ML060860245

Ro/Sro Initial Exam

ML060860245
Person / Time
Site:	Cook
Issue date:	02/06/2006
From:	NRC/RGN-III
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Download: ML060860245 (132)

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DC Cook 2006 NRC Exam 1

1. 001 11 Rod Control was in AUTO with Unit 1 power at 79% when MPC-253, Turbine Impulse Pressure Channel 1, failed low. The crew determined that a Turbine Runback was NOT in progress and placed rod control in Manual.

Which ONE of the following actions are required to "Restore Equilibrium Conditions" in accordance with 01-OHP-4022-012-003, Continuous Control Bank Movement?

A. Initiate boration.

B. Reduce Turbine Load.

C. Insert Control Rods.

D. Initiate dilution.

Answer: B Per 01-OHP-4022-012-003, Continuous Control Bank Movement, a reduction in steam demand is the prefered method to restore Tave to Tref. The failure of MPC-253 would have caused the control Rods to Insert (Tref lowered).

A - Incorrect - Rods inserted and Tave would be low for the current power level.

C - Incorrect - Rods inserted and Tave would be low for the current power level.

D - Incorrect - The procedure directs a reduction of Turbine load or a withdrawal of the control rods.

REFERENCE:

01-OHP-4022-012-003, Continuous Control Bank Movement, Step 3 LESSON PLAN/OBJ: RO-C-AOP-7 / #5 KA - 000001 2.1.20 Continuous Rod Withdrawal Conduct of Operations Ability to execute procedure steps.

RO/SRO Value - (4.3 / 4.2) CFR - 41.10 / 43.5 / 45.12 Question #1 KA# - 000001 2.1.20 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 2

2. 002 1 During a power accension, with reactor power at 48%, control Bank C - Group 1 rod B-8 drops.

Prior to the drop it was at 230 steps. While restoring the rod, control rod urgent failure alarm occurs.

Which one of the following explains why the alarm actuated?

A. All other Bank C - Group 1 rod lift coil disconnect switches are open.

B. All Bank C - Group 2 rod lift coil disconnect switches are open.

C. The step counter of the pulse to analog (P/A) converter was not reset to 0.

D. Group C rod moving with group D rods withdrawn.

Answer: B Since the dropped rod is completely inserted, the lift coil disconnect switches for all operable rods within the affected bank are opened. An Urgent failure will occur when the misaligned rod begins to move. This is caused by the non-movement of the group without the misaligned rod.

A - Incorrect - While all other Bank C Group 1 rods lift coils deenergized, the Alarm is generated from the failure of Group 2 movement (System monitors current through the lift coils - Since Bank C group 1 rod B-8 still has current the alarm is from group 2)

C - Incorrect - While the P/A Converter is reset during rod recovery, failure to do so would not cause an urgent failure.

D - Incorrect - Group C is moved in the bank select mode. This would not cause an urgent failure alarm.

REFERENCE:

02-OHP-4024-210, Annunciator #210 Response: Flux Rod, Drop 26 Rod Control Urgent Failure, RO-C-AOP-6, Abnormal Operating Procedures - Day 6 pg. 56 LESSON PLAN/OBJ: RO-C-AOP-6/#AOP 6.20, RO-C-01200 / #4 KA - 000003 AA1.02 Dropped Control Rod Ability to operate and/or monitor the following as they apply to the Dropped Control Rod:

Controls and components necessary to recover rod RO/SRO Value - (3.6 / 3.4) CFR - 41.7 / 45.5 / 45.6 SOURCE: INPO # 27278 Ginna 1 - 4/27/2004 Original Quest. KA - 000003AK2.05 Question #2 KA# - 000003 AA1.02 Exam Level - RO Question Source - DIRECT-INPO - GINNA 2004-27278 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 3

3. 003 3 During power escalation of Unit 2, a single rod in Control Bank D is misaligned. Procedure OHP 4022.012.005, Dropped Or Misaligned Rod is being performed. A CAUTION prior to step

1. 2 requires that any required power reductions must be accomplished using boration rather than control rods.

The purpose of this requirement is to prevent:

A. Moving the control rods until a failure analysis can be initiated.

B. Another rod failure until the cause of the initial failure can be identified.

C. Exaggerating any existing flux tilts.

D. Exceeding Rod Insertion Limits during the power reduction.

Answer: C With a single misaligned rod, flux tilts could develop across the core. Further movement of the rods could cause a greater tilt and subsequent oscillations.

A - Incorrect - Failure analysis is not the reason for using Boron B - Incorrect - While this may be prudent it is not the reason for the caution.

D - Incorrect - Insertion Limits wouldn't be exceeded with 1 rod misaligned.

REFERENCE:

1(2)-OHP 4022.012.005 LESSON PLAN/OBJ: RO-C-AOP-6/#21; RQ-C-2514/#4 KA - 000005 AK1.02 Inoperable/Stuck Control Rod Knowledge of the operational implications of the following concepts as they apply to Inoperable/Stuck Control Rod:

Flux tilt RO/SRO Value - (3.1 / 3.9) CFR - 41.8 / 41.10 / 45.3 Original question Source: Bank 01AOPC621-1 Original KA APE 003 AA1.05 (4.1/4.1)

ORIGINATION DATE: 10/24/00 REVISION DATE: 10/16/2003 EXAM/QUIZZES: RO21AOP3; RQ2526C-R/S; RO22EOP4A; RO22EOP4B Question #3 KA# - 000005 AK1.02 Exam Level - RO Question Source - DIRECT - BANK 01AOPC621-1 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 4

4. 004 3 The following plant conditions exist:

- A Reactor trip from 100% power has occurred on Unit 2. OHP-4023-E-0, Reactor Trip Or Safety Injection, is being implemented.

- Containment pressure is 0.6 psig and stable.

- SG NR Levels are offscale low.

- RCS pressure is 2150 psig and lowering.

- Control Rod H-8 is indicating 32 steps.

- All systems responded normally to actuation signals.

Which ONE of the following actions would be taken?

A. Transition to 02-OHP-4023-ES-0-1, Reactor Trip Response, and initiate boration for a stuck rod.

B. Transition to 02-OHP-4023-ES-0-1, Reactor Trip Response. Rod H-8 condition is expected so boration is not required for a stuck rod.

C. Initiate Safety Injection and continue with 02-OHP-4023-E-0, Reactor Trip Or Safety Injection, as pressurizer pressure is too low.

D. Initiate Safety Injection and continue with 02-OHP-4023-E-0, Reactor Trip Or Safety Injection, as Steam Generator levels are too low.

Answer: B Unit 2 Rod H-8 is required to be less than 35 steps.

All other rods should be less than 10 steps. Plant conditions do not require a Safety Injection.

A - Incorrect - Rod H-8 is expected to be less than 10 steps on Unit 1.

