

b.

CORRECT. Check valve leakage is occurring, and the path to the PRT is via the RHR Discharge Relief valve. The RHR system is rated for 700 psig, but is currently being exposed to RCS pressure, causing the relief to lift.

c.

Incorrect. Plausible, since candidate may believe that 1-FCV-63-93 being open is an abnormal condition; however, per ECCS normal alignment 1-FCV-63-93 is open, and therefore is not the source of any elevated temperatures. Also, the path to the PRT is NOT via the suction relief valve.

d.

Incorrect. Plausible, since candidate may believe that 1-FCV-63-93 being open is an abnormal condition; however, per ECCS normal alignment 1-FCV-63-93 is open, and therefore is not the source of any elevated temperatures. The correct path to the PRT is via the discharge relief valve.

K/A Applicability:

Candidate is presented with conditions involving a LOCA outside containment, via check valve leakby, into the RHR system, which relieves to the PRT. Candidate must then analyze this failure mode and apply knowledge of the system components and how they relate to each other (RHR relieves to the PRT),

specifically WHICH part of the RHR system is overpressurized as a result of this LOCA.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 18 of 77 Question Number: 17 K/A: E11 EA2.1 Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Tier:

1 RO Imp:

3.4 RO Exam:

17 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

17 Source:

New Applicable 10CFR55 Section: (CFR: 43.5 / 45.13)

Learning Objective:

3-OT-EOP0100,Rev.11, Objective 20, Discuss and justify the priority of usage given to procedure ES-1.3, "Transfer to RHR Containment Sump".

References:

ES-1.3, "Transfer to Containment Sump", Rev. 17 Question: 17 Given the following plant conditions:

- A Large Break LOCA is in progress on Unit 1.

- The 1A 6.9 KV Shutdown Board is damaged, resulting in the loss of power to all of its loads.

- After 40 minutes the crew observes the following annunciators and indications:

126-C RWST LEVEL LOW RECIRC INTLK annunciator LIT.

127-E CNTMT LEVEL HI RECIRC INTERLOCK annunciator LIT.

1-FCV-74-3, RHR Pump A Suction indicating lights are DARK.

1-FCV-74-21, RHR Pump B Suction RED light LIT.

1-FCV-63-72, Containment Sump to RHR Pump A Suction indicating lights are DARK.

1-FCV-63-73, Containment Sump to RHR Pump B Suction GREEN light LIT.

The OAC will place the handswitch for ____(1)_____ to OPEN, in accordance with _____(2)______.

(1)

(2)

a.

1-FCV-63-72 ECA-1.1 "Loss of RHR Sump Recirculation".

b.

1-FCV-63-73 ECA-1.1 "Loss of RHR Sump Recirculation".

c.

1-FCV-63-72 ES-1.3, "Transfer to Containment Sump".

d.

1-FCV-63-73 ES-1.3, "Transfer to Containment Sump".

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, if candidate fails to recognize that 1-FCV-63-73 has power available and could be opened. Use of ECA-1.1 is plausible since sump swapover has not occurred; however, candidate fails to realize that ES-1.3 contains guidance for manual alignment of sump suction valves if automatic function fails.

b.

Incorrect. Plausible, since the valve to be opened is correct, however, with the valve open, entry into ECA-1.1 is not required or appropriate.

c.

Incorrect. Plausible, because if 1-FCV-63-72 had power available, the OAC would open it per ES-1.3.

d.

CORRECT. Since 1A 6.9kV SD Board is deenergized 1-FCV-63-72 MOV has NO power. Opening 1-FCV-63-73 per ES-1.3 will provide suction to B RHR Pump from the containment sump.

K/A Applicability:

Candidate must determine that a loss of coolant recirculation has occurred, and then interpret other conditions to then make a decision on which valve manipulation will correct the condition, and which procedure is in effect.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 19 of 77 Question Number: 18 K/A: E12 G2.2.36 Excessive Heat Transfer Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Tier:

1 RO Imp:

3.1 RO Exam:

18 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

18 Source:

New Applicable 10CFR55 Section: (CFR: 41.10 / 43.2 / 45.13)

Learning Objective:

3-OT-T/S0307, Rev. 3, Obj. 3, Given plant conditions and parameters, correctly determine the OPERABILITY of components associated with different Plant Systems in Section 7 of Technical Specifications.

References:

T.S. 3.7.1, "Main Steam Safety Valves", Action A.

Question: 18 Consider the following four full power plant conditions which involve corrective actions taken to mitigate an excessive heat transfer event. Which ONE of these actions, if taken, results in a Technical Specification LCO entry?

a.

The Turbine Bldg. AUO manually isolates a leaking steam dump valve.

b.

Maintenance "gags" closed a leaking steam generator (SG) code safety valve.

c.

The controller for a failed open Feed Regulating Valve is placed to MANUAL to maintain SG level.

d.

A leak on the 1B Main Feed Pump high pressure steam supply valve is furmanited and then isolated.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since this is an excessive heat transfer condition, and there are numerous valves in the plant that if manually isolated, do require LCO entry. However in this case the candidate incorrectly believes the steam dumps are controlled by Tech. Specs.

b.

CORRECT. Any one inoperable SG code safety valve requires entry into T.S. 3.7.1, Action A.

c.

Incorrect. Plausible, since there are Tech. Specs. associated with FW Reg. valves, but taking manual control does not require LCO entry. A failed open FRV constitutes excessive heat transfer to excessive cooler feedwater being supplied to a SG.

d.

Incorrect. Plausible, since the steam leak constitutes an excessive heat transfer condition, but candidate incorrectly believes disabling of the steam supply valve requires LCO entry.

K/A Applicability:

In the context of an excessive heat transfer event, the candidate must be able to analyze each of four actions, any of which could be needed during this event, and then determine which one requires LCO entry.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 20 of 77 Question Number: 19 K/A: 000001 AA2.01 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

Determine/interpret reactor tripped brkr indicator.

Tier:

1 RO Imp:

4.2 RO Exam:

19 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

19 Source:

WBN Bank Mod Applicable 10CFR55 Section: (CFR: 43.5 / 45.13)

Learning Objective:

3-OT-AOI0200, Obj. 3, Describe Initial Actions for Continuous Rod Withdrawal.

References:

AOI-2, "Malfunction of Reactor Control System", Rev. 37 Question: 19 Given the following plant conditions:

- Unit 1 is at 14% power during a plant startup.

- After rod withdrawal of several steps, the IN-HOLD-OUT lever is returned to the HOLD position.

- The Operator-at-the-Controls observes that the RODS OUT light (RED) remains illuminated and outward rod motion continues at 8 steps/minute.

For the above conditions, the operator will.

a.

Place the IN-HOLD-OUT lever to the IN position and confirm rod motion stopped.

b.

Place the Rod Bank Selector in IND BANK SEL position and confirm rod motion stopped.

c.

Initiate a reactor trip, and then confirm that the reactor trip breaker GREEN lights are LIT and that the bypass breaker GREEN lights are LIT.

d.

Initiate a reactor trip, and then confirm that the reactor trip breaker GREEN lights are LIT and that the bypass breaker lights are DARK.

DISTRACTOR ANALYSIS

a.

Incorrect. Placing rods to "IN" is not the correct action directed by AOI-2. Plausible, since taking manual control is a typical action taken for various controller malfunctions in the plant.

b.

Incorrect. Placing rods to "IND BANK SEL" is not the correct action directed by AOI-2. Rods are already in MANUAL, therefore NO rod motion should be occurring without a demand from the IN-HOLD-OUT switch. Plausible since the given switch manipulation is a control associated with rod control, and because desiring to stop undesired rod motion seems appropriate.

c.

Incorrect. Plausible, since a reactor trip is required, but the method for determining reactor trip and bypass breaker position is incorrect. The bypass breaker indicating lights change state (and indicate GREEN) only if they were racked up. The bypass breakers are only racked up for testing, and this is not a condition given in the stem.

d.

CORRECT. A reactor trip will be initiated, and the method for determining reactor trip and bypass breaker position stated is correct. To correctly answer this question requires the candidate to recall that the reactor trip bypass breakers are racked up ONLY for special testing, and not during a plant startup.

Applying this knowledge, the candidate will then realize the reactor trip bypass breaker indicating lights will both be DARK.

K/A Applicability:

Candidate must determine that a continuous rod withdrawal is occurring, and after taking the appropriate action (which is to trip the reactor), must interpret indications for the reactor trip breakers.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 21 of 77 Question Number: 20 K/A:000003 G2.2.44 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.

Tier:

1 RO Imp:

2.9 RO Exam:

20 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

20 Source:

Modified Applicable 10CFR55 Section: 41.5 / 43.5 / 45.12 Learning Objective:

3-OT-AOI0200, Obj. 05, Discuss the Symptoms of a Dropped RCCA.

References:

AOI-2, "Malfunction of Reactor Control System", N3-85-4003, Control Rod Drive System.

Question: 20 Given the following plant conditions:

- With Unit 1 initially at full power, control rod M-12 drops fully into the core.

- AOI-2, "Malfunction of Reactor Control System", was entered and dropped rod recovery is in progress.

- Rod M-12 (in Bank D, group 2) has been withdrawn 30 steps, resulting in turbine load rising to 74%.

- To address the above conditions, operators have reconnected the lift coil(s) for the appropriate rod(s),

per AOI-2.

A CONTROL ROD URGENT FAILURE is LIT.

With 1-RBSS, Rod Bank Select in CBD (Control Bank D) position, which one of the following identifies the rod(s) that will move when rod control is operated to IN? [Assume NO other actions have been taken other than reconnecting the lift coil(s).]

a.

ONLY Group 1 Bank D rods.

b.

ONLY Group 2 Bank D rods.

c.

No rod motion will occur.

d.

All Bank D rods.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since candidate may believe that only the rod group (Group 1) with no dropped rod is used to maintain power stable. Candidate may also incorrectly think that the Control Rod Urgent Failure alarm was due to the dropped rod, which was in Group 2, and therefore, Group 1 is "OK" to use for stabilizing power. However, Group 1 Bank D rods are blocked from movement until the Control Rod Urgent Failure alarm is reset.

b.

CORRECT. In Individual Bank Select for Bank D, only Group 2 rods are capable of motion, since the Control Rod Urgent Failure alarm affected Power Cabinet 1BD (the power cabinet for Group 1 rods).

The Control Rod Urgent Failure alarm originated from the 1BD power cabinet. With all of the Group 1 lift coils disconnected, the 1BD sensed an Urgent Failure when M-12 rod was withdrawn. Reconnecting the lift coils will not reset the Control Rod Urgent Failure alarm on Power Cabinet 1BD.

c.

Incorrect. Plausible, since the Control Rod Urgent Failure alarm does block rod movement for one of the groups in Bank D (Group 1), but Group 2 Control Bank D rods will move on demand.

d.

Incorrect. Plausible, since candidate may fail to recall that the effect of the standing Control Rod Urgent Failure alarm is group specific - until this alarm is reset Group 1 rods are blocked from movement.

Group 2 rods WILL move, since they are on a separate power cabinet. An URGENT FAILURE exists on the 1BD power cabinet due to the previous rod withdrawal.

K/A Applicability:

Stem conditions involve a dropped control rod, and candidate must interpret various given conditions, including a key fact that an alarm has NOT been reset, and then determine how the subsequent operator action of moving rod control to IN affects the rods.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 22 of 77 Question Number: 21 K/A: AK1.01 Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions:

PZR reference leak abnormalities Tier:

1 RO Imp:

2.8 RO Exam:

21 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

21 Source:

Bank Applicable 10CFR55 Section: (CFR 41.8 / 41.10 / 45.3)

Learning Objective: 3-OT-AOI2000, Obj 2, Discuss the Result of specific PZR Level Channel failures.

References:

AOI-20, "Malfunction of Pressurizer Level Control System", Rev. 31.

Question: 21 With the plant at full power, a reference leg leak develops on the controlling PZR Level Transmitter.

As a result, the controlling PZR level channel will indicate slightly ______(1)______ than actual level, and remain ______(2)______ than the cold-calibrated PZR level instrument.

(1)

(2)

a.

higher lower

b.

lower lower

c.

higher higher

d.

lower higher DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the response of the controlling channel is correct. The cold calibrated level will indicate higher than the hot calibrated channel.

b.

Incorrect. Plausible if the candidate confuses the response of a variable leg leak on the controlling channel. The cold calibrated level will indicate higher than the hot calibrated channel.

c.

CORRECT. The cold calibrated pressurizer level instrument is calibrated for temperatures far lower than normal operating temperatures and will indicate lower. When the containment atmospheric temperature rises, the pressurizer reference leg heats up, causing density to decrease, and exerting less pressure on the reference leg side of the transmitter. This results in an increase in indicated level.

d.

Incorrect. Plausible, since the response of the controlling channel is incorrect, but the response of the cold calibrated channel is correct.

K/A Applicability:

Stem conditions are given for a leak in the PZR reference leg. Candidate must then determine the operational implication of the result, specifically effect on level indications.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 23 of 77 Question Number: 22 K/A:000033 AA2.01 Loss of Intermediate Range Nuclear Instrumentation Determine/interpret equivalency between source range, intermediate range, and power range channel readings.

Tier:

1 RO Imp:

3.0 RO Exam:

22 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

22 Source:

New Applicable 10CFR55 Section: 43.5 / 45.13 Learning Objective:

3-OT-SYS092A, Obj. 7, Discuss how each set of detectors overlap readings.

References:

Watts Bar System Description, N3-92-4003, Neutron Monitoring System, Rev. 8, Page 50 Question: 22 Given the following plant conditions:

- A plant startup is in progress, with the reactor currently at 8% power.

- The instrument power fuse for Intermediate Range Monitor N136A opens (blows).

Assuming the plant was in its normal alignment at the time of the failure, which ONE of the following statements describes the impact of the Intermediate Range Monitor (IRM) failure on plant operation, and the effect, if any, of the failure on the Source Range Monitors (SRMs)?

Impact of failure Effect on SRM

a.

Reactor trip.

When the operable IRM indicates below 1.66 x 10-4% power, the SRMs will automatically reinstate.

b.

Reactor trip.

When the operable IRM indicates below 1.66 x 10-4% power, the operator will place both SR Trip RESET-BLOCK switches to RESET.

c.

Operation will continue.

When the operable IRM indicates below 1.66 x 10-4% power, the SRMs will automatically reinstate.

d.

Operation will continue.

When the operable IRM indicates below 1.66 x 10-4% power, the operator will place both SR Trip RESET-BLOCK switches to RESET.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since a reactor trip occurs when the instrument power fuse opens, as long as the bypass switch is in the Normal position. The SRMs do NOT reenergize automatically when power drops below 1.66 x 10-4% power.

b.