C - Incorrect - RCS pressure will decrease post trip it is still above the SI setpoint.

D - Incorrect - SG levels will decrease to offscale low post trip. AFW will recover levels.

Modified to Unit 2 which allows rod H-8 to indicate up to 35 steps. This changed answer B to correct Answer.

REFERENCE:

02-OHP-4023-E-0, Reactor Trip Or Safety Injection; 02-OHP-4023-ES 1, Reactor Trip Response LESSON PLAN/OBJ: RO-C-EOP03 / #24 KA - 000007 EK3.01 Reactor Trip - Stabilization Knowledge of the reasons for the following responses as they apply to the reactor trip:

Actions contained in EOP for reactor trip RO/SRO Value - (4.0 / 4.6) CFR - 41.5 / 41.10 / 45.6 / 45.13 SOURCE: COOK 2002 NRC #027 (RO#N/A, SRO#25)

Original KA : 000007 - 2.4.48 Reactor Trip

- Emergency Procedures/Plan

- Ability to interpret control room indications to verify the status and operation of system, and understand how operator actions and directives affect plant and system conditions.

DC Cook 2006 NRC Exam 5

NOTE: Unit difference KA 2.2.3 (3.3 / 3.1) - If on Unit 1, correct answer would be "Transition to 01-OHP-4023-ES-0-1. Rod H-8 condition is NOT expected so boration is required for a stuck rod."

Question #4 KA# - 000007 EK3.01 Exam Level - RO Question Source - MODIFIED - COOK 2002 NRC 027 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 6

5. 005 2 Which ONE of the following correctly describes the plant response if the listed instrument fails HIGH while in the LTOP mode?

A. RVLIS Wide Range pressure instrument NPS-110 will cause PORV NRV-153 to OPEN.

B. RVLIS Wide Range pressure instrument NPS-110 will cause PORV NRV-152 to OPEN.

C. Wide Range RCS pressure instrument NPS-121 will cause PORV NRV-153 to OPEN.

D. Wide Range RCS pressure instrument NPS-121 will cause PORV NRV-152 to OPEN.

Answer: C Pressurizer PORV NRV-153 will open if NPS-121 fails open.

A - Incorrect - RVLIS WR pressure does not feed the PRZ Porvs.

B - Incorrect - RVLIS WR pressure does not feed the PRZ Porvs.

D - Incorrect - WR RCS pressure NPS-122 opens PORV NRV-152 Changed Stem & distractors slightly LESSON PLAN/OBJ: RO-C-00202/#22;

REFERENCE:

SOD-00202-001 Rev 1 KA - 000008 AK2.03 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

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

Controllers and positioners RO/SRO Value - (2.5 / 2.4) CFR - 41.7 / 45.7 Original Question Source: RO23 Audit 037-4 (RO#036 /SRO#036) from #1022.doc 1 Original KA - APE 008 AA206 (3.3/3.6)

Question #5 KA# - 000008 AK2.03 Exam Level - RO Question Source - DIRECT - RO23 AUDIT 037-4 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 7

6. 006 1 Which of the following is the basis for establishing / maintaining S/G Narrow Range level between 27%-50% (non-adverse containment) for small or intermediate LOCA's?

A. Maintains a static head of water to reduce any existing S/G tube leakage.

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

C. A RCP may have to be started if 02-OHP-4023-FR-C.1 is entered later in the event.

D. Maintains the water level above the top of the U-tubes to prevent depressurizing S/G.

Answer: B E-1 Basis Document states the purpose of establishing 27% level is to ensure adequate feed flow or S/G inventory to maintain a secondary heat sink for small or intermediate size LOCAs and secondary break accidents. 50% is the S/G narrow range upper control band limit.

A - Incorrect - because it is the basis for S/G level / feed flow for a LARGE break LOCA.

C - Incorrect - The RCP is started in FR-C.1 even if level is not available D - Incorrect - This is the reason for maintaining level during a tube rupture.

REFERENCE:

12-OHP-4023-ES-1.2 Step 9 pg. 25 LESSON PLAN/OBJ: RO-C-EOP09/#36 KA - 000009 EK2.03 Small Break LOCA Knowledge of the interrelations between the small break LOCA and the following:

S/Gs RO/SRO Value - (3.0 / 3.3) CFR - 41.7 / 45.7 SOURCE: INPO # 24053 Salem Unit 1 - 5/5/2003 Original Quest. KA - 000009.K2.03 Question #6 KA# - 000009 EK2.03 Exam Level - RO Question Source - DIRECT-INPO - SALEM 2003 - 24053 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 8

7. 007 1 Unit 2 Reactor Startup is in progress with Reactor Power at 2% and rising.

The following conditions exist:

- RCP No. 3 Lower Bearing water temperature is 208°F and stable.

- RCP No. 3 Motor Bearing temperature is 187°F and stable.

- RCP No. 3 Seal Leakoff temperature is 179°F and stable.

- RCP No. 3 Motor Temperature is 148°C and rising.

- Annunciator 107 Drop 40, RCP Motor Overheated - LIT Which ONE of the following operator actions MUST be taken based upon these conditions?

A. Do NOT trip the reactor. Trip the No. 3 RCP. Power must be maintained less than 5%

(Mode 2).

B. Initiate reactor shutdown per 02 OHP 4021.001.003, Power Reduction and trip the No. 3 RCP within 8 Hours.

C. Manually trip the reactor, Enter 02 OHP 4023.E 0, Reactor Trip or Safety Injection, perform immediate actions, then trip the No. 3 RCP.

D. Immediately Trip the No. 3 RCP. Initiate reactor shutdown per 02 OHP 4021.001.003, Power Reduction and be in Mode 3 within 6 Hours.

Answer: B The RCP Motor Temperature has exceeded the limit of 145°C. This requires the reactor to be shutdown and the RCP tripped within 8 Hours.

A - Incorrect - RCP Lower bearing temperature is elevated but has not exceeded the trip setpoint of 225°F. (Motor Bearing Temperature is limited to 200°F). Additionally, the Plant is not analyzed/licensed to operate with less than 4 RCPs. Tech Specs require Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with less than 4 RCP loops.

C - Incorrect - RCP Lower bearing temperature is elevated but has not exceeded the trip setpoint of 225°F. (Motor Bearing Temperature is limited to 200°F). An Immediate trip is not required for the RCP Motor Temperature.

D - Incorrect - RCP Lower bearing temperature is elevated but has not exceeded the trip setpoint of 225°F. (Motor Bearing Temperature is limited to 200°F). An Immediate trip is not required for the RCP Motor Temperature. The Plant is not analyzed/licensed to operate with less than 4 RCPs.

Modified Question to Add Motor Temperature and Alarm. Changed Lower Bearing temperature to elevated status (but < trip). Changed Distractors A & D to include power requirements. Answer Changed from C to B.

REFERENCE:

02-OHP-4022-002-001, Malfunction Of A Reactor Coolant Pump step 10.

LESSON PLAN/OBJ: RO-C-AOP-11 / #19 KA - 000015 AA2.09 017 Reactor Coolant Pump (RCP) Malfunctions Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions:

When to secure RCPs on high stator temperatures RO/SRO Value - (3.4 / 3.5) CFR - 43.5 / 45.13

DC Cook 2006 NRC Exam 9

SOURCE: INPO # 27639 Cook 1 - 4/19/2004 Original Quest. KA - 000015 AK1.02 Question #7 KA# - 000015 AA2.09 Exam Level - RO Question Source - MODIFIED - COOK 2004-27639 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 10

8. 008 2 With 120 gpm letdown in service and blender controls in automatic on Unit 1, the VCT level transmitter QLC-452 fails to 100%.

With NO operator action, actual VCT level will:

A. lower to 2.5% and RWST swapover will occur automatically.

B. lower to 14% and stabilize due to auto makeup flow.

C. lower to zero and eventually result in loss of CCP suction.

D. rise to 87% and stabilize due to full divert flow.

Answer: C A Failure of QLC-452 high will cause Letdown to divert to the HUT. Auto VCT makeup will attempt to control VCT level between 14 and 24% (Makeup will not keep up with 120 gpm letdown). Eventually the VCT level will lower to the RWST switchover setpoint.

With 452 failed high, the coincidence for switchover will not be made up and level will continue lowering to 0%. At this time suction will be lost to the charging pumps and they will trip.

A - Incorrect - The auto swapover will not occur with QLC-452 failed high.

B - Incorrect - VCT Makeup is not sufficient to replace the 120 gpm lost through diversion to the HUT.

D - Incorrect - The level will lower as the letdown is diverted to the HUT. If QLC-451 failed low this would be the correct response.

REFERENCE:

RO-C-AOP-6 Abnormal Operating Procedures - Day 6 pg. 39 LESSON PLAN/OBJ: RO-C-AOP-6/#AOP-6.15, RO-C-00300/#12, 14,19 KA - 000022 AA2.02 Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Pump Makeup:

Charging pump problems RO/SRO Value - (3.2 / 3.7) CFR - 43.5 / 45.13 Original Question Source: 01003C0013-3 Original KA: SYS 004 A.03 (3.3/3.3)

EXAM/QUIZZES: R921517; RO15 AUDIT EXAMNRC(286); RO20SRO3; SRO Cert 2 &

3; RQ2703V; RQ2703E; RQ2703B Question #8 KA# - 000022 AA2.02 Exam Level - RO Question Source - DIRECT - 01003C0013-3 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 11

9. 009 1 The following Unit 2 conditions exist:

- The unit is in MODE 5 at 150oF with the RCS intact.

- Both trains of RHR are available, with the 2 West RHR train is in operation.

- Both PORVs are lined up for LTOP mode of operation.

- The West CCP and Both SI pumps are Tagged out.

Due to indication of cavitation on 2 West RHR Pump, the crew enters 02-OHP-4022-017-001, Loss of RHR Cooling.

Per the procedure, the crew isolates both trains of RHR by placing both RHR pumps in Pull-to-Lock and closes RHR suction valves IMO-128 and ICM-129.

How does this action affect the Technical Specification operability of Overpressure Protection Systems?

A. Technical Specification requirements are met. Both PORVs are available for overpressure control.

B. Technical Specification requirements are NOT met. The RHR suction relief must be available for overpressure protection.

C. Technical Specification requirements are NOT met. The Charging Pump must be Isolated/

Tagged -Out.

D. Technical Specification requirements are met. Isolating RHR suction valves does not affect operability of the RHR suction relief.

Answer: A Tech. Spec requirements are met. Both PORVs are available for overpressure control.

T.S. 3.4.12 requires 2 overpressure protection flowpaths, whether both PORVs a PORV and RHR suction relief. Closing the RHR suction valves isolates the RHR suction reliefs, but the PORVs are still available.

B - Incorrect - The RHR suction Relief is not required since 2 PORVs are available and only 1 CCP is available. (True if both CCPs available)

C - Incorrect - 1 CCP may be available with 2 PORVs.

D - Incorrect - The Suction Relief is downstream of the RCS to RHR isolation Valves.

REFERENCE:

Technical Specifications 3.4.12 LESSON PLAN/OBJ: RO-C-01700/#13 KA - 000025 2.2.23 Loss of Residual Heat Removal System (RHRS)

Equipment Control Ability to track limiting conditions for operations.

RO/SRO Value - (2.6 / 3.8) CFR - 43.2 / 45.13 SOURCE: INPO # 24615 Seabrook 1 - 5/30/2003 Original Quest. KA - 025.2.1.12 Question #9 KA# - 000025 2.2.23 Exam Level - RO Question Source - DIRECT-INPO - SEABROOK 2003-24615 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 12

10. 010 2 The following plant conditions exist:

- RCS pressure is 1600 psig and lowering

- Pressurizer is empty

- Containment pressure is stable at 1.5 psig You have been directed to close 2-WMO-734, ESW outlet valve for CCW HX 2 East.

What actions; if any, must be taken to close the valve?

A. The valve control switch must be put in pull-to-stop and must be closed locally.

B. When the valve reaches its preset position you can close the valve using the control switch.

C. No action is necessary, the valve has already been closed by an automatic signal.

D. SI must be reset to close the valve using the control switch.

Answer: B Three-position control switches in control room CLOSE - Closes valve PULL TO STOP - No longer needed since modified to hold-to-operate OPEN - Opens valve The listed plant conditions indicate an SI so the valve will automatically move to preset intermediate position on receipt of an SI signal corresponding to an approximate flow rate of 5000 gpm. The valve can be closed by holding the control switch to close after it has reached its throttle position.

A - Incorrect - The valve can be closed with the hand switch after it has reached throttle position.

C - Incorrect - The valve will position open to a preset position.

D - Incorrect - The valve will NOT close until it reaches it throttle position.

Removed SG level from Stem. Changed from "told" to directed. Changed "position" switch to "control" switch. Changed distractor A to Pull-to-Stop vs. lock and "locally" vs.

Manually. Distractor D changed to include After SI reset.