CORRECT. A reactor trip signal is generated, and manual actions are required to reinstate the SRMs after power drops below 1.66 x 10-4% power.

c.

Incorrect. Plausible, since an instrument power fuse failure WITH the channel bypassed does NOT result in a reactor trip. The SRMs does NOT reenergize automatically when power drops below 1.66 x 10-4% power.

d.

Incorrect. Plausible, since an instrument power fuse failure WITH the channel bypassed does NOT result in a reactor trip, but manual actions are required in order to reinstate the SRMs after power drops below1.66 x 10-4% power.

K/A Applicability:

Stem conditions involve a loss of the intermediate Range NI. The Gamma Metrics Intermediate Range fully overlaps the Power Range monitors, and reads out in % power, so there is no equivalency associated with the Intermediate Range and Power Range.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 24 of 77 Question Number: 23 K/A: 000059 AK2.02 Knowledge of the interrelations between the Accidental Liquid Radwaste Release and the following:

Interrelations with radioactive gas monitors.

Tier:

3 RO Imp:

2.7 RO Exam:

23 Cognitive Level:

High Group:

n/a SRO Imp:

n/a SRO Exam:

23 Source:

Sig. WBN Bank Mod Applicable 10CFR55 Section: (CFR: 41.12 / 43.4 / 45.9)

Learning Objective:

3-OT-SYS090A, Obj. 27, Describe the Design Basis of the Process Radiation Monitoring System per FSAR 11.4.1.

References:

3-OT-SYS090A, Radiation Monitoring System.

Question: 23 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- A containment purge is in progress.

- A small leak develops on 1-FCV-62-69 Letdown Isolation valve bonnet.

- 174-A, CNTMT PURGE EXH 1-RM-130/131 RAD HI annunciator is LIT.

Which ONE of the following describes the expected purge system response, and the radiation monitor that is now INOPERABLE?

Purge System Response Inoperable Radiation Monitor

a.

Remains in service.

1-RM-90-106, Lower Containment Radiation Monitor

b.

Remains in service.

1-RM-90-400, Shield Building Vent Radiation Monitor

c.

Isolates.

1-RM-90-106, Lower Containment Radiation Monitor

d.

Isolates.

1-RM-90-400, Shield Building Vent Radiation Monitor DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the radiation monitor which is now inoperable is correct.

b.

Incorrect. Plausible, since the reason for the containment isolation is correct, but the incorrect radiation monitor is given.

c.

CORRECT. LP 3-OT-SYS090A, states that Hi Radiation initiates a Containment Vent Isolation signal which shuts down the Purge System so that for a small LOCA during purge, pressure relief through the purge system will not prevent initiation of a ØA Containment Isolation & SI on high containment pressure. When containment isolation occurs, 1-RM-90-106 is inoperable due to its flowpath being isolated.

d.

Incorrect. Plausible, since the correct radiation monitor, 1-RM-90-106 is listed as inoperable. The exhaust air is filtered to limit releases to the environment.

K/A Applicability:

An Accidental Liquid Radwaste Release is occurring due to the leak on the letdown isolation valve.

Candidate is then tested on not only what automatic action is initiated by the radiation monitor's response, but also on the effect (or the interrelationship between the monitor and the release) on the monitor itself.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 25 of 77 Question Number: 24 K/A: 000076 AK3.05 Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity :

Corrective actions as a result of high fission-product radioactivity level in the RCS.

Tier:

1 RO Imp:

2.9 RO Exam:

24 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

24 Source:

Bank Applicable 10CFR55 Section: (CFR 41.5,41.10 / 45.6 / 45.13)

Learning Objective: 3-OT-AOI2800, Obj. 4, Explain how activity is reduced if activated corrosion/erosion products are the reason for Hi activity.

References:

AOI-28, "High Activity in Reactor Coolant", Rev. 21.

Question: 24 Unit 1 has tripped from 100% power. The operators transitioned from E-0, "Reactor Trip or Safety Injection" to ES-0.1, "Reactor Trip Response". There are indications of failed fuel based on the results of post-trip activity samples. AOI-28, "High Activity in Reactor Coolant" is being implemented.

Which ONE of the following describes the correct actions to take with the CVCS system, including the reason for the action?

a.

Letdown is diverted to the CVCS Holdup Tank to limit radiation levels in the charging pump area.

b.

Excess Letdown is placed in service in order to increase overall flow from the reactor coolant system.

c.

Charging and Letdown are isolated to contain the high radioactivity within the containment building.

d.

Charging and Letdown flows are increased through the mixed bed demineralizers to maximize clean up.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since diverting to the Holdup Tank tends to contain the radioactive materials. This however causes an increase in the amount of liquid radioactive waste required to be processed, and would not be performed.

b.

Incorrect. Plausible, since placing excess letdown in service increases the amount of flow from the RCS, and provides some additional filtration using the Seal Water return filter. Maximizing letdown flow is accomplished by AOI-28; however, normal letdown is routed to the demineralizers.

c.

Incorrect. Plausible, since isolating charging and letdown would contain the radioactive material in the RCS, but it is not the prescribed action.

d.

CORRECT. AOI-28 directs the operators to place the CVCS mixed-bed demineralizer in service at 120 gpm flow. Both charging and letdown flows have to be increased to accomplish this.

K/A Applicability:

Candidate is presented with a high reactor coolant activity condition, and asked for what is the corrective action AND what is the reason for it.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 26 of 77 Question Number: 25 K/A: E08 EA1.3 Ability to operate and / or monitor the following as they apply to the (Pressurized Thermal Shock)

Desired operating results during abnormal and emergency situations.

Tier:

1 RO Imp:

3.6 RO Exam:

25 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

25 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.7 / 45.5 / 45.6)

Learning Objective: 3-OT-FRP0001, Obj. 6, Given a set of plant conditions, use procedure FR-P.1 or P.2 to identify any applicable cooldown and/or pressure limitations.

References:

FR-P.1 "Pressurized Thermal Shock", WOG Background Document FR-P.1, Rev. 2.

Question: 25 Given the following plant conditions:

- A Main Steam Line Break has occurred inside containment on Unit 1.

- Containment pressure is 5.5 psid.

- The crew has entered FR-P.1, "Pressurized Thermal Shock".

- An RCS pressure reduction is in progress.

- RCS subcooling is at 75°F.

Which one of the following describes the required action?

a.

Continue with the depressurization of the RCS.

b.

Dump steam from an intact S/G to raise subcooling.

c.

Start an additional charging pump to raise RCS subcooling.

d.

Close the PORV to stop RCS depressurization until subcooling is recovered.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since 75° subcooling is the value associated with normal containment conditions.

However, conditions given in the stem are adverse containment values, and 95° subcooling is the correct value for minimum required subcooling.

b.

Incorrect. A large cooldown has already occurred, and no further cooldown is allowed until after a soak has taken place. Plausible, since dumping steam would improve the subcooling, but restrictions to further cooldown apply.

c.

Incorrect. Plausible, since starting an additional charging pump will raise RCS pressure, and increase subcooling. A pressure increase is not desired due to the PTS concern.

d.

CORRECT. The step for RCS depressurization provides criteria for stopping the depressurization. The subcooling requirement for stopping the depressurization is <95°F subcooling (adverse containment values).

K/A Applicability:

Stem conditions include implementation of the procedure for Presssurized Thermal Shock. Candidate must evaluate these conditions and determine which of the listed choices will result in the desired operating result for mitigation of the PTS concern.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 27 of 77 Question Number: 26 K/A: E14 EK2.1 Knowledge of the interrelations between the (High Containment Pressure) and the following:

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

Tier:

1 RO Imp:

3.4 RO Exam:

26 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

26 Source:

Bank Applicable 10CFR55 Section: (CFR: 41.7 / 45.7)

Learning Objective:

3-OT-SYS063A, Obj. 25, Describe how to reset the safety injection signal, include P-4 interlock, also how and when to block the SI signal.

References:

1-47W611-63-1, Logic Diagram Question: 26 Given the following plant conditions:

- The plant was initially at full power.

- A failed Pressurizer safety valve resulted in an ATWS, Safety Injection (SI), and Phase B actuation.

- The reactor trip breakers could NOT be opened.

- The rod drive MG set supply breakers were successfully opened.

- After fifteen minutes, the crew is ready to reset SI per ES-1.1, "SI Termination".

- RCS pressure is 1930 psig and slowly recovering.

- Containment pressure is currently 1.8 psid and slowly lowering.

- All S/G pressures at 600 psig and slowly rising.

Which ONE of the following is true regarding resetting of the SI signal?

The SI signal...

a.

CAN be RESET, since the 90 second timer has timed out.

b.

CAN be RESET if BOTH of the reactor trip breakers are locally OPENED.

c.

CANNOT be RESET until containment pressure is less than SI setpoint.

d.

CANNOT be RESET until S/G pressures are greater than SI setpoint.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the SI signal can be reset after the 90 second timer has elapsed. But under the conditions of the stem, the reactor trip breakers are in the incorrect position to block the auto SI.

b.

CORRECT. Reactor trip breakers would have to be OPEN in order for the SI reset and Auto SI block functions to work properly.

c.

Incorrect. With containment pressure at 1.8 psid, a standing SI signal still exists. The reactor trip breakers are in the closed position, which prevents the auto SI block feature from functioning.

d.

Incorrect. Plausible, since a low S/G pressure does result in an SI actuation, and does prevent resetting a standing SI signal, but only IF both reactor trip breakers are still closed. Restoration of S/G pressures, by itself, does not allow resetting of the SI.

K/A Applicability:

Knowledge of the interrelations between the (High Containment Pressure) and the following:

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

A high containment pressure condition is given in the stem. Candidate must then evaluate numerous parameters involving signals, interlocks, and automatic features, including reactor trip breaker status, to exhibit knowledge of the interrelations between these conditions and the high containment pressure, to determine the effect on the ability to reset SI.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 28 of 77 Question Number: 27 K/A: E15 EK1.2 Knowledge of the operational implications of the following concepts as they apply to the (Containment Flooding): Normal, abnormal and emergency operating procedures associated with (Containment Flooding).

Tier:

1 RO Imp:

2.7 RO Exam:

27 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

27 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.8 / 41.10, 45.3)

Learning Objective: 3-OT-FRZ0001, Obj. 7, Identify all sources of water to the containment which might cause containment flooding.

References:

FR-Z.1, "Containment Flooding", WOG Background Doc. FR-Z.2 "Containment Flooding".

Question: 27 Given the following plant conditions:

- A large break LOCA has occurred.

- Accumulators have discharged and are isolated.

- ES-1.3, "Transfer to Containment Sump" has been completed.

- Containment sump level is now at 84% and slowly rising.

- The SM directs performance of FR-Z.2, "Containment Flooding".

FR-Z.2 requires that containment sump be sampled for activity and chemistry. Which one of the following describes (1) where the sample is taken from and (2) the reason for evaluating sump activity?

(1)

(2)

a.

RHR System To determine the source of the leak causing the level rise.

b.

Containment Sump To determine the source of the leak causing the level rise.

c.

RHR System To evaluate the possibility of transferring water from the sump to tanks outside of containment.

d.

Containment Sump To evaluate the possibility of transferring water from the sump to tanks outside of containment.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the sample is taken from the RHR System, but sampling for activity does not discriminate the source of the leak.

b.

Incorrect. Plausible, since containment sump would be the correct sample point if sump swapover had NOT been completed. Candidate fails to recognize this and believes sump sample is required.

c.

CORRECT. Since sump swapover has occurred, FR-Z.2 directs obtaining a sample from the RHR system to aid in determining if activity levels will allow transferring water to locations outside containment, to alleviate containment flooding.

d.

Incorrect. Plausible, since containment sump would be the correct sample point if sump swapover had NOT been completed. Further plausibility is added because the reason given is correct.

K/A Applicability:

With a containment flooding condition, the candidate is tested on procedural requirements for what sample must drawn, and how it is used from an operational implication perspective; i.e., what operation may be subsequently performed, based on the sample results.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 29 of 77 Question Number: 28 K/A: 003 K2.02 Knowledge of bus power supplies to the following:

CCW pumps Tier:

1 RO Imp:

2.5 RO Exam:

28 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

28 Source:

New Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective: 3-OT-SYS068B, Obj. 13, List and Explain the limitation for RCP operation without Component Cooling Water (CCS) aligned.

References:

System Description, N3-68-4001, Reactor Coolant System,; AOI-15, Loss of Component Cooling Water (CCS); SOI-70.01 Component Cooling Water (CCS) System, Checklist 1.

Question: 28 Which one of the following describes which 480 V SD Boards are the NORMAL and the ALTERNATE power supplies to C-S CCS pump?

NORMAL ALTERNATE

a.

1A1-A 2B2-B

b.

2A1-A 1B1-B

c.

2B2-B 1A2-A

d.

1A2-A 2B2-B DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the alternate source listed is one of the potential power supplies, but it is the NORMAL supply, not the alternate.

b.

Incorrect. Plausible, since the two listed sources have similar symmetrical separation as the corrrect answer, but train sources and unit sources are incorrect.

c.

CORRECT. Both listed sources are correct, per SOI-70.01 Electrical Checklists.

d.

Incorrect. Plausible, since both sources are actual sources available, but they are reversed (normal, alternate).

K/A Applicability:

Question tests in a straightforward manner, what are the power supplies to the pumps.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 30 of 77 Question Number: 29 K/A: 004 A2.05 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

RCP seal failures Tier:

2 RO Imp:

4.0 RO Exam:

29 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

29 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.5/ 43/5 / 45/3 / 45/5)

Learning Objective:

3-OT-SYS062A, Obj. 2, Explain the functions of the following subsystems of the CVCS system: Charging, letdown and seal injection water.

References:

AOI-24, "RCP Malfunctions During Pump Operation", Rev. 28.

Question: 29 Given the following plant conditions:

- The Unit is operating at 100%.

- The following alarms and indications are noted for #2 RCP:

- RCP #1 SEAL LEAKOFF FLOW HI.

- RCP 2 STANDPIPE LEVEL HI/LO.

- RCP #1 seal leak-off flow recorder indicates off-scale high.

- Charging flow has risen 40 gpm to maintain pressurizer level.

Based on these indications the ______(1)______ on #2 RCP has failed, and the operator will trip the reactor, then trip #2 RCP, and ______(2)_______.

(1)

(2)

a.

1. 1 seal close 1-FCV-62-61 and 1-FCV-62-63, Seal Water Return Isolation Valves.

b.

1. 2 seal close 1-FCV-62-61 and 1-FCV-62-63, Seal Water Return Isolation Valves.

c.

1. 1 seal after 3-5 minutes close 1-FCV-62-22, RCP 2 Seal Return Valve.

d.