REFERENCE:

RO-C-01900, Essential Service Water System pg. 22 LESSON PLAN/OBJ: RO-C-01900/#11 KA - 000026 AK3.01 Loss of Component Cooling Water (CCW)

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

The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the CCW/nuclear service water coolers RO/SRO Value - (3.2 / 3.5) CFR - 41.5 / 41.10 / 45.6 / 45.13 SOURCE: INPO # 19517 Cook 1 - 5/21/2001 Original Quest. KA - 000026.K3.01 Question #10 KA# - 000026 AK3.01 Exam Level - RO

DC Cook 2006 NRC Exam 13 Question Source - DIRECT - COOK 2001 Q#65 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 14

11. 011 2 Assume that the Pressurizer master pressure controller output were failed AS IS just before a large, rapid secondary load rejection.

Which of the following would occur naturally in the Pressurizer to limit the magnitude of the resulting pressure transient on the primary system?

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

B. An insurge of hotter water heats the Pzr. More liquid then flashes to steam helping to limit the resulting pressure drop in the RCS.

C. An insurge of cooler water compresses the steam space in the Pzr. Steam is condensed to water helping to limit the overall pressure increase in the RCS.

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

Answer: C A secondary load rejection will cause RCS temperature to rise. Pressurizer pressure will initially follow the level trend. As Tavg initially increases, level will increase causing an insurge which causes pressure to rise. Since the water from the insurge is cooler than the pressurizer liquid, steam is condensed to water helping to limit the overall pressure increase in the RCS.

A - Incorrect - The load rejection will cause an insurge. Plausible since this is the response for a secondary load increase.

B - Incorrect - The insurge liquid from the RCs hot leg is still significantly cooler than the Pressurizer liquid temperature.

D - Incorrect - The load rejection will cause an insurge. Plausible since an outsurge would remove some energy from the pressurizer.

REFERENCE:

RO-C-GF26 OBSERVING REACTOR BEHAVIOR DURING REACTOR TRANSIENTS pg. 6-12 LESSON PLAN/OBJ: RO-C-GF26/#1 KA - 000027 AK1.03 Pressurizer Pressure Control (PZR PCS) Malfunction Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions:

Latent heat of vaporization/condensation RO/SRO Value - (2.6 / 2.9) CFR - 41.8 / 41.10 / 45.3 SOURCE: INPO # 21479 Braidwood 1 - 7/17/2002 Original Quest. KA - 027.ak1.03 Question #11 KA# - 000027 AK1.03 Exam Level - RO Question Source - DIRECT-INPO - BRWD 2002 - 21479 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 15

12. 012 1 The following plant conditions exist:

- Unit 1 has experienced an Anticipated Transient Without Scram (ATWS), and has implemented 01-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS. QMO-410, emergency boration to CCP suction valve will NOT OPEN.

- SI has been initiated.

Which one of the following identifies the correct actions and reason to satisfy 01-OHP-4023-FR-S.1 Step 5, Initiate Emergency Boration Of RCS?

A. 1) Open at least one CCP suction from RWST valve: 1-IMO-910 -OR-1-IMO-911

2) Close at least one CCP suction from VCT valve: 1-QMO-451 -OR-1-QMO-452

3) Maximize charging flow.

4) Verify letdown flow established.

This ensures at least 70 gpm of Borated RWST water is injected through the Normal Charging flowpath.

B. 1) Open both boration valves: 1-QRV-400 -AND-1-QRV-421

2) Start both boric acid transfer pumps in FAST speed.

This ensures at least 36 gpm of Borated BAST water is injected through the Normal Charging flowpath.

C. 1) Open at least one of the boration valves: 1-QRV-400 -OR-1-QRV-421

2) Start both boric acid transfer pumps in FAST speed.

This ensures at least 36 gpm of Borated BAST water is injected through the Normal Charging flowpath.

D. 1.) Verify Pressurizer Pressure is <2335 psig.

This ensures at least 70 gpm of Borated RWST water is injected through the BIT injection lines.

Answer: D On a Safety Injection, Both CCPs start, IMO-910 & 911 Open, and the BIT Injection Lines Open providing RWST injection to the RCS. Step 5b RNO has the Operator go to step 5c which checks Pressurizer pressure if a SI is actuated.

A - Incorrect - The Normal Charging flowpath will isolate on a SI. Plausible since this is an acceptable flowpath if SI is not actuated.

B - Incorrect - The Normal Charging flowpath will isolate on a SI. Plausible since this is an acceptable flowpath if SI is not actuated.

C - Incorrect - The Normal Charging flowpath will isolate on a SI and BOTH QRV-400 and 421 are required to be open. Plausible since this is similar to the flowpath if SI is not actuated. The option for aligning the CCPs to the RWST require only 1 of the valves to be opened.

REFERENCE:

01-OHP-4023-FR-S.1 Response to Nuclear Power Generation/ATWS Step 5 LESSON PLAN/OBJ: RO-C-EOP04/#15 KA - 000029 EA1.05 Anticipated Transient Without Scram (ATWS)

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

BIT outlet valve switches RO/SRO Value - (3.7 / 3.6) CFR - 41.7 / 45.5 / 45.6 Question #12

DC Cook 2006 NRC Exam 16 KA# - 000029 EA1.05 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 17

13. 013 6 The following plant conditions exist:

- Ann. 205, Drop 28, SFP Temp High - LIT.

- South (U2) SFP pump out for maintenance.

- South SFP Hx is available for use.

- North SFP Hx shell develops a large leak requiring Hx to be isolated.

Procedure 12-OHP-4022-018-001, Loss of Spent Fuel Pit Cooling allows cross-tie of the North SFP Pump through a 3" line to the South SFP Hx.

This action will...

A. maintain the SFP temperature at normal conditions until repairs are made.

B. prevent the SFP from boiling, but result in operation at elevated temperature.

C. prevent the SFP from boiling if used in conjunction with 125 gpm makeup.

D. extend the time to SFP boiling to allow time for repair of system.

Answer: D Crosstieing the North Pump through the 3" line to the South HX will delay the time to boiling but will not prevent boiling unless makeup flow of 175 gpm is also used.

A - Incorrect - Temperatures will rise to the boiling point.

B - Incorrect - Temperatures will rise to the boiling point.

C - Incorrect - Temperatures will rise to the boiling point unless makeup flow of 175 gpm is also used.

REFERENCE:

12-OHP-4022-018-001, Loss of Spent Fuel Pit Cooling Step 9 & Note LESSON PLAN/OBJ: RO-C-AOP12/AOP12.9 KA - 000036 AK3.03 Fuel Handling Incidents Knowledge of the reasons for the following responses as they apply to the Fuel Handling Incidents:

Guidance contained in EOP for fuel handling incident RO/SRO Value - (3.7 / 4.1) CFR - 41.5 / 41.10 / 45.6 / 45.13 SOURCE: Master Bank AOP1CAOP12.9-3 Question #13 KA# - 000036 AK3.03 Exam Level - RO Question Source - DIRECT - AOP1CAOP12.9-3 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 18

14. 014 10 During power operation, SG tube leakage was detected and estimated at 100 gpm when RCS pressure was 2200 psig and SG pressure was 800 psig. The plant was shutdown and a cooldown initiated.