1. 2 seal after 3-5 minutes close 1-FCV-62-22, RCP 2 Seal Return Valve.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since the correct seal failure is included, but the compensatory action is incorrect.

The compensatory actions are plausible since the closure of the listed valves does isolate seal return.

b.

Incorrect. Plausible, since the candidate must determine which seal has failed from the indications provided. The compensatory actions are plausible since the closure of the listed valves does isolate seal return.

c.

CORRECT. Indications listed are associated with a failure of the #1 Seal, and the action listed is taken from AOI-24.

d.

Incorrect. Plausible, since the candidate must determine which seal has failed from the indications provided. The action listed is taken from AOI-24, "RCP Malfunctions during Pump Operation".

K/A Applicability:

Candidate must be able to determine which RCP seal has failed and then by recalling which procedure is in effect, to determine which action is appropriate to mitigate the consequences.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 31 of 77 Question Number: 30 K/A: 004 K5.08 Chemical and Volume Control System Knowledge of the operational implications of the following concepts as they apply to the CVCS:

Estimation of subcritical multiplication factor (Keff) by means other than the 6-factor formula: relationship of count rate changes to reactivity changes Tier:

1 RO Imp:

2.6 RO Exam:

30 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

30 Source:

New Applicable 10CFR55 Section: (CFR: 41.5/ 45.7)

Learning Objective:

3-OT-GO0200, Obj. 1, Identify the reason for each prerequisite and precaution discussed in this lesson or provided in GO-2.

References:

GO-2, "Reactor Startup", 3.2.E.

Question: 30 A dilution to critical is being performed. Initial counts are 10 cps on both Source Range Channels N132 and N133. The first batch of primary water results in count rate changing from 10 cps to 20 cps. If the count rate is doubled again, the volume of primary water required for the doubling will be _____(1)_____ than the first batch, and as a result the reactor will then be _______(2)_________.

(1)

(2)

a.

greater than subcritical.

b.

less than subcritical.

c.

greater than critical.

d.

less than critical.

DISTRACTOR ANALYSIS Assume initial Keff is 0.98.

Using equations ((CR1) (1-Keff 1)) = ((CR2)(1-Keff2)) and = 1-Keff/Keff

((CR1) (1-Keff 1)) = ((CR2)(1-Keff2))

Keff2 = 1- ((10)(1-0.98))/20 = 0.99 After the first batch of primary water was added, Keff was increased to 0.99.

Change in reactivity from 0.98 to 0.99 was +0.010307 K/K Assessing the second batch:

((CR1) (1-Keff 1)) = ((CR2)(1-Keff2))

Keff2 = 1- ((20)(1-0.99))/40 = 0.995 After the second batch of primary water was added, Keff was increased to 0.995 Change in reactivity from 0.99 to 0.995 was +0.005076 K/K The amount of reactivity to achieve the second doubling was less than the amount of reactivity to achieve the first doubling, and since Keff2 is 0.995, the reactor remains subcritical.

a.

Incorrect. See proof above. Plausible if the candidate does not properly recall the relationship between count rate and the change in Keff.

b.

CORRECT. See proof above.

c.

Incorrect. See proof above. Plausible, since the candidate may confuse the concept of count rate doubling, where after a doubling occurs, if the SAME AMOUNT OF REACTIVITY IS ADDED AGAIN, THE REACTOR WILL BE CRITICAL.

d.

Incorrect. See proof above. Plausible, since the candidate may confuse the concept of count rate doubling, where after a doubling occurs, if the SAME AMOUNT OF REACTIVITY IS ADDED AGAIN, THE REACTOR WILL BE CRITICAL.

K/A Applicability:

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 32 of 77 The core of the concept being tested is generic fundamentals in nature. However, the K/A also asks for the operational implications of the concept. This is addressed in the question by, 1.) having the candidate evaluate how much dilution is needed, and 2.) what is the effect on the reactor.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 33 of 77 Question Number: 31 K/A:005 K3.06 Knowledge of the effect that a loss or malfunction of the RHRS will have on the following:

CSS (containment spray system)

Tier:

2 RO Imp:

3.1 RO Exam:

31 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

31 Source:

Bank Applicable 10CFR55 Section: (CFR: 41.7 / 45.6)

Learning Objective:

3-OT-SYS072A, Obj. 19, Identify the basis, requirements and interlocks required to place the RHR spray in service.

References:

E-1, Loss of Reactor or Secondary Coolant.

Question: 31 Which one of the following conditions will result in a less than adequate functioning of RHR Spray to control containment pressure after a large break LOCA? (Evaluate each condition separately.)

a.

1B RHR pump trips on instantaneous overcurrent.

b.

1A CCP pump trips on instantaneous overcurrent.

c.

1-FCV-74-41, RHR SPRAY HDR B ISOLATION, will NOT OPEN.

d.

1-FCV-74-33 AND 1-FCV-74-35, RHR Crosstie Valves, CANNOT be CLOSED.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since this renders the 1B RHR train incapable of spray flow. This would NOT prevent alignment of the 1A RHR train to supply spray flow, assuming all other conditions were satisfied.

b.

Incorrect. Plausible, since the conditions required to be satisfied for placing RHR spray in service require at least 1 charging and 1 SI pump running. With no additional information provided the candidate should assume all other pumps are running.

c.

Incorrect. Plausible, since this renders the 1B RHR train incapable of spray flow. This would NOT prevent alignment of the 1A RHR train to supply spray flow, assuming all other conditions were satisfied.

d.

CORRECT. With the RHR trains incapable of being split, ES-1.3, "Transfer to RHR Containment Sump", actions would not be complete. This means that RHR suction is not aligned to the containment sump, one of the conditions for placing RHR spray in service. If only ONE cross-connect valve had failed, the other RHR train could be aligned.

K/A Applicability:

The correct answer involves a malfunction of RHR components. Candidate must then understand the effect of that malfunction on the containment spray aspect of RHR design.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 34 of 77 Question Number: 32 K/A: 006 A4.05 Ability to manually operate and/or monitor in the control room:

Transfer of ECCS flowpaths prior to recirculation Tier:

2 RO Imp:

3.9 RO Exam:

32 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

32 Source:

New Applicable 10CFR55 Section: (CFR: 41.7 / 45.5 to 45.8)

Learning Objective:

3-OT-SYS063A, Obj. 8, Explain the RHR suction valve logic from the RWST.

References:

ES-1.3, Transfer to Containment Sump, Rev. 17; 1-47W611-63-2, 63-5 ECCS Logic Diagrams.

Question: 32 Given the following plant conditions:

- Safety Injection (SI) is actuated.

- RWST level is 32%.

- Containment sump level is 18%.

Based on these conditions, it is expected that sump swapover will

a.

NOT occur since containment sump level is too low.

b.

occur, but ONLY if manual actions are taken.

c.

NOT occur since RWST level is too high.

d.

occur, since all required conditions have been met.

DISTRACTOR ANALYSIS

a.

Incorrect. Containment sump level is greater than the required 16.1% to satisfy the logic for swapover.

b.

Incorrect. Plausible, since the lights above the handswitches are lit when an SI signal is present.

c.

Incorrect. RWST level is less than the required 34.63% to satisfy the logic for swapover.

d.

CORRECT. RWST level, containment sump level and the presence of the SI signal satisfy all conditions for the swapover signal to occur.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 35 of 77 Question Number: 33 K/A:007 K4.01 Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following:

Quench tank cooling Tier:

2 RO Imp:

2.6 RO Exam:

33 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

33 Source:

New Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective: 3-OT-SYS068C, Obj. 21, Describe the flow path of sources of supply, discharges, vents, drains, leakoff, and connections/penetrations that intertie this system to other systems.

References:

System Description N3-68-4001, Reactor Coolant System, Rev. 25.

Question: 33 The pressurizer relief tank (PRT) is designed to condense and cool a discharge of steam equal to...

a.

100% of the PZR steam volume above 25% level.

b.

110% of the PZR steam volume above 60% level.

c.

100% of the PZR steam volume above 60% level.

d.

110% of the PZR steam volume above 25% level.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the candidate may not recall that the steam volume is based on full power level in the PZR.

b.

CORRECT. Per the RCS system description, the tank design is based on the requirement to condense and cool a discharge of PZR steam equal to 110% of the volume above the full-power PZR water level set point. The volume of water in the tank is capable of absorbing the heat from the assumed discharge.

c.

Incorrect. Plausible, since the initial level is correct, but percent of steam volume given is incorrect.

d.

Incorrect. Plausible, since the percent of steam volume given is correct, but the initial level is incorrect.

K/A Applicability:

Tests knowledge of PRT capacity design features for accomplishing one of its functions (maintain the tank cooled during a steam discharge).

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 36 of 77 Question Number: 34 K/A: 007 K5.02 Knowledge of the operational implications of the following concepts as they apply to PRTS: Method of forming a steam bubble in the PZR Tier:

2 RO Imp:

3.1 RO Exam:

34 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

34 Source:

New Applicable 10CFR55 Section: (CFR: 41.5 / 45.7)

Learning Objective:

3-OT-GO0100, Obj. 5, Describe the basic steps necessary to establish a steam bubble in the Pressurizer (PZR) with or without a Nitrogen blanket.

References:

GO-1, Unit Startup From Cold Shutdown to Hot Standby, Rev. 60. pp. 140-144; SOI-68.01, Reactor Coolant System Pressurizer Relief Tank Operations, Rev. 46.

Question: 34 In accordance with GO-1, "Unit Startup From Cold Shutdown To Hot Standby", the Shift Manager has directed the crew to establish a steam bubble in the pressurizer (PZR) with a nitrogen blanket present. The flow path from the PZR to the pressurizer relief tank (PRT) is required to be closed when...

a.

PZR level reaches 30%.

b.

PRT pressure reaches 6 psig.

c.

PRT temperature rises to 112°F.

d.

PZR Liquid Temperature reaches 235°F.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since this is the upper limit of the desired PZR level during bubble formation.

b Incorrect. Plausible, since the pressure in the PRT will be maintained between 6.5 an 8.0 psig.

c.

Incorrect. Plausible, since PRT temperature is a concern and would be rising as PZR temperature increases. Also, 112° is a number listed in GO-1 associated with PRT parameters.

d.

CORRECT. Per GO-1, this is the specific indication to be used for evaluating when to close the PZR PORVs, with a temperature band of 230-240°F.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 37 of 77 Question Number: 35 K/A: 008 G2.4.2 Component Cooling System Knowledge of system setpoints, interlocks, auto actions associated with EOP entry conditions.

Tier:

2 RO Imp:

RO Exam:

35 Cognitive Level:

Low Group:

1 SRO Imp:

SRO Exam:

35 Source:

New Applicable 10CFR55 Section:

Learning Objective:

3-OT-SYS070A, Obj. 19, Given a set of plant conditions, determine the correct response of the CCS system.

References:

AOI-15, Loss of Component Cooling System (CCS), Rev. 31.

Question: 35 Which one of the following component cooling system conditions would result in implementation of E-0, "Reactor Trip or Safety Injection"?

a.

Thermal barrier heat exchanger flow path isolates due to high differential flow.

b.

The Letdown Heat Exchanger develops a tube leak.

c.

ERCW flow is lost to CCS Heat Exchanger A.

d.

CCS pump 1A flow exceeds 6800 gpm.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the isolation affects RCP parameters, but with seal injection still available, the plant can remain at power.

b.

Incorrect. Plausible, since the leak requires isolation of the normal letdown and charging. Chemistry parameters may degrade to the point that a unit shutdown is required but a reactor trip is not warranted.

c.

CORRECT. Per AOI-15, "Loss of Component Cooling Water", step 6, a loss of ERCW flow to CCS heat exchanger A requires the crew to trip the reactor (sub step 6.e.).

d.

Incorrect. Plausible, since this value is the maximum flow allowed through a CCS pump and may require the crew to stop the pump. This does not require a reactor trip to be initiated.

K/A Applicability:

Tests knowledge of which component cooling water system component parameter value (i.e., 0 gpm of cooling water flow) requires entry into E-0.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 38 of 77 Question Number: 36 K/A: 010 K6.04 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PRT Tier:

2 RO Imp:

2.9 RO Exam:

36 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

36 Source:

INPO Bank Point Beach 2003 Applicable 10CFR55 Section: (CFR: 41.7 / 45.7)

Learning Objective: 3-OT-SYS068C, Obj. 11: Describe the indication an operator has that a PORV is open or leaking through.

References:

INPO Bank question; ARI 88-C, "PRT Press Hi".

Question: 36 Unit 1 is operating at 100% power when the following sequence of events occurs:

- PZR PORV 1-PCV-68-334 opens and sticks open.

- PZR PORV Block valve, 1-FCV-68-332 cannot be closed.

- The reactor trips.

- Safety Injection actuates.

- PRT pressure rises to the point that the PRT rupture disc ruptures.

Which of the following is an effect of the disc rupturing?

a.

N2 header pressure lowers.

b.

H2 concentration in containment lowers.

c.

PRT level drains below the sparging nozzles.

d.

Pressurizer Relief Valve Outlet temperature lowers.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible if candidate mistakenly believes that because of the interface between the PRT and the nitrogen supply header, when pressure suddenly reduces (from the rupture disc blowing),

nitrogen header pressure is also affected.

b.

Incorrect. Plausible, if candidate misapplies the concept of recombination of hydrogen in containment when exposed to the inventory released from the PRT.

c.

Incorrect. Plausible, since there is a level maintained in the tank, and if candidate believes that the liquid inventory flashes to steam and reduces the level, however, this inventory is not at saturated conditions.

d.

CORRECT. Per the TMI lessons learned, and basic thermodynamic principles, this temperature will be lower.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 39 of 77 Question Number: 37 K/A: 012 A2.01 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 bistable operation.

Tier:

2 RO Imp:

3.1 RO Exam:

37 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

37 Source:

New Applicable 10CFR55 Section: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

Learning Objective:

3-OT-SYS099A, Rev. 6, Objective 8, Briefly discuss the input relays, Logic Section and Output Section of the SSPS.

References:

Watts Bar System Description, N3-99-4003 Reactor Protection System, Rev. 21; AOI-44, Eagle 21 Malfunctions, Rev. 1.

Question: 37 With the plant operating at 90% power, what is the impact of the BISTABLES in NIS Power Range Channel N41 drawer failing to the tripped condition, and what actions will be directed by procedures in response to this condition?

Impact of Failed Bistable Actions per Procedure

a.

Control Rods insert at maximum speed and a 2/4 SSPS trip is processed.

AOI-4, "Nuclear Instrumentation Malfunctions,"

directs placing rod control in MANUAL, and then to slowly restore rods to pre-event position.

b.

Rod withdrawal is blocked and a 1/4 SSPS trip is processed.