Which ONE of the following is the approximate current leak rate if RCS pressure is 1350 psig and SG pressure is 1000 psig? Assume the break size has not changed.

A. Approximately 25 gpm B. Approximately 50 gpm C. Approximately 75 gpm D. equal to the initial leak rate of 100 gpm Answer: B Break flow is Proportional to the Square Root of the Pressure Differential.

)

800 2200

(

int

Flow

)

1000 1350

(

final Flow Flow is 50% of initial or 50 gpm A - Incorrect - Differential pressure is 1/4 of original but break flow should be proportional to the square root of DP.

C - Incorrect - This value is 75% of break flow (NRC reccomendation).

D - Incorrect - This is original value. The DP changed and so does break flow.

Modified from Cook 2004 NRC Exam #10 - Changed Question Stem from 50 gpm to 150 gpm. Also changed Distracters accordingly and removed percentages.

REFERENCE:

RO-C-EOP05, SI Termination, ECCS Flow Reduction, and SI Reinitiation and Actuation; RO-C-GF27, Sensors and Detectors LESSON PLAN/OBJ: RO-C-GF27 / #10 KA - 000038 EK1.02 Steam Generator Tube Rupture (SGTR)

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

Leak rate vs. pressure drop RO/SRO Value - (3.2 / 3.5) CFR - 41.8 / 41.10 / 45.3 SOURCE:2004 NRC Exam Q#10 Cook 1 - 4/29/2004 Original Quest. KA - 000037 AK1.02 Steam Generator (S/G) Tube Leak Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak:

Leak rate vs. pressure drop RO-3.5 SRO-3.9 Question #14 KA# - 000038 EK1.02 Exam Level - RO Question Source - MODIFIED - COOK04-010

DC Cook 2006 NRC Exam 19 Cognitive/Difficulty Level - H/4

DC Cook 2006 NRC Exam 20

15. 015 2 The following plant conditions exist:

- A loss of all FW and AFW has occurred.

- The operating crew has entered FR-H.1, Loss Of Secondary Heat Sink.

- The RO identifies that the East CCP has tripped and that the West CCP will NOT Start.

Which one of the following describes the consequences of the CCP failures?

A. RCS Bleed and Feed cooling must be established immediately to ensure sufficient SI flow.

B. A Red Path on the Core Cooling Critical Safety Function will develop due to loss of RCS inventory with no available makeup.

C. The RCS will not depressurize quickly enough to ensure sufficient SI flow to provide RCS heat removal, so other RCS openings will have to be established.

D. RCS Bleed and Feed cooling must NOT be initiated and secondary depressurization to inject condensate pump flow must be immediately initiated.

Answer: A Step 2 of OHP-4023-FR-H.1 requires that Bleed and Feed be initiated immediately initiated if the CCPs are not available. If it is known that no CCP is available for bleed and feed, then all RCPs are stopped and bleed and feed is initiated immediately. The RCS will have to depressurize below the shutoff head of the SI pumps before any ECCS injection will occur, so the bleed and feed must be initiated prior to the loss of heat removal capability of the SGs.

B - Incorrect. Although a red condition on Core Cooling may eventually occur, there is available makeup with SI.

C - Incorrect. Bleed and Feed is started early enough so that the RCS can depressurize far enough to allow SI injection. Additional Openings are used if all PORVs are NOT available.

D -Incorrect. Action to align condensate pumps may still be taken, but not as a contingency to Bleed and Feed.

Modified BV question Stem from loss of PORV to loss of CCP. Changed Answer &

Distracters.

REFERENCE:

12-OHP-4023-FR-H.1 Loss of Heat Sink Background Step 2. OHP-4023-FR-H.1 Loss of Heat Sink Step 2.

LESSON PLAN/OBJ: RO-C-EOP11/#7 KA - 000054 AA1.04 Loss of Main Feedwater (MFW)

Ability to operate and/or monitor the following as they apply to the Loss of Main Feedwater (MFW):

HPI, under total feedwater loss conditions RO/SRO Value - (4.4 / 4.5) CFR - 41.7 / 45.5 / 45.6 SOURCE: INPO # 25008 Beaver Valley 1 - 12/1/2002 Original Quest. KA - E05.EK1.1 Question #15

DC Cook 2006 NRC Exam 21 KA# - 000054 AA1.04 Exam Level - RO Question Source - MODIFIED - BEAVERVAL 2002 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 22

16. 016 3 The plant has experienced a Loss of All Offsite AC and only the 2AB EDG is operating.

Due to the Loss of Power, the PPC is not available. What contingency action is taken to compensate for the unavailability of the PPC?

A. Emergency Response Data System (ERDS) is activated.

B. Safety Parameter Display System (SPDS) is transmitted to the NRC.

C. PMP-2080-EPP-100 Data sheet 1, Technical Information sheet, is completed every 15 minutes.

D. PMP-2080-EPP-100 Data sheet 4, Plant Status sheet, is completed every 60 minutes.

Answer: C Upon loss of the PPC and/or EDR the PMP-2080-EPP-100 Data sheet 1, Technical Information sheet, is completed every 15 minutes and sent to applicable facilities.

A - Incorrect - ERDS is normally activated within 60 Minutes but will not be available if the PPC is not working.

B - Incorrect - SPDS will not be available if the PPC is not working.

D - Incorrect - This is the information sheet that is used for guidance when communicating with the NRC.

REFERENCE:

PMP-2080-EPP-100 Section 3.2.3.b.4 LESSON PLAN/OBJ: ST-C-EP04/#3 KA - 000055 2.4.29 Loss of Offsite and Onsite Power (Station Blackout)

Emergency Procedures/Plan Knowledge of the emergency plan.

RO/SRO Value - (2.6 / 4.0) CFR - 43.5 / 45.11 Original Question Source: Bank 12EPPC0403-1 Original KA 2.4.40 (2.3/4.0)

EXAM/QUIZZES: RQ2603C; RQ2603D; RQ2603E; RQ2603A; RQ2603B; RQ2603C; RQ2603E; RO22EPPRO; RO22EPPSRO Question #16 KA# - 000055 2.4.29 Exam Level - RO Question Source - DIRECT - 12EPPC0403-1 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 23

17. 017 1 Unit 2 was operating at 100% power when a reactor trip occurred due to a loss of offsite power. The operators completed the actions of 02-OHP-4023-ES-0.1, Reactor Trip Response, and have transitioned to 02-OHP-4023-ES-0.2, Natural Circulation Cooldown, where they are initiating a natural circulation cooldown.

At the onset of the natural circulation cooldown, which ONE of the following processes will remove the MOST heat from the Reactor Vessel HEAD?