AOI-2, "Malfunction of Reactor Control System,"

directs placing rod control in MANUAL and then reducing turbine load to match Tave and Tref.

c.

Control Rods insert at maximum speed and a 2/4 SSPS trip is processed.

AOI-2 directs placing rod control in MANUAL and then reducing turbine load to match Tave and Tref.

d.

Rod withdrawal is blocked and a 1/4 SSPS trip is processed.

AOI-4, directs defeating the rod withdrawal block to permit AUTO and MANUAL rod withdrawal.

DISTRACTOR ANALYSIS

a.

Incorrect. Although the high power level bistable will be tripped, there will be no high power signal sent to the rod control circuit. Plausible, since the control rods would insert at maximum speed for a power range channel failure, and rods would be taken to manual per AOI-4, Step 1. AOI-4 Step 8 supports the actions to match Tavg to Tref prior to returning rods to AUTO.

b.

Incorrect. Plausible, since the candidate may believe that a high power condition exists. If it did, then these actions would be appropriate and accomplished using AOI-2.

c.

Incorrect. Plausible, since the candidate may believe that a high power condition exists. If it did, then these actions would be appropriate and accomplished using AOI-2.

d.

CORRECT. The Overpower Rod Stop bistable would be tripped, and since this is a 1/4 coincidence, the rod stop would have to be defeated to regain the ability to withdraw rods. This action would be appropriate and accomplished using AOI-4, Attachment 1.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 40 of 77 Question Number: 38 K/A: 012 G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Tier:

2 RO Imp:

3.8 RO Exam:

38 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

38 Source:

New Applicable 10CFR55 Section: (CFR: 41.10 / 43.5 / 45.13)

Learning Objective:

3-OT-EOP0100, Obj 8, Given a set of plant conditions, use E-1, ES-1.1, ES-1.2, ES-1.3, and ES-1.4 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes, and Cautions.

References:

ES-1.1, SI Termination, Rev. 15; SEN 268, Invalid Safety Injection with Failure to Reset, 8/17/2007.

Question: 38 Which one of the following describes how the Operator-at-the-Controls will determine the TRAIN of SI that failed to RESET, and the local action that will by itself, RESET the SI signal.

Indication of SI train that failed to RESET Local Operator Actions

a.

ICS points SD9000, SI ACT TRAIN A and/or SD9001, SI ACT TRAIN B will display ACT in RED letters on the train that failed to reset.

Open both 48v breakers in panel 1-R-47 (Train A)

OR 1-R-50 (Train B).

b.

ICS points SD9000, SI ACT TRAIN A and/or SD9001, SI ACT TRAIN B will display ACT in RED letters on the train that failed to reset.

Place 1-R-51 or 1-R-52 Safeguards Test Cabinet RESET switch to RESET and release.

c.

SI ACTUATED permissive will be LIT, AUTO SI BLOCKED permissive will be DARK.

Open both 48v breakers in panel 1-R-47 (Train A)

OR 1-R-50 (Train B).

d.

SI ACTUATED permissive will be LIT, AUTO SI BLOCKED permissive will be DARK.

Place 1-R-51 or 1-R-52 Safeguards Test Cabinet RESET switch to RESET and release.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the first half of the distractor is correct. Opening both 48v breakers locally requires the OAC to reset the SI from the 1-M-6 panel.

b.

CORRECT. The ICS computer points listed are correct, and the red ACT on the computer informs the operator that the associated SI train is still actuated. Per Appendices G and H, placing the Safeguards Cabinet RESET switch to RESET should cause the SI ACTUATED annunciator to become DARK and the AUTO SI BLOCKED annunciator to be LIT.

c.

Incorrect. Plausible, since the candidate may believe that the annunciators will allow diagnosis of the failed SI train. The local action provided is taken from Appendices G and H, but will not by itself result in SI being RESET. Opening both 48v breakers locally requires the OAC to reset the SI from the 1-M-6 panel.

d.

Incorrect. Plausible, since the candidate may believe that the annunciators will allow diagnosis of the failed SI train. Per Appendices G and H, placing the Safeguards Cabinet RESET switch to RESET should cause the SI ACTUATED annunciator to become DARK and the AUTO SI BLOCKED annunciator to be LIT.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 41 of 77 Question Number: 39 K/A: 013 K2.01 Knowledge of bus power supplies to the following:

ESFAS/safeguards equipment control Tier:

2 RO Imp:

3. 6 RO Exam:

39 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

39 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective:

3-OT-SYS063A, Rev, 10, Objective 21, Identify and explain the initiation signals to the ECCS, include setpoints.

References:

3-OT-SYS099A, Reactor Protection System. Rev. 6, Appendix C; N3-99-4003, Reactor Protection System, Rev. 21.

Question: 39 Given the following plant conditions:

- The operating crew is responding to a reactor trip due to a loss of 120V AC Vital Instrument Power Board I-I.

- PZR pressure transmitter 68-334 (Channel II) fails LOW.

Which ONE of the following describes the plant response?

a.

Both trains of SSPS SI master relays would actuate AND both trains of ECCS equipment auto start.

b.

Both trains of SSPS SI master relays would actuate BUT only "B" train ECCS equipment auto start.

c.

Only the "B" train SSPS SI master relays would actuate BUT both trains of ECCS equipment auto start.

d.

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

DISTRACTOR ANALYSIS

a.

Incorrect. Master relays would actuate since they are fed by the power supplies to the logic bay but slave relays for train A ECCS equipment cannot energize to actuate. Examinee may think the safeguards panel has redundant power supplies also.

b.

CORRECT. Both trains of master relays will actuate since they are fed by the redundant power supplies of the logic bay, but A train ECCS will not start since only channel I supplies power to the safeguards panel to actuate A train slave relays.

c.

Incorrect. Examinee may be confused as to which relays have redundant power supplies, and may think the master relays do not have redundant power supplies - the slave relays do.

d.

Incorrect. Examinee may confuse the power supplies of master relays and think they are powered by the same power supply as the safeguards panel.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 42 of 77 Question Number: 40 K/A: 022 A3.01 Ability to monitor automatic operation of the CCS, including:

Initiation of safeguards mode of operation Tier:

2 RO Imp:

4.1 RO Exam:

40 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

40 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.7 / 45.5)

Learning Objective:

3-OT-SYS030D, Obj. 3, Describe the automatic start signal for the air return fans.

References:

1-47W611-30-3 Logic Diagram.

Question: 40 Given the following plant conditions:

- Reactor trip and Safety Injection occurred at 0340 due to a LOCA inside Containment.

- Phase B Containment Isolation actuated at 0347.

Which ONE of the following describes the status of the Containment Air Return System at 0352?

Air Return Fans Air Return Backdraft Dampers

a.

RUNNING CLOSED

b.

RUNNING OPEN

c.

OFF OPEN

d.

OFF CLOSED DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since a Phase B signal is present, but only 7 minutes have elapsed since the actuation and the logic circuit requires 9 minutes. Backdraft dampers will be open if the fan was running.

b.

Incorrect. Plausible since a Phase B signal is present, but only 7 minutes have elapsed since the actuation and the logic circuit requires 9 minutes. Backdraft dampers will be open if the fan was running.

c.

Incorrect. Plausible, since insufficient time has elapsed to provide a start signal. Backdraft dampers will be closed under these conditions.

d.

CORRECT. Insufficient time has elapsed AND the backdraft dampers are closed under the stated conditions.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 43 of 77 Question Number: 41 K/A: 025 K1.02 Knowledge of the physical connections and/or cause/effect relationships between the ice condenser system and the following systems: Refrigerant systems Tier:

2 RO Imp:

RO Exam:

41 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

41 Source:

WBN Bank Mod Applicable 10CFR55 Section: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Learning Objective:

3-OT-SYS061A, Obj. 16, Describe the logic for the glycol containment isolation valves.

References:

System Description N3-61-4001, Ice Condenser System, Section 3.1.

Question: 41 Which ONE of the following conditions will directly cause the refrigerant system to be isolated from the glycol piping in the floor of the Ice Condenser?

a.

Containment Phase A actuation.

b.

Containment Phase B actuation.

c.

A valid ICE COND FLOOR COOLANT P LO alarm annunciates.

d.

A valid GLYCOL EXP TNK LEVEL HI/HI-HI alarm annunciates.

DISTRACTOR ANALYSIS

a.

CORRECT. Per System Design document, Phase A actuation isolates 1-FCV-61-191, and 1-FCV 193 (Glycol inlet piping to Containment).

b.

Incorrect. Plausible, since a Phase B is another, more extensive containment isolation; however, Phase A signal is the only isolation signal for the glycol valves.

c.

Incorrect. Plausible, since this condition is related to an isolation of the glycol system to containment.

The given alarm would actually result from an isolation of the glycol system from containment, not cause the isolation. Plausible, since this could be a misconception on the cause and effect aspect of this topic.

d.

Incorrect. Plausible, since there is an alarm condition associated with the Glycol Expansion Tank which does cause isolation of the glycol system from containment, but it is Glycol Expansion Tank LO LO level, not HI HI level.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 44 of 77 Question Number: 42 K/A: 025 K4.02 Knowledge of ice condenser system design feature(s) and/or interlock(s) which provide for the following:

System control Tier:

2 RO Imp:

2.8 RO Exam:

42 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

42 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective:

3-OT-SYS061A, Rev, 2, Objective 15, Describe the glycol pumps, include power supply, logic and capacity.

References:

1-47W760-61-1.

Question: 42 Which ONE of the following lists 2 conditions that will DIRECTLY cause a glycol circulating pump to trip?

a.

Low expansion tank level, Low glycol temperature.

b.

Low expansion tank level, High discharge pressure.

c.

Low suction pressure, High discharge pressure.

d.

Low suction pressure, Low glycol temperature.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since a low expansion tank level causes isolation of 1-FCV-61-191 and 1-FVC 193. The closure of these valves results in a trip of the glycol circ pumps and the chillers INDIRECTLY.

Low glycol temperature trips the chillers DIRECTLY, and results in an INDIRECT trip of the glycol circ pumps due to high discharge pressure.

b.

Incorrect. Plausible, since a low expansion tank level causes isolation of 1-FCV-61-191 and 1-FVC 193. High discharge pressure causes a DIRECT trip of the glycol circ pumps.

c.

CORRECT. Both high discharge pressure and low suction pressure trip the pumps.

d.

Incorrect. Low suction results in a DIRECT trip of the pumps, however low glycol temperature trips the chillers DIRECTLY, and results in an INDIRECT trip of the glycol circ pumps due to high discharge pressure.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 45 of 77 Question Number: 43 K/A: 026 K4.09 Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following:

Prevention of path for escape of radioactivity from containment to the outside (interlock on RWST isolation after swapover)

Tier:

2 RO Imp:

3.7 RO Exam:

43 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

43 Source:

New Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective:

3-OT-SYS072A, Rev. 7, Objective 8, Describe the logic (interlocks) on the Containment Spray suction, discharge header, and containment sump valves.

References:

WBN System Description N3-72-4001, Rev. 18; 1-SI-72-901-B, Containment Spray Pump 1b-B Quarterly Performance Test, Rev. 17.

Question: 43 Given the following conditions:

- A large break LOCA occurred.

- Containment pressure is 3.5 psid.

- RWST level is 7%.

- ES-1.3, "Transfer to RHR Containment Sump", is being performed.

FCV-72-21, RWST Suction to the 1B CS Pump, failed to close.

For the above conditions, which one of the following describes the required action?

a.

Open 1-FCV-72-45, 1B CS Sump Suction, from the MCR because the interlock only functions in the AUTO swapover circuit.

b.

Open 1-FCV-72-45, 1B CS Sump Suction, from the Reactor MOV board because the Aux. circuit bypasses the interlock logic.

c.

Manually close 1-FCV-72-21, RWST Suction to the 1B CS Sump, then open 1-FCV-72-45, 1B CS Sump Suction.

d.

Align only the A Train of CS, since 1-FCV-72-45 must be locally operated and it is located inside containment.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the given operation for 1-FCV-72-45 is correct; however, candidate incorrectly believes there an auto swapover circuit.

b.

Incorrect. Plausible, since there are components in the plant where use of the aux. circuit does bypass the logic; however, candidate misapplies the concept here.

c.

CORRECT. Per procedure, this is the correct desired operation.

d.

Incorrect. Plausible, since candidate recognizes that an alternate method must be used, but misapplies it here.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 46 of 77 Question Number: 44 K/A: 039 A3.02 Ability to monitor automatic operation of the MRSS, including:

Isolation of the MRSS Tier:

2 RO Imp:

3.1 RO Exam:

44 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

44 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.5 / 45.5)

Learning Objective:

3-OT-SYS001A, Obj. 21, List the automatic closure signals for the MSIVs.

References:

1-47W611-1-1, 1-47W611-63-1 and 1-47W611-88-1 Logic Diagrams.

Question: 44 Given the following plant conditions:

A reactor heatup is in progress, per GO-1, "Unit Startup From Cold Shutdown To Hot Standby Reactor Startup".

- RCS pressure is stable at 1600 psig.

Which of the following will automatically close the main steam isolation valves (MSIVs)?

a.

Steam line pressure has lowered to 700 psig.

b.

Containment pressure 2.6 psid on 2/4 channels.

c.

Containment pressure 1.6 psid on 2/3 channels.

d.

A rapid steam line pressure drop of 100 psig in 5 seconds.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the logic and coincidence for the actuation of the Low Steam Header Pressure SI are correct, and the setpoint of 675 psig has not been reached.

b.

Incorrect. Plausible, since the measured parameter and coincidence logic are correct, but the given value is incorrect.

c.

Incorrect. Plausible, since this is the correct setpoint and logic for the High Containment Pressure SI.

d.

CORRECT. The plant is greater than P-11, so the steam line pressure rate circuit is not enabled, but a pressure drop of this magnitude satisfies the Low Steam Line pressure SI and MSIV closure logic. This signal is lead-lag compensated. The signal is boosted or amplified to make the bistable think the pressure is lower than actual. The higher the rate of pressure drop, the quicker the setpoint is reached.

Any sudden drop in steam pressure is amplified 10 times by the circuit.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 47 of 77 Question Number: 45 K/A: 059 A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Overfeeding event Tier:

2 RO Imp:

2.7 RO Exam:

45 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

45 Source:

New Applicable 10CFR55 Section: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Learning Objective:

3-OT-SYS003A, Objective 6, Describe the MFPT Speed Control Program; 3-OT-SYS003D, Objective 10, For a given input parameter failure to the steam generators level control system, determine the effect on steam generator level.

References:

AOI-16, Loss of Normal Feedwater, Rev. 30.

Question: 45 Given the following plant conditions:

- The plant is at 70% power.

- Main Steam header pressure transmitter 1-PT-1-33 fails full scale high.