A. All CRDM fans running.

B. Upper head bypass flow.

C. Heat losses to ambient.

D. The 25 F/hr natural circulation cooldown of the RCS.

Answer: A CRDM fan operation results in a considerable amount of heat removal from the upper head region. The results from several tests at domestic and foreign plants indicate that the Control Rod Drive Mechanism (CRDM) cooling fans aid significantly in removing heat from the upper head area. At 600°F this will result in a cooldown rate of the upper head of approximately 21°F/hr.

B - Incorrect - Analysis assumed for Cook Nuclear Plant results in this average cooldown rate of the upper head fluid to be about 10°F/hr when a 25°F/hr N/C cooldown rate of the bulk of the RCS is in progress.

C - Incorrect - Losses to ambient result in <1°F/hr cooldown rate.

D - Incorrect - This cooling would come from upper head bypass flow which is about 10°F/hr when a 25°F/hr N/C cooldown rate of the bulk of the RCS is in progress.

REFERENCE:

12-OHP 4023 ES-0.2 PSBD Step 5, RO-C-EOP03 Plant Trips, Diagnosing Accidents, Natural Circulation Cooldown, E-0 Series EOPs, and Background Information pg. 74-78 LESSON PLAN/OBJ: RO-C-EOP03/#8, 12 KA - 000056 AK1.01 Loss of Offsite Power Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power:

Principle of cooling by natural convection RO/SRO Value - (3.7 / 4.2) CFR - 41.8 / 41.10 / 45.3 SOURCE: INPO # 20211 Cook 1 - 9/10/2001 Retake #46 From Bank 01EOPC0309-1 Original Quest. KA - 056.AK1.01 Question #17 KA# - 000056 AK1.01 Exam Level - RO Question Source - DIRECT-INPO - COOK 2001 RT Q#46 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 24

18. 018 1 The following plant conditions exist:

- Unit 1 is in Mode 5

- Unit 2 is in Mode 6.

- Unit 2 Core offload is currently underway.

- The protected train is A. All equipment needed for a proper protected train alignment is available.

- DC bus 2AB is deenergized for maintenance.

- DC bus 2CD deenergizes due to equipment failure.

What action must occur due to the deenergized bus?

A. Immediately suspend core alterations and movement of irradiated fuel.

B. Immediately suspend core alterations. Fuel movement may continue in the Auxiliary Building.

C. Re-energize DC bus 2AB or 2CD within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or declare the Unit 1 ESW system inoperable.

D. Re-energize DC bus 2AB or 2CD within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or suspend core alterations and movement of irradiated fuel.

Answer: A Technical Specification 3.8.5 requires one DC train to be OPERABLE. Immediate suspension of all core alterations fuel movement is required if both are lost.

B - Incorrect - Fuel Movement in the Auxiliary Building must also be stopped. Plausible since several specs allow Fuel Movement while requiring Core Alts to be stopped.

C - Incorrect - The Unit 1 ESW System/Train is declared inoperable during Modes 1-4 if the Unit 2 DC is inoperable (TS 3.8.4 action E)

D - Incorrect - Action is required immediately. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action is from Technical Specification 3.8.4 DC Sources-Operating loss of one bus.

Changed to loss of all DC and required suspension of core alts/fuel movement

REFERENCE:

ITS 3.8.5 LESSON PLAN/OBJ: RO-C-ADM13/#ADM13.3 KA - 000058 2.2.26 Loss of DC Power Equipment Control Knowledge of refueling administrative requirements.

RO/SRO Value - (2.5 / 3.7) CFR - 43.5 / 45.13 SOURCE: INPO # 24707 Seabrook 1 - 5/30/2003 Original Quest. KA - 058.AK1.01 Question #18 KA# - 000058 2.2.26 Exam Level - RO Question Source - MODIFIED - SEABROOK 2003 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 25

19. 019 6 Which of the following is the expected response to a High Radiation Signal on the ERS-7401 Unit 1 Control Room Area Radiation Monitor?

A. Only the Unit 1 East and West Pressurization fans start and Dampers align.

B. Either the Unit 1 East -OR-West Pressurization fan starts and Dampers align depending on which Control Room Air Handling Unit is in Service.

C. Only the Unit 1 and Unit 2 East Pressurization fans start and Dampers align.

D. Both the Unit 1 and Unit 2 East and West Pressurization fans start and Dampers align.

Answer: A 12-OHP-4024-139 drop #12 states that both the East and West Pressurization fans will start and the associated dampers align.

B - Incorrect - Both the East and West Pressurization fans will start and the associated dampers align. One Pressurization fan must be manually stopped.

C - Incorrect - Both the Unit 1 East and West Pressurization fans will start and the associated dampers align. A SI signal on either Unit starts All Pressurization fans on both units. One Pressurization Fan on each Unit is manually stopped.

D-Incorrect - Both the Unit 1 East and West Pressurization fans will start and the associated dampers align. A SI signal on either Unit starts All Pressurization fans on both units. One Pressurization Fan on each Unit is manually stopped.

REFERENCE:

12-OHP-4024-139 drop #12 LESSON PLAN/OBJ: RO-C-01350/#4 KA - 000061 AK2.01 Area Radiation Monitoring (ARM) System Alarms Knowledge of the interrelations between the Area Radiation Monitoring (ARM) System Alarms and the following:

Detectors at each ARM system location RO/SRO Value - (2.5 / 2.6) CFR - 41.7 / 45.7 Question #19 KA# - 000061 AK2.01 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 26

20. 020 4 Unit 1 is operating at 100%, steady state condition, when Annunciator Panel 118, Drop 84, ESW PIPE TUNNEL SUMP LEVEL HI-HI, alarm is received. East ESW supply and return flows indicate an abnormal high differential.

Unit 2 has reported that 02-OHP-4022-019-001, ESW SYSTEM LOSS/RUPTURE has been implemented.

What action(s) would be taken by BOTH Unit control room operators?

A. Align alternate ESW cooling to non affected diesel generators.

B. Stop affected header ESW pumps.

C. Place associated diesel generators in tripped condition.

D. Close affected header unit crosstie valves.

Answer: D When ESW supply and return flows indicate an abnormal high differential both units are directed to isolate the affected header unit crosstie valves.

A - Incorrect - The alternate ESW cooling to the EDGs is isolated per step 15.

B - Incorrect The header crosstie valves are closed first to prevent loss of the opposite unit.

C - Incorrect - The affected EDG is placed in the tripped condition after the headers are isolated and ESW pumps are stopped.

REFERENCE:

RO-C-AOP-5, Abnormal Operating Procedures - Day 5 pg. 52; 01-OHP 4022.019.001, ESW SYSTEM RUPTURE, Step 10 LESSON PLANS/OBJ: RO-C-AOP-5/#AOP5.15 KA - 000062 AK3.03 Loss of Nuclear Service Water Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water:

Guidance actions contained in EOP for Loss of nuclear service water RO/SRO Value - (4.0 / 4.2) CFR - 41.5 / 41.10 / 45.6 / 45.13 Original Question Source: Bank 12AOPS0515-1 K/A: 076000 A2.01 (3.5/3.7)

EXAM/QUIZZES: Q1805D; Q1805E; R1800B5; 1800B3; Q1908C; Q1908A; RQ2804E; RQ2804B Question #20 KA# - 000062 AK3.03 Exam Level - RO Question Source - DIRECT - 12AOPS0515 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 27

21. 021 1 The control room has been evacuated per 02-OHP-4025-001-001, Emergency Remote Shutdown.

The crew has performed 02-OHP-4025-LS-6-1, Seal Injection from CVCS Cross-Tie and 02-OHP-4025-LS-6-3, BIT Injection alignment.

Which ONE of the following describes the source and method of monitoring borated water addition?

Prior to aligning RHR for shutdown cooling, RCS boration is accomplished by...

A. manually aligning Unit 1 Boric Acid makeup to feed Unit 2 through the crosstie. Flow is tracked by Unit 1 Boric Acid flow recorder.

B. locally aligning Unit 2 Boric Acid makeup to feed Unit 2 through the crosstie. Flow is tracked by Boric Acid Flow indicator on the Local Instrument Panel.

C. manually aligning Unit 1 CCP suction to the RWST to feed Unit 2 through the crosstie.

Flow is tracked by the sum of local seal injection flows and the Unit 1/2 CCP Crosstie Flow indicator.

D. manually aligning Unit 1 CCP suction to the RWST to feed Unit 2 through the crosstie.

Flow is tracked by the Unit 1/2 CCP Crosstie Flow indicator.

Answer: D Prior to initiating the Unit Crosstie, Unit 1 is directed to trip the unit and align CCP suction to the RWST. The Total flow from Unit 1 to Unit 2 is indicated on the Unit 1/2 CCP Crosstie Flow indicator.

A - Incorrect - Unit 1 is aligned to the RWST. If Unit 1 was aligned to the VCT, this would be a valid flowpath.

B - Incorrect - Unit 2 CCPs are placed in Lock-out prior to evacuating the control room.

Unit 2 Boric Acid flow would be at too low of pressure to inject into the discharge pressure of the Unit 1 CCP through the crosstie. The Pressurizer level is monitored at the local panels.

C - Incorrect - All Unit Crosstie flow travels through the Unit 1/2 CCP Crosstie Flow indicator. Adding the seal injection flow would count that flow twice.

REFERENCE:

02-OHP-4025-001-001, Emergency Remote Shutdown, 02-OHP-4025-LS-6-1, Seal Injection from CVCS Cross-Tie, 02-OHP-4025-LS-6-3, BIT Injection, SOD00300-001 LESSON PLAN/OBJ: RO-C-EC002/#4 KA - 000068 AA1.08 Control Room Evacuation Ability to operate and/or monitor the following as they apply to the Control Room Evacuation:

Local boric acid flow RO/SRO Value - (4.2 / 4.2) CFR - 41.7 / 45.5 / 45.6 Question #21 KA# - 000068 AA1.08 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 28

22. 022 10 Which ONE of the following conditions would require an LCO Action Statement entry for Technical Specification Section 3.6, Containment Systems?

A. While in MODE 1, an electrician opens the outer airlock door at the lower containment access without prior approval.

B. While in MODE 3, Containment internal pressure is found to be -0.5 psig prior to placing Containment Purge in service.

C. While in MODE 4, The outer airlock door interlock is discovered to be non-functional.

D. While in MODE 5, during performance of the Overall Integrated Containment Leakage Rate Test, Containment leakage exceeds the maximum allowable Technical Specification leakage rates.

Answer: C The airlock door interlocks are required to be operable in Modes 1 - 4 per T.S. 3.6.2.

A - Incorrect - One airlock door may be opened for normal entry / exit while in Modes 1 -

4 per T.S. 3.6.2.

B - Incorrect - Containment pressure limits are -1.5 to +0.3 psig during Modes 1 - 4 per T.S. 3.6.4.

D - Incorrect - Overall Containment leakrate only applies in Modes 1 - 4 per T.S. 3.6.1 Changed correct answer by modifying distractors B & C to reflect airlock interlock vs.

equipment hatch & Internal pressure to Mode 3 vs. 4. Modified wording in stem to ask for LCO entry instead of "Loss of Cont. Operability" & feedback slightly for ITS. 1-05

REFERENCE:

Tech Spec 3.6.1 & Basis 3.6.1 pg. B3.6.1-1 LESSON PLAN/OBJ: RO-C-TS01 / #12 KA - 000069 AK2.03 Loss of Containment Integrity Knowledge of the interrelations between the Loss of Containment Integrity and the following:

Personnel access hatch and emergency access hatch RO/SRO Value - (2.8 / 2.9) CFR - 41.7 / 45.7 SOURCE: INPO # 22812 Cook 1 - 12/9/2002 #016 (RO#12/SRO#15)

From Dev bank TS-38 Original Quest. KA - 000069.K2.03 Loss of Containment Integrity

- Knowledge of the interrelations between the Loss of Containment Integrity and the following: Personnel access hatch and emergency access hatch Question #22 KA# - 000069 - AK2.03 Exam Level - RO Question Source - MODIFIED - COOK 2002 NRC 016-1 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 29

23. 023 2 Unit 2 experienced a reactor trip from 100% power. The operators have transitioned from 02-OHP-4023-E-0 Reactor Trip or Safety Injection to 02-OHP-4023-ES-0.1 Reactor Trip Response.

ERA-8309, Reactor Coolant Filter Room has a high alarm.

ERA-8303, East CCP Room shows a rising trend, but has yet to alarm.

The SRO implements 2-OHP-4022-002-019, High Reactor Coolant Activity or Failed Fuel.

Which ONE of the following describes the correct actions to take with the CVCS system and the basis for these actions?

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

B. Letdown is diverted to the CVCS HUT to limit radiation levels in the charging pump area.

C. Charging and Letdown are maximized through the letdown demineralizers to maximize clean up.

D. Excess Letdown is placed in service through the letdown demineralizers to maximize clean up.

Answer: C 12-OHP-4022-002-019 directs the operator to verify Letdown lineup and maximize letdown flow to help reduce RCS Activity.

A - Incorrect - CVCS is used to cleanup the RCS so flow is maximized not isolated.

B - Incorrect - Letdown is not diverted. The flow is maximized through the demeralizers to cleanup the RCS. Diverting would create added waste.

D - Incorrect - Excess letdown does not flow through the letdown demineralizers.