Based on this failure, the steam generators (SGs) will be_________(1)_________, and the crew will take manual control of ________(2)__________ to mitigate the consequences of the transient.

(1)

(2)

a.

overfed main feedwater regulating valve (MFRV) and reduce flow

b.

overfed main feed pump speed control and reduce flow

c.

underfed main feedwater regulating valve (MFRV) and raise flow

d.

underfed main feed pump speed control and raise flow DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the event will result in the SGs being overfed, but the MFRVs will not be the cause of the increase in feedwater flow.

b.

CORRECT. The failure of the steam header pressure transmitter high will cause the main feed pump speed control circuit to raise speed to compensate for an apparent drop in differential pressure.

c.

Incorrect. Plausible, since the candidate may confuse the impact of the failure on the MFRVs.

d.

Incorrect. Plausible, since the candidate may confuse the impact of the failure on feed pump speed control.

K/A Applicability:

The procedure aspect of the K/A is implied in the question, since the listed actions are procedurally directed.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 48 of 77 Question Number: 46 K/A: 061 K5.03 Knowledge of the operational implications of the following concepts as they apply to the AFW:

Pump head effects when control valve is shut.

Tier:

2 RO Imp:

2.6 RO Exam:

46 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

46 Source:

New Applicable 10CFR55 Section: (CFR: 41.5 / 45.7)

Learning Objective:

3-OT-SYS003B, Objective 23, Using plant drawings, determine the effect of a loss of instrument air/control power on the following valves/components:

a.MDAFWP regulating valve (main and bypass) b. TDAFWP regulating valve c.

AFW pumps

References:

AOI-10, Loss of Control Air, Rev. 38.

Question: 46 Given the following plant conditions:

- A plant startup is in progress.

- 1A and 1B Auxiliary feedwater (AFW) pumps are supplying each of the steam generators at 100 gpm.

- The air supply to 1-PCV-3-122, 1A AFW Pump Backpressure Control Valve, fails.

Comparing the response of the 1A AFW Pump to the 1B AFW Pump, which one of the following describes the effect of the air supply failure on motor amps and discharge pressure?

Motor Amps Pump Discharge Pressure

a.

1A Pump greater than 1B Pump 1A Pump greater than 1B Pump

b.

1A Pump greater than 1B Pump 1B Pump greater than 1A Pump

c.

1B Pump greater than 1A Pump 1A Pump greater than 1B Pump

d.

1B Pump greater than 1A Pump 1B Pump greater than 1A Pump DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the pump discharge pressure response is correct, but the motor amp response is incorrect.

b.

Incorrect. Plausible, since the candidate must evaluate parameters.

c.

CORRECT. The loss of air to 1-PCV-3-122 will cause the valve to close. This will result in 1A Pump discharge pressure rising. 1A Pump motor amps will lower.

d.

Incorrect. Plausible, since the motor amp response is correct, but the pump discharge pressure response is incorrect.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 49 of 77 Question Number: 47 K/A: 062 G2.4.11 A.C. Electrical Distribution Knowledge of abnormal condition procedures.

Tier:

2 RO Imp:

4.0 RO Exam:

47 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

47 Source:

New Applicable 10CFR55 Section: (CFR: 41.10 / 43.5 / 45.13)

Learning Objective:

3-OT-AOI4300, Rev. 1, Objective 4, Demonstrate ability/knowledge of AOI, by: a.

Recognizing Entry conditions b. Responding to Actions c. Responding to Contingencies (RNO) d. Responding to Notes/Cautions

References:

AOI-43.01 Loss of Unit 1 Train A Shutdown Boards, Rev. 6 Question: 47 Given the following plant conditions:

- Unit 1 is operating at 50% power, during a power escalation, with 1A CCP in service.

- 1A-A 6.9KV Shutdown Board has tripped on differential relay operation.

- The Control Building AUO reports extensive damage to the 1A-A 6.9 KV Shutdown Board bus bars.

Under these conditions, RCP seal cooling is ______(1)___________ and the crew will _____(2)______.

(1)

(2)

a.

Unavailable.

Continue performance of AOI-43.01, "Loss of Unit 1 Train A", then manually start 1B-B CCS pump.

b.

Available, via thermal barrier heat exchangers.

Continue performance of AOI-43.01, "Loss of Unit 1 Train A Shutdown Boards", then reestablish charging and letdown using.

c.

Unavailable.

Isolate letdown, then reestablish charging and letdown using AOI-43.01 Attachment 1.

d.

Available, via seal injection flow Continue performance of AOI-43.01, "Loss of Unit 1 Train A Shutdown Boards", then reestablish CCS flow to the thermal barrier heat exchangers using SOI-70.01.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the loss of the 1A-A 6.9KV Shutdown Board will cause a loss of the 1A CCS pump, which will in turn cause a low pressure start signal and an automatic start of the 1B-B CCS pump.

b.

CORRECT. The loss of the 1A-A 6.9KV Shutdown Board will cause a loss of the 1A CCS pump, which will in turn cause a low pressure start signal and an automatic start of the 1B-B CCS pump. Seal flow will be supplied from the RCS and will be cooled by the CCS flowing through the thermal barrier heat exchanger.

c.

Incorrect. Plausible, since the loss of the 1A-A 6.9KV Shutdown Board will cause a loss of the 1A CCP.

Loss of charging flow will require letdown isolation and performance of AOI-43.01 Attachment 1 to restore charging and letdown. However, the 1B-B CCS pump will be running, providing thermal barrier cooling.

d.

Incorrect. Plausible, since the loss of the 1A-A 6.9KV Shutdown Board will cause a loss of the 1A CCP.

Loss of charging flow will require letdown isolation and performance of AOI-43.01 Attachment 1 to restore charging and letdown. However, the 1B-B CCS pump will be running, providing thermal barrier cooling.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 50 of 77 Question Number: 48 K/A: 063 A1.01 Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including:

Battery capacity as it is affected by discharge rate Tier:

2 RO Imp:

2.5 RO Exam:

48 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

48 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.5 / 45.5)

Learning Objective:

3-OT-SYS057P, Rev.8, Objective 10, Given the condition/status of the 125V DC Vital system/component and the appropriate sections of Tech Specs, determine if operability requirements are met and what actions, if any, are required.

References:

WB-DC-30-27 AC and DC Control Power Systems - (Unit 1/Unit 2), Rev. 28.

Question: 48 In the event of a loss of all AC power, with an accident coincident with a single failure, the vital batteries have adequate capacity to provide for the required load continuously for ______(1)_______.

If DC load shedding is performed such that the loading on the battery is reduced, the battery will be available to supply the remaining loads for _______(2)_________.

(1)

(2)

a.

30 minutes 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

b.

30 minutes 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

c.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 8 hours

d.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4 hours DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since 30 minutes is the correct required load capacity.

b.

CORRECT. Per system design, this is the correct required load capacity, and extended capacity if load shedding is performed.

c.

Incorrect. Plausible, if candidate misapplies a common time in various plant design functions.

d.

Incorrect. Plausible, since 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is correct.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 51 of 77 Question Number: 49 K/A: 064 K6.07 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers Tier:

2 RO Imp:

2.7 RO Exam:

49 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

49 Source:

New Applicable 10CFR55 Section: (CFR: 41.7 / 45.7)

Learning Objective:

3-OT-SYS082D, Rev. 2, Objective 3, Identify the air pressure in the air receiver at which the DG start air pressure Lo alarm annunciates.

References:

Tech Spec 3.8.3, Diesel Fuel Oil, Lube Oil, and Starting Air.

Question: 49 Given the following plant conditions:

- The unit is at 100% power.

- The Control Room Operator is responding to 195-C, DG START AIR PRESS LO annunciator.

- The Outside AUO reports from the 1A Diesel Generator that both of the local low start air pressure alarms are lit, due to an air leak on an individual start air receiver.

Based on these conditions, the 1A Diesel Generator is

a.

INOPERABLE until the local LOW START AIR PRESSURE TANK alarm (200 psig) is cleared.

b.

INOPERABLE until the local LOW START AIR PRESSURE ENGINE INLET alarm (160 psig) is cleared.

c.

OPERABLE if Start Air Header Pressure is 165 psig.

d.

OPERABLE if Start Air Receiver pressure is 185 psig.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the reset point for the alarm is 200 psig. At this point the DG is operable.

b.

Incorrect. Plausible, since the reset point of the starting air headers is 160 psig. When this alarm is cleared, the DG is still INOPERABLE.

c.

Incorrect. Plausible, since the reset point of the starting air headers is 160 psig. When this alarm is cleared, the DG is still INOPERABLE, because start air header pressure is less than the specification (170 psig).

d.

CORRECT. The DG remains OPERABLE, as long as the locally read air receiver tank pressure is greater than 170 psig and recovered to greater than 200 psig in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 52 of 77 Question Number: 50 K/A: 064 K6.08 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system:

Fuel oil storage tanks Tier:

2 RO Imp:

3.2 RO Exam:

50 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

50 Source:

New Applicable 10CFR55 Section: (CFR: 41.7 / 45.7)

Learning Objective:

3-OT-SYS082E, Rev.1, Objective 6, Identify the alarms associated with the Diesel Generator Fuel Oil System.

References:

ARI-195-201, 1A-A Diesel Generator, Rev.12; System LP 082E.

Question: 50 If the DG Day Tank level is unexpectedly decreasing, which one of the following is the LOWEST of the listed quantities of fuel oil remaining in the Day Tank that will maintain the DG OPERABLE?

a.

275 gallons.

b.

250 gallons.

c.

200 gallons.

d.

175 gallons.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since this number is close to the mid range for acceptable lube oil inventory for a DG. LCO 3.8.3 requires greater than 267 and less than 287 gallons of lube oil inventory to declare a DG OPERABLE.

b.

CORRECT. The minimum amount of fuel oil required in each skid mounted day tank is 218.5 gallons.

c.

Incorrect. Plausible, since this level corresponds to the amount of fuel oil remaining in the day tank when the low level alarm is received (~ 3/8 full).

d.

Incorrect. Plausible, since this value (175) is a value associated with certain DG parameters (175°F is the minimum lube oil temperature for normal operating range; 175 psig is a value associated with air receiver pressure). Candidate confuses the values and makes the incorrect selection here.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 53 of 77 Question Number: 51 K/A: 073 K3.01 Knowledge of the effect that a loss or malfunction of the PRM system will have on the following:

Radioactive effluent releases Tier:

2 RO Imp:

3.6 RO Exam:

51 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

51 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.7 / 45.6)

Learning Objective: 3-OT-SYS077B, Obj 10, Describe the general procedure to make a gaseous release.

References:

SOI-77.02, Waste Gas Disposal System, Rev. 34.

Question: 51 During a Waste Gas Decay Tank release, which ONE of the following conditions occurring during the release could result in a potential release outside of permitted limits, assuming NO operator actions?

a.

Emergency Gas Treatment System fan trips off.

b.

Waste Gas Monitor, 0-RM-90-118 fails low at 10 cps.

c.

Shield Building Vent Monitor, RM-90-400, loses power.

d.

Waste Gas Monitor, 0-RM-90-118, was not completely purged after the last release.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since the ABGTS fan, not the EGTS fan is used for dilution flow during a gas release. Additionally, the EGTS fan flow is an input to the isokinetic flow measurement for 1-RM-90-400 Shield Building Exhaust Radiation Monitor.

b.

CORRECT. Failure of 0-RM-90-118 to an arbitrarily low value renders the monitor incapable of detecting an out-of-limits high condition, and may result in exceeding release limits.

c.

Incorrect. The loss of power to the Shield Building Vent Monitor causes 0-FCV-77-119 to close, terminating the release.

d.

Incorrect. Plausible, since the radiation monitor is purged with nitrogen after a release. If the purge was incomplete, residual radioactivity would be present and would tend to terminate the subsequent release early (well before a limit is exceeded).

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 54 of 77 Question Number: 52 K/A: 076 A1.02 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.

Tier:

RO Imp:

RO Exam:

52 Cognitive Level:

High Group:

SRO Imp:

SRO Exam:

52 Source:

New Applicable 10CFR55 Section: (CFR: 41.5 / 45.5)

Learning Objective:

3-OT-SYS067A, Objective 8: State the ERCW System normal discharge path and given a failure of the path, discuss the alternate discharge paths.

References:

AOI-13, LOSS OF ESSENTIAL RAW COOLING WATER, Rev. 35.

Question: 52 Given the following plant conditions:

- The plant is operating at 100% power.

- 1A CCP is in service.

The following annunciators and indications are noted:

- 225-B, ERCW PUMP C-A DISCH PRESS LO annunciator is LIT.

EI-67-35A, ERCW C-A AMPS indicates 12 amps.

PI 37A, ERCW PUMP C-A DISCH PRESS indicates 0 psig.

Which ONE of the following describes the impact of the failure on the listed parameters?

(Assume no other operator action.)

1A CCP Oil CCS flow through the CVCS Temperature Letdown Heat Exchanger A.

Rises Rises B.

Rises Remains constant C.

Remains constant Rises D.

Remains constant Remains constant DISTRACTOR ANALYSIS

a.

CORRECT. The C-A ERCW pump supplies flow to the A header. When the pump shaft shears, flow to both the 1A and 2A ERCW headers drops. A reduction in flow to the 1A CCS heat exchanger will cause the A ESF header to heat up as well as the Reactor Building and the Miscellaneous Equipment Headers to heat up. The 1A CCP will show an increase in oil temperature. The Letdown Exchanger, supplied off the Miscellaneous Equipment Header would also heat up, causing the TCV to OPEN in order to maintain letdown temperature.

b.

Incorrect. Plausible, because oil temperature does rise, however, Letdown Heat Exchanger temperature also rises resulting in more CCS flow as the TCV repositions.

c.

Incorrect. Plausible, since the candidate must evaluate the effect of the C-A Pump shaft shearing. As stated in distractor analysis for a., CCP oil temperature remains rises and the Letdown Heat Exchanger temperature rises.

d.

Incorrect. Plausible, since candidate incorrectly recalls cooling water supply to this component.

K/A Applicability:

Candidate is presented with conditions involving an Emergency Raw Cooling Water pump with several operating parameters in an alarm condition. Though not directly "operating the SWS controls", alarms and indicators are at least associated with a form of control and monitoring, and the candidate must evaluate these operating parameters and predict the impacts of them on two parameters associated with a closed cooling water system: component cooling water.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 55 of 77 Question Number: 53 K/A: 078 A4.01 Ability to manually operate and/or monitor in the control room:

Pressure gauges Tier:

2 RO Imp:

RO Exam:

53 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

53 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.7 / 45.5 to 45.8)

Learning Objective:

3-OT-SYS032A, Obj. 16, List the events and their corresponding set points that take place on decreasing control air pressure.

References:

AOI-10, Loss of Control Air, Rev. 38 Question: 53 Operators are monitoring Control Air Pressure due to a rupture in the system. As air pressure lowers, the Auxiliary Air Compressors will start at __(1)__ psig, and if pressure continues to lower to ___(2)___ psig, air to the Reactor Building will automatically isolate.

(1)

(2)

a.

83 80

b.

83 70

c.

79.5 75

d.

79.5 70 DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since the Auxiliary Air Compressor starts at 83 psig, but the essential and non-essential air systems do not isolate at 80 psig.

b.

CORRECT. Auxiliary Air Compressor starts at `83 psig, and the essential and non-essential air systems isolate at 70 psig.

c.

Incorrect. Plausible since the auxiliary air system isolates at 79.5 psig, but the essential and non-essential air systems do not isolate at 75 psig. 75 psig is plausible since this is the required pressure for reopening air to containment valve.

d.

Incorrect. Plausible since the auxiliary air system isolates at 79.5 psig, but the essential and non-essential air systems isolate at 70 psig.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 56 of 77 Question Number: 54 K/A: 078 K1.01 Instrument Air Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: Sensor air Tier:

2 RO Imp:

2.8 RO Exam:

54 Cognitive Level:

High Group:

1 SRO Imp:

n/a SRO Exam:

54 Source:

New Applicable 10CFR55 Section: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Learning Objective:

3-OT-SYS001B, Rev.6, Objective 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.

References:

1-47W611-1-2 Logic Diagram.

Question: 54 Given the following plant conditions:

- A reactor trip from 35% power has just occurred.

- The steam dumps are armed.

- The Turbine Building AUO reports that an air leak has developed on the supply to FM-1-103, Current to Pneumatic (I/P) Controller.

(Complete the following sentence.)

As air pressure drops, the output air signal sent to the flow controllers (FC-1-103A thru D) will go to zero,...

a.

resulting in the dump valves all opening.

b. resulting in the dump valves all closing.

c. but 6 dump valves will remain open until the HI bistable reaches its reset point.

d. but all 12 dump valves will remain open until the HI-HI bistable reaches its reset point.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the operator must understand how the output of the I/P converter is applied to the steam dump valves.

b.

CORRECT. The output of the I/P converter ranges from 3-15 psig (which corresponds to a 10-50 ma input signal). This signal is sent to the 4 flow controllers. A 3-6 psig signal positions the cooldown dumps from full closed to full open. A 6-9 psig signal is sent to FM-103B, and controls the second group of 3 dump valves. A 9-12 psig signal is sent to FM-103C, and controls the third group of 3 dump valves.

Finally, a 12-15 psig signal is sent to FM-103D and controls the last group of 3 dump valves. A greater air signal causes MORE dump valves to open. For the leak described, air pressure would be decreasing and would result in dump valves closing.

c.

Incorrect. A trip from 35% power will result in a Tavg-to-T no-load mismatch of approximately 12°F, which is not sufficient to pick up the HI bistable.

d.

Incorrect. A trip from 35% power will result in a Tavg-to-T no-load mismatch of approximately 12°F, which is not sufficient to pick up the HI-HI bistable.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 57 of 77 Question Number: 55 K/A: 103 K1.01 Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems: CCS Tier:

2 RO Imp:

3.6 RO Exam:

55 Cognitive Level:

Low Group:

1 SRO Imp:

n/a SRO Exam:

55 Source:

New Applicable 10CFR55 Section: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Learning Objective:

3-OT-SYS030C, Obj 6, Describe the Lower Compartment Air Cooling system including: a. General description of the fan/coil units b. Suction/discharge flow paths c. Automatic starts of the fans d. Fan trips e. Number of fans required for normal operation f. Points of control

References:

N3-30RB-4002 Reactor Building Ventilation System Description Question: 55 Given the following plant conditions:

- Unit 1 is at 100% power.

- 1 A-A, B-B, and C-A Lower Compartment Coolers are operating in A-AUTO.

- 1 D-B Lower Compartment Cooler is shutdown and the handswitch is in A-AUTO.

Which one of the following will occur if the temperature control valve (TCV) for Lower Compartment Cooler 1A-A fails to the closed position?

a.

Containment temperature rises, causing TCVs for the running coolers to open more.

b.

Lower Compartment Cooler 1 D-B starts due to the rise in lower containment temperature.

c.

Lower Compartment Cooler 1 D-B starts due to the low ERCW flow through Lower Compartment Cooler 1 A-A.

d.

Lower Compartment Cooler 1A-A trips due to low ERCW flow and auto starts Lower Compartment Cooler 1 D-B.

DISTRACTOR ANALYSIS

a.

CORRECT. No auto initiation of the fan in A-P AUTO would occur and temperature would increase causing TCVs to open to compensate.

b.

Incorrect. There is no auto start function for increase in temp.

c.

Incorrect. There is no auto start function for low cooling water flow.

d.

Incorrect. Loss of cooling water to 1A-A fan would not trip the fan nor start the fan in A-P Auto.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 58 of 77 Question Number: 56 K/A: 001 K6.03 Knowledge of the effect of a loss or malfunction on the following CRDS components:

Reactor trip breakers, including controls.

Tier:

2 RO Imp:

3.7 RO Exam:

56 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

56 Source:

New Applicable 10CFR55 Section: (CFR: 41.7/45.7)

Learning Objective:

3-OT-SYS085A, Objective 24, Explain how a normal reactor trip occurs and how to perform an emergency reactor trip from outside the main control room.

References:

WOG Background Document, E-0, Rev 2.

Question: 56 An automatic reactor trip and Safety Injection is generated, but the Train A reactor trip breaker fails to open.

Which one of the following describes the effect of this failure?

a.

Feedwater isolation will NOT occur when RCS temperature decreases to the low Tavg setpoint, resulting in excessive RCS cooldown after the trip.

b.

The condenser steam dumps will NOT receive an arming signal from the associated P-4 contact, resulting in lifting of the SG power operated relief valves (PORVs).

c.

Prevents manual block of automatic safety injection reinitiation, requiring local actions to block reinitiation.

d.

Prevents the automatic trip of the main turbine, requiring a manual turbine trip from the main control room.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the Train A functions are affected. The feedwater isolation signal from Train B P-4 causes the main feed pumps, Standby MFP, condensate booster pumps and the main turbine trip signals to occur. The Train A feedwater and bypass isolation valves will not close, requiring local manual action to close them. AFW pumps will start. An excessive cooldown should not occur as a result of these conditions.

b.

Incorrect. Plausible since the Train A reactor trip breaker does control the reactor trip arming signal.

Although it is true that the Train A P-4 contact failing to reposition properly, will prevent a reactor trip arming signal, the steam dumps will arm from the C-7 function. The steam dumps would receive a positioning signal from the reactor trip controller and dump steam to the condenser. SG PORVs should not operate under these circumstances.

c.

CORRECT. The A train of SI will be affected by the failure of the P-4 contact associated with the Train A reactor trip breaker. Local actions WILL be required to compensate for the failure and allow blocking of the AUTO SI.

d.

Incorrect. Plausible, since the P-4 contacts are used by the turbine trip circuit to detect and respond to a reactor trip. With the Train A P-4 contact in the incorrect position, the candidate may believe that a manual trip is necessary. However, only one P-4 contact in the correct position results in a turbine trip from a reactor trip.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 59 of 77 Question Number: 57 K/A: 002 K3.03 Knowledge of the effect that a loss or malfunction of the RCS will have on the following:

Containment Tier:

2 RO Imp:

4.2 RO Exam:

57 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

57 Source:

New Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective:

3-OT-AOI0600, Obj. 2, Identify 10 Alarms and/or Indications of a small RCS leak.

References:

AOI-6, Small Reactor Coolant System Leak, Rev. 32; ARI-88-94, Reactor Coolant System, Rev 19; ARI-0-L-2D Liquid Waste Panel, Rev. 7.

Question: 57 With the plant in Mode 1, which one of the following conditions is expected to cause a small leak of Reactor Coolant into containment? (Consider each condition separately.)

a.

The RCDT level controller malfunctions, resulting in 0-L-2D, Window 1, REACTOR COOLANT DRAIN TANK HI-HI LEVEL LO-LO-LEVEL UNIT 1 alarm, and level on 1-LI-77-1 indicating 78%.

b.

1-PCV-77-158, RCDT N2 PRESS CONTROL malfunctions, resulting in Reactor Coolant Drain Tank pressure at 9 psig.

c.

Due to leakage past the reactor head inner seal ring, 88-A, RX VESSEL FLNG LEAKOFF TEMP HI alarm annunciates.

d.

Due to valve leakage, 88-E, RX HEAD VENT TEMP HI alarm annunciates.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the candidate may believe that the RCDT level at 78% causes the tank to overflow into containment.

b.

CORRECT. SOI-77.01 and SOI-68.01 both contain a precaution that states Allowing RCDT pressure to exceed 6 psig may allow RCP seals to divert to the Reactor Building floor. The HI pressure alarm setpoint is 10 psig, so at 9 psig the RCP seals would be diverting to the Reactor Building floor.

c.

Incorrect. Plausible, since leakage is indicated, but the leakage is being directed to a closed volume, specifically the RCDT and would not result in a release to the Reactor Building floor.

d.

Incorrect. Plausible, since leakage is indicated, but the leakage is being directed to a closed volume, specifically the PRT and would not result in a release to the Reactor Building floor.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 60 of 77 Question Number: 58 K/A: 016 K5.01 Knowledge of the operational implication of the following concepts as they apply to the NNIS:

Separation of control and protection circuits Tier:

2 RO Imp:

2.7 RO Exam:

58 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

58 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.5 / 45.7)

Learning Objective:

3-OT-SYS068C, Obj 8, Describe the operation of the master pressure controller.

References:

N3-69-4001, Reactor Coolant System, Rev. 25, Page 50.

Question: 58 A short circuit occurs internally on the Master Pressurizer Pressure Controller (1-PIC-68-340).

The short circuit will

a.

feed back into the protection circuit, causing the associated channel to trip.

b.

NOT feed back into the protection circuit, due to the use of isolation amplifiers.

c.

feed back into the protection circuit, preventing the associated channel from tripping.

d.

NOT feed back into the protection circuit, since completely separate pressure transmitters are used.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible if the candidate does not know that an isolation amplifier is installed in the circuit to prevent this from occurring.

b.

CORRECT. The circuit is provided with isolation amplifiers which ensure that the fault will not feed back to the protection circuit.

c.

Incorrect. Plausible if the candidate does not know that an isolation amplifier is installed in the circuit to prevent this from occurring.

d.

Incorrect. Plausible if the candidate believes that separate transmitters are used for protection and control, in a manner similar to the separate instrumentation used in the Auxiliary Control Room.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 61 of 77 Question Number: 59 K/A: 017 K1.01 Knowledge of the physical connections and/or cause/effect relationships between the ITM system and the following systems:

Plant computer Tier:

2 RO Imp:

3.2 RO Exam:

59 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

59 Source:

New Applicable 10CFR55 Section: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Learning Objective:

3-OT-SYS068F, Rev. 4, Objective 6, Use the RVLIS indications on the ICCM and the FR-C status tree to diagnose core cooling conditions and direct implementation of the appropriate Function Restoration Instruction.

References:

3-OT-SYS068F, Reactor Vessel Level Instrumentation System, Rev. 4; Inadequate Core Cooling Monitor Remote Display Manual, Rev. 0.

Question:

59 If a calculated analog input point quality code of BAD is assigned to INCORE TCs HI QUAD, then a reverse video ___(1)____ will be displayed next to the new value of ____(2)____ on the RVLIS-ICCM Plasma Display.

(1)

(2)

a.

B BBB

b.

D BBB

c.

B XXX

d.

D XXX DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since a reverse video B will be displayed for a BAD point, but the displayed value will be indicated with Xs, not Bs.

b.

Incorrect. Plausible since a SUSPECT data point would be indicated by a value comprised of Ds.

c.

CORRECT. The reverse video would indicate a B for BAD and apply a value comprised of Xs.

d.

Incorrect. Plausible since a SUSPECT data point would be indicated by a value comprised of Ds.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 62 of 77 Question Number: 60 K/A: 027 K2.01 Knowledge of bus power supplies to the following:

Fans Tier:

2 RO Imp:

3.1 RO Exam:

60 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

60 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective:

3-OT-SYS065A, Rev. 9, Objective 9, Describe the EGTS fans power supplies.

References:

N3-65-4001, Emergency Gas Treatment System. Rev. 9, Table 1 EGTS Power Supplies.

Question: 60 Given the following conditions:

- Unit 1 is in Mode 3 when a Safety Injection occurs.

- Following the Safety Injection, both the 1A-A Shutdown Board and the 2B-B 6.9 KV Shutdown Boards trip on differential.

- All other systems and components respond per design.

Which ONE of the following describes the status of the EGTS system?

a. Both trains EGTS fans are running.

b. Only EGTS train A fan is running.

c.

Only EGTS train B fan is running.

d. Neither EGTS fans are running.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since it is possible for both trains of EGTS to start on the Phase A isolation signal resulting from a Safety Injection signal.

b.

Incorrect. Plausible, since the Train A equipment could be powered from Unit 2, like the ABGTS fans are.

c.

CORRECT. EGTS Train B is supplied from Unit 1 1B Shutdown Power.

d.

Incorrect. Plausible, if the candidate assumes that the EGTS equipment is COMMON and could have power for independent trains from each unit (i.e., Train A Unit 1 A Train, Train B Unit 2 B Train).

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 63 of 77 Question Number: 61 K/A: 033 K4.01 Knowledge of design feature(s) and/or interlock(s) which provide for the following:

Maintenance of spent fuel level Tier:

2 RO Imp:

2.9 RO Exam:

61 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

61 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.7)

Learning Objective:

3-OT-SYS078A, Obj. 6, State the design features of the Spent Fuel System which prevent dewatering.

References:

N3-78-4001, Spent Fuel Pool Cooling and Cleaning System, Rev. 15, Page 23 of 68.

Question: 61 Which ONE of the following describes a feature of the Spent Fuel Pool Cooling System designed to limit the loss of water from the pool in the event of a SFP Cooling pump discharge header break?

Spent Fuel Pool pumps...

a.

take suction from and discharge to the skimmer loop.

b.

take suction approximately 4 feet below the normal level of the pool.

c.

will trip and the associated discharge valve(s) close when the low level alarm is received in the pool.

d.

are equipped with a vacuum breaker located in the pump suction which, if uncovered, will result in a loss of suction to the pumps.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since there is a pump (the skimmer pump) connected to the skimmer loop.

b.

CORRECT. The SFPCCS is designed and located to minimize the probability and effects of pipe ruptures. The suction for the SFPCCS is four feet below the normal SFP water level elevation to prevent draining of the SFP in the event of a suction line break.

c.

Incorrect. Plausible, since low level conditions in the Spent Fuel Pool are annunciated. Some plant systems are designed to trip pumps on low level conditions.

d.

Incorrect. Plausible, since there is a one-inch diameter hole in the discharge piping located in the SFP two feet below the normal water level to prevent draining the SFP in the event of a discharge line break.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 64 of 77 Question Number: 62 K/A: 034 A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fuel Handling System controls including:

Water level in the refueling canal Tier:

2 RO Imp:

2.9 RO Exam:

62 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

62 Source:

WBN Bank Applicable 10CFR55 Section:

(CFR: 41.5 / 45.5)

Learning Objective:

3-OT-SYS079A, Rev. 6, Objective 18, Describe the possible consequences of a reactor cavity seal failure during refueling.

References:

N3-79-4001, Fuel Handling and Storage System, Rev. 16; AOI-29, Dropped Or Damaged Fuel Or Refueling Cavity Seal Failure, Rev 20.

Question: 62 Given the following plant conditions:

- Core load is in progress.

- A failure of the Reactor Cavity Seal occurred.

- Reactor cavity level is at el. 748' and dropping slowly.

Which ONE of the following actions is required per AOI-29, "Dropped or Damaged Fuel or Refueling Cavity Seal Failure"?

a.

Start one SI pump in the cold leg injection flowpath for cavity makeup.

b.

Align CCP suction to RWST and discharge to the RCS through normal charging.

c.

Align RHR suction to the RWST and discharge to the RCS through the hot legs.

d.

Align Refueling Water Purification pumps suction to RWST and discharge directly to refueling cavity.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the candidate may believe this is an acceptable make-up flowpath for loss of inventory events since it is viable in other plant instructions.

b.

CORRECT. CCP suction is aligned to the RWST with discharge to the normal charging flow path.

c.

Incorrect. Plausible if candidate mistakes make-up through the hot legs with the cold legs since aligning RHR to the RWST through the cold legs is a correct flow path per AOI-29.

d.

Incorrect. Plausible, since the candidate may believe this is an acceptable make-up flow path since the Refueling Water Purification pumps may be used to dewater the cavity and fill the transfer canal.

K/A Applicability:

Candidate is presented with conditions involving a loss of water level in the reactor cavity due to a failure of the cavity seal. At first glance, this may not appear to match the K/A, however, during this evolution, the reactor cavity and the refueling canal are connected, so that water level changes in the reactor cavity also result in water level changes in the refueling canal. Further, "operating the Fuel Handling System controls" aspect is met because the correct answer involves controlling one aspect of fuel handling; i.e., flow through the core.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 65 of 77 Question Number: 63 K/A: 045 g2.2.37 Ability to determine operability and/or availability of safety related equipment.

Tier:

2 RO Imp:

3.6 RO Exam:

63 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

63 Source:

New Applicable 10CFR55 Section:

(CFR: 41.7 / 43.5 / 45.12)

Learning Objective:

3-OT-SYS047A, Obj. 6, Identify turbine trip inputs to the SSPS.

References:

1-47W611-99-1 Reactor Protection System.

Question: 63 Given the following plant conditions:

- The unit is at 62% power during a load escalation.

PS-47-74, Turbine Auto Stop Oil Pressure transmitter, fails low.

Which one of the following is a complete description of the alarm(s) expected to be LIT as a result of this failure and the operability of the Turbine Trip function for the Reactor Protection System?

Alarms OPERABILITY of Turbine Trip Function

a.

121-D TURB AUTO STOP OIL PRESS LO 71-A AUTO STOP OIL PRESS LO 76-B TURBINE TRIP OPERABLE

b.

121-D TURB AUTO STOP OIL PRESS LO OPERABLE

c.

121-D TURB AUTO STOP OIL PRESS LO 71-A AUTO STOP OIL PRESS LO 76-B TURBINE TRIP INOPERABLE

d.

121-D TURB AUTO STOP OIL PRESS LO INOPERABLE DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since the low auto stop oil pressure causes 121-D TURB AUTO STOP OIL PRESS LO to alarm. Since the trip is a 2/3 logic, no trip would occur. The turbine trip function is OPERABLE.

b.

CORRECT. 121-D TURB AUTO STOP OIL PRESS LO will be lit since the alarm is a 1/3 logic. Since the trip is a 2/3 logic, no trip occurs. The turbine trip function is OPERABLE.

c.

Incorrect. Plausible since the low auto stop oil pressure sensed would cause 121-D TURB AUTO STOP OIL PRESS LO to alarm. Since the trip is a 2/3 logic, no trip would occur. The turbine trip function is OPERABLE.

d.

Incorrect. Plausible, since 121-D TURB AUTO STOP OIL PRESS LO alarm is a 1/3 logic. Since the trip is a 2/3 logic, no trip would occur. The turbine trip function is OPERABLE.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 66 of 77 Question Number: 64 K/A: 068 A3.02 Ability to monitor automatic operation of the Liquid Radwaste System including: Automatic isolation.

Tier:

2 RO Imp:

3.6 RO Exam:

64 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

64 Source:

New Applicable 10CFR55 Section: (CFR: 41.7 / 45.5)

Learning Objective:

3-OT-SYS077A, Obj 19, Discuss how processed water is released. 3-OT-SYS090A, Obj 7, 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.

References:

ARI-159-165, Rev 34, Page 21 of 48; ARI-180-187, Rev. 30, Page 9 of 47; Page 11 of 47 Question: 64 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- A planned release of the Monitor Tank is in progress.

- Power is lost to the rate meter associated with 0-RM-90-122A Waste Disposal System Release Line Radiation Monitor.

Which one of the following describes the effect of the failure on the release in progress, and which of the main control room alarms listed below are expected?

Main Control Room Alarms 1 - 161-F 0-L-2 RAD WASTE PNL TROUBLE 2 - 181-A WDS RELEASE LINE 0-RM-122 LIQ RAD HI 3 - 181-C WDS RELEASE LINE 0-RM-122 INSTR MALF Effect on Release Expected Alarms

a.

Must be MANUALLY terminated.

1 and 2 ONLY.

b.

MUST be MANUALLY terminated.

1, 2, and 3.

c.

Will be AUTOMATICALLY terminated.

1 and 2 ONLY.

d.

Will be AUTOMATICALLY terminated.

1, 2, and 3.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the 181-C WDS RELEASE LINE 0-RM-122 INSTR MALF alarms on the loss of power to the rate meter. 161-F 0-L-2 RAD WASTE PNL TROUBLE alarms. The release will be terminated automatically, NOT manually.

b.

Incorrect. Plausible, since the 181-C WDS RELEASE LINE 0-RM-122 INSTR MALF alarms on the loss of power to the rate meter. 161-F 0-L-2 RAD WASTE PNL TROUBLE alarms. The release will be terminated automatically, NOT manually.

c.

Incorrect. Plausible, since the 181-A WDS RELEASE LINE 0-RM-122 LIQ RAD HI alarms on the loss of power to the rate meter. Release is terminated automatically.

d.

CORRECT. Both 181-A WDS RELEASE LINE 0-RM-122 LIQ RAD HI and 181-A WDS RELEASE LINE 0-RM-122 INSTR MALF annunciators alarm when power is lost to the rate meter, resulting in the termination of the release.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 67 of 77 Question Number: 65 K/A: 072 A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Erratic or failed power supply Tier:

2 RO Imp:

2.7 RO Exam:

65 Cognitive Level:

High Group:

2 SRO Imp:

n/a SRO Exam:

65 Source:

New Applicable 10CFR55 Section: (CFR: 41.5 / 43.5 / 43.3 / 45.13)

Learning Objective:

3-OT-SYS090A, Obj. 9, Explain actions to take if an Area Rad Monitor alarms in a work area.

References:

ARI 174-E, Rev.44; AOI-31, Abnormal Release of Radioactive Material, Rev. 22.

Question: 65 Given the following plant conditions:

- The unit is operating at 100% power.

- 174-B, 1-RR-90-1 AREA RAD HI, is LIT.

- 174-E, 1-RR-90-1 AREA MONITORS INSTR MALF, is LIT.

- The GREEN indicating light on 1-RM-90-8, AFW Pumps is NOT LIT.

Which one of the following describes (1) the cause of the alarm, and (2) the actions required to be taken in response to the problem?

(1)

(2)

a.

High radiation in the AFW pump area.

Enter AOI-31, Abnormal Release of Radioactive Material, evacuate non-essential personnel from the affected area, evaluate the release using ICS EFF1 screen.

b.

High radiation in the AFW pump area.

Since the GREEN indicating light is NOT lit, check counts indicating on the readout module. If no counts are indicated, then perform a source check on 1-RM-90-8. Contact Radiation Protection to evaluate compensatory actions.

c.

Failure of 1-RM-90-8.

Enter AOI-31, Abnormal Release of Radioactive Material, evacuate non-essential personnel from the affected area, evaluate the release using ICS EFF1 screen.

d.

Failure of 1-RM-90-8.

Since the GREEN indicating light is NOT lit, check counts indicating on the readout module. If no counts are indicated, then perform a source check on 1-RM-90-8. Contact Radiation Protection to evaluate compensatory actions.

DISTRACTOR ANALYSIS

a.

Incorrect. It is plausible to believe that an actual high radiation in a plant area would cause alarm actuation, however, with the malfunction alarm also in, candidate mistakenly applies it.

b.

Incorrect. Plausible, since the second column is correct.

c.

Incorrect. Plausible, since AOI-31 is related to an actual alarm condition.

d.

CORRECT. ARI-174-B directs the operator to check 1-RR-90-1 and associated radiation monitors to determine the area that is affected. The ARI also directs the operator to determine if any loss of power has occurred. If a loss of power is indicated AND 174E annunciator is also LIT, the operator is directed to implement actions contained in ARI 174-E.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 68 of 77 Question Number: 66 K/A: G2.1.4 Operator responsibilities for shift staffing.

Tier:

3 RO Imp:

3.3 RO Exam:

66 Cognitive Level:

Low Group:

n/a SRO Imp:

n/a SRO Exam:

66 Source:

INPO Bank Applicable 10CFR55 Section: (CFR 43.10/43.5/45.13)

Learning Objective:

3-OT-SPP1000, Rev. 9, Objective 6, Describe shift staffing requirements.

References:

OPDP-1, Conduct of Operations, Rev.10 Question: 66 Given the following plant conditions:

- The plant is in Mode 4.

- A cooldown to Mode 5 has been initiated.

What is the difference in staffing requirements for the minimum on-duty shift complement per OPDP-1, "Conduct of Operations", when Mode 5 is achieved?

a.

Only ONE Unit Operator is required and the STA is NOT required.

b.

The Unit Supervisor and the STA are NOT required.

c.

Appendix R Fire Safe Shutdown personnel can be reduced to FOUR persons.

d.

Only TWO Non-Licensed Auxiliary (AUO) Unit Operators are required.

DISTRACTOR ANALYSIS

a.

CORRECT. Per the table in OPDP-1, during Mode 5 and 6 there is no requirement for the STA position to be manned. Additionally, the requirement for the number of Unit operators is reduced from 2 to 1.

b.

Incorrect. Plausible, since the STA is NOT required but the Unit Supervisor is required in Mode 5.

c.

Incorrect. Plausible, since the number of AUOs may be reduced to 3, but the fire brigade still requires at least 5 members.

d.

Incorrect. Plausible, since only 3 AUOs are required per OPDP-1.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 69 of 77 Question Number: 67 K/A: G2.1.42 New and spent fuel movement procedures.

Tier:

3 RO Imp:

RO Exam:

67 Cognitive Level:

Low Group:

n/a SRO Imp:

n/a SRO Exam:

67 Source:

New Applicable 10CFR55 Section:

Learning Objective:

3-OT-SYS079A, Rev. 6, Objective 4, Identify the maximum quantity of fuel that shall be out of approved storage locations during fuel handling operations.

References:

FHI-7 Fuel Handling and Movement, Rev. 31.

Question: 67 Per FHI-7, "Fuel Handling and Movement", what is the maximum number of IRRADIATED fuel assemblies allowed outside of approved storage?

a.

1

b.

3

c.

4

d.

5 DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since one fuel assembly is allowed within the spent fuel storage pool boundary (excluding the inspection, reconstitution, or cleaning locations with appropriate evaluation for each configuration that must be performed prior to implementation). The spent fuel storage pool boundary includes the cask loading area, fuel transfer canal (excluding the transfer cart), and spent fuel pool.

b.

Incorrect. Plausible, since three fuel assemblies are allowed within the refueling canal. The refueling canal includes the fuel transfer tube boundary (including the transfer cart) and the rod cluster control changing fixture. This allows for two fuel assemblies to be in the rod cluster control changing fixture while the third fuel assembly is being transferred through the fuel transfer tube, is in the Upender, or is in transit to or from the reactor cavity.

c.

Incorrect. Plausible, since the combination of assemblies allowed in the spent fuel storage pool; boundary and the refueling canal totals 4 assemblies. This is credible if the candidate does not recall the assembly allowed in the reactor vessel.

d.

CORRECT. One allowed in the spent fuel storage pool + one in the reactor vessel + 3 in the refueling canal = 5 total.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 70 of 77 Question Number: 68 K/A: G2.2.20 Knowledge of the process for managing troubleshooting activities.

Tier:

3 RO Imp:

2.6 RO Exam:

68 Cognitive Level:

Low Group:

n/a SRO Imp:

n/a SRO Exam:

68 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.10 / 43.5 / 45.13)

Learning Objective:

3-OT-OPDP-7, Obj. 4, Describe the procedure for replacing blown fuses in safety related circuits.

References:

OPDP-7, Fuse Control, Rev.3.

Question: 68 Given the following plant conditions:

- Unit 1 is at 100% power.

- The CRO notices that SG Blowdown Valve 1-FCV-1-15 has no indicating lights.

- The Control Building AUO reports that a fuse has blown in the valve's control circuit.

Which one of the following describes how troubleshooting for this condition is performed, including attempts at replacing the fuse?

a.

One attempt may be made to replace the fuse. If the fuse blows a second time, a Work Order must be written for the purpose of initiating troubleshooting on the circuit PRIOR to any further attempt at fuse replacement.

b.

One attempt may be made to replace the fuse. If the fuse blows a second time, Electrical Maintenance must be contacted to determine the exact cause of the problem. A Work Order is then written to perform needed repairs.

c.

Write a Work Order to have Electrical Maintenance perform troubleshooting PRIOR to any attempt at replacing the fuse. If they determine the cause was a faulty fuse, replace it. No additional Work Order is needed.

d.

Electrical Maintenance must be contacted to determine the exact cause of the blown fuse PRIOR to any attempt at replacing the fuse. If they determine the cause was a faulty fuse, replace it, and THEN document the condition by writing a Work Order.

DISTRACTOR ANALYSIS

a.

CORRECT. Whenever a fuse clears, (the first time a fuse clears it is not considered a condition adverse to quality) it may be replaced one time with the proper fuse as identified in accordance with this procedure. The operations personnel replacing the blown fuse shall prepare and submit a WO documenting the fuse replacement and any pertinent information associated with malfunction. For FLAS-5 fuses, include the circuit and the new fuse lot number for tracking. The proper verification (i.e.,

IV/2nd) shall be included on the WO. These WOs shall be sent to maintenance history. If the fuse should clear a second time in the same operational sequence a WO shall be initiated to troubleshoot the circuit before further fuse replacement. (In an emergency, SM may authorize further fuse replacement.) Forward all blown FLAS-5 fuses to NE-Electrical supervision for evaluation.

b.

Incorrect. Plausible, since one attempt may be made to replace fuse; however, the given followup action is incorrect, but plausible since Electrical Maintenance does perform troubleshooting activities related to the components.

c.

Incorrect. Plausible, since the action seems prudent, but this level of caution is not required since one attempt may be made to replace the fuse.

d.

Incorrect. Plausible, due to frequent use of related terms for the condition; however, as a whole these actions are incorrect.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 71 of 77 Question Number: 69 K/A: G2.2.21 Knowledge of pre-and post-maintenance operability requirements.

Tier:

3 RO Imp:

2.9 RO Exam:

69 Cognitive Level:

Low Group:

n/a SRO Imp:

n/a SRO Exam:

69 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.10 / 43.2)

Learning Objective:

3-OT-SPP0801, Obj. 2, Identify the applicability of SPP-8.1 to other tests or instructions.

References:

SPP-6.3, Pre-/Post-Maintenance Testing, Rev. 2; SPP-8.0, Testing Programs, Rev. 4; GOI-7, Generic Equipment Operating Guidelines, Rev. 33.

Question: 69 Given the following conditions:

- Unit 1 is in Mode 1.

- Maintenance has requested a clearance for 1B Safety Injection (SI) Pump.

- The clearance included taking the handswitch to STOP, PULL-TO-LOCK and racking down its supply breaker only.

- A visual inspection showed that NO work is required.

- NO disassembly work was performed on the equipment.

- The breaker has been racked up using the guidelines of GOI-7, "Generic Equipment Operating Guidelines".

Which one of the following describes the operability of the 1B Safety Injection (SI) Pump?

a.

1B Safety Injection (SI) Pump is OPERABLE now.

b.

1B Safety Injection (SI) Pump will be OPERABLE after the handswitch is in A-AUTO.

c.

1B Safety Injection (SI) Pump will be OPERABLE after the breaker is tested and the handswitch is in A-AUTO.

d.

1B Safety Injection (SI) Pump will be OPERABLE after performance of 1-SI-63-901-B, "Safety Injection Pump 1B-B Quarterly Performance Test".

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since the action returns the breaker to the cubicle, however, the pump handswitch is in an incorrect position for the current operating mode.

b.

CORRECT. Since no maintenance was performed after the visual inspection, the breaker being racked up AND the control handswitch being returned to A-AUTO satisfies the post maintenance testing for restoring the 1B SI pump to OPERABLE.

c.

Incorrect. Plausible, since the candidate may believe that a test of the pump breaker is required.

d.

Incorrect. Plausible, since the candidate may believe that the 1-SI-63-901-B Safety Injection Pump 1B-B Quarterly Performance Test is required to return the pump to OPERABLE status.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 72 of 77 Question Number: 70 K/A: G2.3.12 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.

Tier:

3 RO Imp:

3.2 RO Exam:

70 Cognitive Level:

Low Group:

n/a SRO Imp:

n/a SRO Exam:

70 Source:

New Applicable 10CFR55 Section: (CFR: 41.12 / 45.9 / 45.10)

Learning Objective:

3-OT-RAD0003, Obj. 8. Identify the responsibilities of the following concerning the ALARA program: a. Radiation Protection Manager/Radiation Safety Officer

b. TVA NPG Organization c. Employee

References:

RCI-100, Control of Radiological Work, Rev. 35.

Question: 70 In addition to a thermoluminescent dosimeter (TLD), which one of the following describes the acceptable monitoring method for personnel entering a Very High Radiation Area and the MINIMUM level of radiation protection coverage for the entry?

1)

Electronic Alarming Dosimeter that continuously integrates the radiation dose rates in the area and alarms when the dose rate or dose alarm setpoints are reached.

2)

Electronic Alarming Dosimeter that continuously transmits dose rate and cumulative dose to a remote receiver monitored by RP personnel.

3)

Continuous surveillance by an individual qualified in radiation protection procedures, via a closed circuit television.

4)

Continuously accompanied by, and in direct communications with, a qualified RP technician equipped with a radiation monitoring device that continuously displays the dose rate in the area.

a.

1 and 3.

b.

1 and 4.

c.

2 and 3.

d.

2 and 4.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible since this combination of equipment and RP coverage is applicable for entry into a Locked High Radiation Area, per the "Requirement for Entry" matrix in RCI-100 Control of Radiological Work.

b.

CORRECT. This combination is correct per the matrix in RCI-100 Control of Radiological Work.

c.

Incorrect. Plausible since this combination of equipment and RP coverage is applicable for entry into a Locked High Radiation Area, per the "Requirement for Entry" matrix in RCI-100 Control of Radiological Work.

d.

Incorrect. Plausible since this combination of equipment and RP coverage is applicable for entry into a Locked High Radiation Area, per the "Requirement for Entry" matrix in RCI-100 Control of Radiological Work.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 73 of 77 Question Number: 71 K/A: G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Tier:

3 RO Imp:

3.4 RO Exam:

71 Cognitive Level:

High Group:

n/a SRO Imp:

n/a SRO Exam:

71 Source:

Bank Applicable 10CFR55 Section: (CFR: 41.12 / 43.4 / 45.10)

Learning Objective:

3-OT-RAD0003, Obj. 7, Explain the ALARA concept.

References:

SOI-62.01, CVCS-Charging and Letdown, AOI-20, "Malfunction of Pressurizer Level Control System", Rev. 31; SOI-78.01, Spent Fuel Pool Cooling And Cleaning System, Rev.

56.

Question: 71 Which one of the following conditions or evolutions REQUIRES contacting Radiological Protection?

(Consider the effects of the described action only.)

a.

Reestablishing letdown flow using AOI-20, Attachment 1, "Alignment of Charging and Letdown".

b. 1A Safety Injection (SI) pump is started to perform a Technical Specification surveillance test.

c. Increasing spent fuel pool (SFP) CCS cooling flow during spent fuel pool fuel moves.

d. Shifting from 1A to 1B Charging Pump.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the candidate may believe that AOI-20, Attachment 1 requires contacting Rad Pro when changing letdown flow.

b.

Incorrect. Plausible, since shifting of CCPs does require Rad Pro support. The source of water to the safety injection pumps is the RWST, and during the injection phase there is no contaminated water recirculated from the containment sump.

c.

Incorrect. Plausible, since the candidate may believe that raising SFP cooling flow will change radiological conditions in the plant and therefore require contacting Rad Pro.

d.

CORRECT. Per SOI-62.01, Section 6.2, Swapping CCPs, NOTE prior to Step 1 states Radiological Protection should be notified when starting or changing CCP alignment for ALARA concerns and to revise Radiological Protection postings.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 74 of 77 Question Number: 72 K/A: G 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Tier:

1 RO Imp:

2.7 RO Exam:

72 Cognitive Level:

Low Group:

2 SRO Imp:

n/a SRO Exam:

72 Source:

WBN Bank Applicable 10CFR55 Section: (CFR 41.7 / 45.7)

Learning Objective:

3-OT-SYS070A, Obj 12, Identify the automatic actions that occur upon detection of CCS high radiation.

References:

1-47W611-70-1, -2 Logic Diagrams.

Question:

72 In response to a 188-A "CCS HX B OUTLET 2-RM-123 LIQ RAD HI" alarm, Unit 1 Component Cooling System (CCS) surge tank vent will ___(1)___ and Unit 2 CCS surge tank vent will ___(2)___.

(1)

(2)

a.

remain OPEN remain OPEN

b.

remain OPEN CLOSE

c.

CLOSE remain OPEN

d.

CLOSE CLOSE DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, since the candidate may believe that a liquid process monitor signal will not close the vent valves.

b.

Incorrect. Plausible, since the surge tanks are independent. A high radiation condition will cause both vent valves to close.

c.

Incorrect. Plausible, since the surge tanks are independent. A high radiation condition will cause both vent valves to close.

d.

CORRECT. Vent valve FCV-70-66 automatically closes if radiation is detected on any of the 3 detectors (0-RM-90-123, 1-RM-90-123, or 2-RM-90-123).

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 75 of 77 Question Number: 73 K/A: G2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Tier:

3 RO Imp:

3.8 RO Exam:

73 Cognitive Level:

High Group:

n/a SRO Imp:

n/a SRO Exam:

73 Source:

INPO Bank Applicable 10CFR55 Section: (CFR: 41.10 / 43.5 / 45.13)

Learning Objective:

3-OT-AOI1400, Obj. 4, Describe 5 ways that RHR Cooling can be lost.

References:

AOI-14, Loss of RHR Shutdown Cooling, Rev. 34; ARI 90-E, ARI-91E, ARI-113E.

Question: 73 Given the following initial conditions:

- Unit 1 reactor is shutdown.

- All RCS temperatures are approximately 160°F

- Transition to water solid operation is in progress.

- Pressure is 330 psig.

- Cold overpressure protection is armed.

- Both PORVs, 1-PCV-68-334 and 1-PCV-68-340, are closed.

- RHR Train A is in service.

While adjusting CVCS charging flow, RCS pressure increases to 600 psig and the following alarms are actuated:

- RCS Approaching COPS Setpoint.

- RHR Suction FCV-74-1,2,8,9 Open & Hi Press.

Which ONE of the following describes (1) the expected PORV response, and (2) the actions which must be taken to mitigate the event prior to entering AOI-14, "Loss of RHR Shutdown Cooling"?

(1)

(2)

a.

Only ONE PORV OPENS.

Adjust charging to restore RCS pressure.

b.

Only ONE PORV OPENS.

Stop A Train RHR Pump and ensure RHR suction valves are closed.

c.

BOTH PORVS OPEN.

Adjust charging to restore RCS pressure.

d.

BOTH PORVS OPEN.

Stop A Train RHR Pump and ensure RHR suction valves are closed.

DISTRACTOR ANALYSIS

a.

Incorrect. Plausible, because COMS does have a staggered lift sepoint design feature (approximately 20 psi setpoint difference between the PORVS. Second column is correct.

b.

Incorrect. Plausible, because COMS does have a staggered lift sepoint design feature (approximately 20 psi setpoint difference between the PORVS. Second column is plausible because this is an action prescribed by the ARI, but candidate fails to recall this is only if RHR is not in service. Also, this used to be an auto isolation feature of the RHR suction valves.

c.

CORRECT. Per COMS design, both PORVs lift for the given condition (when greater than 575 psig),

and the ARI prescribes charging flow adjustment.

d.

Incorrect. Plausible, since the first column is correct. Second column is plausible because this is an action prescribed by the ARI, but candidate fails to recall this is only if RHR is not in service. Also, this used to be an auto isolation feature of the RHR suction valves.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 76 of 77 Question Number: 74 K/A: G2.4.19 EOP layout, symbols and icons.

Tier:

3 RO Imp:

3.4 RO Exam:

74 Cognitive Level:

Low Group:

n/a SRO Imp:

n/a SRO Exam:

74 Source:

INPO Bank Applicable 10CFR55 Section: (41.10/45.13)

Learning Objective:

3-OT-TI1204, 1, Obj 13: Apply rules of usage which relate to performing steps of an EOP in a specified sequence to determine when steps may/must be performed.

References:

TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions.

Question: 74 How are Emergency Operating Instruction (EOI) sub-steps designated if they must be performed in the order in which they are listed?

a.

Substeps are designated by bullets.

b.

The entire step is surrounded by a solid line.

c.

The major step is designated with a double asterisk.

d.

Substeps are designated by alpha-numeric characters.

DISTRACTOR ANALYSIS A. Incorrect. Plausible, since bullets are used if substeps may be completed in any order.

B. Incorrect. Plausible, since Cautions and Notes surrounded by a line in normal operating procedures.

C. Incorrect. Plausible, since double asterisks are used to designate transition points in the EOPs.

D. CORRECT. Letters or numbers listed sequentially are used to identify procedure steps that MUST be performed in sequence.

RO Watts Bar May 2009 NRC Initial License Exam WRITTEN QUESTION DATA SHEET 77 of 77 Question Number: 75 K/A:G2.4.27 Knowledge of fire in the plant procedures.

Tier:

3 RO Imp:

3.4 RO Exam:

75 Cognitive Level:

High Group:

n/a SRO Imp:

n/a SRO Exam:

75 Source:

WBN Bank Applicable 10CFR55 Section: (CFR: 41.10 / 43.5 / 45.13)

Learning Objective:

3-OT-AOI3000, Obj 7, Identify the section of AOI-30.2 giving procedural guidance relative to each of the following: a. Location of component(s) within Auxiliary, Control, or Reactor buildings or Intake pumping station b. Control Air c.

Ventilation Systems with failed fire dampers

References:

AOI-30.2, Fire Safe Shutdown, Rev. 28 Question: 75 After controls are transferred per AOI-30.2, Fire Safe Shutdown, RCS temperature is controlled using

____(1)____ and RCS pressure is controlled using ____(2)_____.

(1)

(2)

a.

Condenser steam dumps PZR back-up heater group C.

b.

Condenser steam dumps PZR back-up heater groups A-A and B-B.

c.

Atmospheric steam dumps PZR back-up heater group C.

d.

Atmospheric steam dumps PZR back-up heater groups A-A and B-B.

DISTRACTOR ANALYSIS

a.

Incorrect. Examinee must know that the MSIVs are closed and condenser steam dumps are not available. Also should know that loss of offsite power is assumed for App. R fire, thus credit is only take for the A-A and B-B backup heaters.

b.

Incorrect. Examinee must know that the MSIVs are closed and condenser steam dumps are not available.

c.

Incorrect. Examinee must know that loss of offsite power is assumed for App. R fire, thus credit is only take for the A-A and B-B backup heaters.

d.

CORRECT. Atmospheric dumps are only available since the MSIVs are closed. Loss of offsite power is assumed for Appendix R fire, thus credit is only taken for the A-A and B-B backup heaters.