REFERENCE:

12-OHP-4022-002-019, High Reactor Coolant Activity or Failed Fuel LESSON PLAN/OBJ: RO-C-AOP07/#AOP7.9 KA - 000076 AA2.02 High Reactor Coolant Activity Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity:

Corrective actions required for high fission product activity in RCS RO/SRO Value - (2.8 / 3.4) CFR - 43.5 / 45.13 SOURCE: INPO # 26703 Indian Point 3 (Unit) - 12/11/2003 Original Quest. KA - 000076AK3.05 Question #23 KA# - 000076 AA2.02 Exam Level - RO Question Source - DIRECT-INPO - IP3 2003 - 26703 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 30

24. 024 1 The following conditions exist:

A plant startup is in progress.

Reactor power is 25%.

All control systems are in AUTOMATIC.

Control Bank D rods are at 150 steps.

The lower detector for power range nuclear instrument channel N41 fails HIGH, causing channel total power indication to increase to 70%.

Assuming no operator action is taken, what will be the response of the rod control system?

A. Rods will not move.

B. Rods will drive all the way in.

C. Rods will drive in some distance then stop.

D. Rods will insert, and then withdraw.

Answer: D When N41 fails high, the rate circuit will sense that Reactor power is rising faster than Turbine power, and will send a signal to insert rods. After the rate of change signal has had time to decay away the Tave deviation generated (by inserting rods) will cause the rods to withdraw.

A - Incorrect - The Power mismatch circuit uses the highest channel, which N41 will become as it fails. Plausible since NI failure low has no impact.

B - Incorrect - The Power Mismatch signal will decay away. Plausible for a Tave ( or Loop Temp) failing high. (Rods continue to insert.)

C - Incorrect - Rods will withdraw due to the Tave mismatch. Plausible since a Rod Stop normally stops rods from withdrawing when the plant is at a higher power level.

REFERENCE:

RO-C-01300, EXCORE NUCLEAR INSTRUMENTATION SYSTEM -

HANDOUT #3 POWER SUPPLY and INSTRUMENT FAILURES pg. 3 LESSON PLAN/OBJ: RO-C-01300/#12 KA - 001000 A1.06 Control Rod Drive System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CRDS controls including:

Reactor power RO/SRO Value - (4.1 / 4.4) CFR - 41.5 / 45.5 SOURCE: INPO # 27108 Millstone 3 - 7/16/2004 Original Quest. KA - 001.K6.02 Question #24 KA# - 001000 A1.06 Exam Level - RO Question Source - DIRECT-INPO - MILLSTONE2004-27108 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 31

25. 025 1 While at 20% power with a power ascension in progress, RCP 11 trips due to an overcurrent condition.

No operator action has been taken and no rod motion has occurred.

Which ONE (1) of the following describes the INITIAL reactor and Loop 11 response?

A. A reactor trip WILL NOT occur and Loop 11 Tavg will LOWER.

B. A reactor trip WILL NOT occur and Loop 11 Tavg will RISE.

C. A reactor trip WILL occur and Loop 11 Tavg will LOWER.

D. A reactor trip WILL occur and Loop 11 Tavg will RISE.

Answer: A A single RCP trip will not cause an automatic reactor trip below P-8. Once the RCP trips, The entire associated loop temperature will go to the Tcold or SG Saturation Temperature causing the Tavg to lower B. Incorrect. Tavg will lower C. Incorrect. No trip below 29% power D. Incorrect. No trip below 29% power and Tavg will lower

REFERENCE:

RO-C-TRANS4, RCS Loop Flow Transients pg. 25-30; RO-C-AOP04, Abnormal Operating Procedures - Day 4 LESSON PLAN/OBJ: RO-C-TRANS4/#TRANS4A.3; RO-C-AOP04/#AOP4B KA - 002000 K6.07 Reactor Coolant System (RCS)

Knowledge of the effect of a loss or malfunction on the following RCS components:

Pumps RO/SRO Value - (2.5 / 2.8) CFR - 41.7 / 45.7 SOURCE: INPO # 23437 Indian Point 3 (Unit) - 3/10/2003 Original Quest. KA - 002.k6.02 Question #25 KA# - 002000 K6.07 Exam Level - RO Question Source - DIRECT-INPO - INDIANPOINT3-2003 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 32

26. 026 1 During a recent degassing operation of the RCS, Volume Control Tank (VCT) level was raised to 70% without any concurrent adjustment in VCT pressure as level rose.

This caused Reactor Coolant Pump (RCP) #1 seal leakoff flow to _____(1)_____, and will require _____(2)_____ to restore seal leakoff flows to normal.

_____(1)_____ _____(2)_____

A. rise opening up 1-QRV-200, Charging Header Pressure Control B. lower venting the VCT C. lower closing down 1-QRV-200, Charging Header Pressure Control D. rise venting the VCT Answer: B Raising VCT level without venting the VCT will cause Pressure to rise. This will place more backpressure on the RCS seals and cause seal leakoff flow to lower.

A - Incorrect - Seal leakoff flow will lower & the VCT must be vented. Plausible if the candidate thinks that the VCT pressure rise would cause higher charging pressure rising seal injection flow.

C - Incorrect - Seal leakoff flow will lower, but the VCT must be vented. Plausible if the candidate thinks that raising charging pressure to provide more seal injection flow will correct the low leakoff flow.

D - Incorrect - Seal leakoff flow will lower & the VCT must be vented. Plausible if the candidate thinks that the VCT pressure rise would cause higher charging pressure rising seal injection flow.

REFERENCE:

RO-C-201, Reactor Coolant Pumps, RO-C-AOP04, Abnormal Operating Procedures - Day 4 LESSON PLAN/OBJ: RO-C-AOP04/#AOP4.21 KA - 003000 A2.05 Reactor Coolant Pump System (RCPS)

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

Effects of VCT pressure on RCP seal leakoff flows RO/SRO Value - (2.5 / 2.8) CFR - 41.5 / 43.5 / 45.3 / 45.13 SOURCE: INPO # 21410 Braidwood 1 - 7/17/2002 Original Quest. KA - 003.a2.05 Question #26 KA# - 003000 A2.05 Exam Level - RO Question Source - DIRECT-INPO - BRAIDWOOD 2002 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 33

27. 027 2 A reactor coolant pump (RCP) is circulating reactor coolant at 100oF. After several hours the reactor coolant temperature has risen to 150oF.

Assuming coolant flow rate (gpm) is constant, RCP motor amps will have _______(1)_______

because ______(2)________.

A.

risen system head losses have lowered B.

lowered system head losses have risen C.

risen coolant density has risen D.

lowered coolant density has lowered Answer: D As RCS temperature raises the coolant density lowers. The pumps will perform less work to move the lighter coolant resulting in a reduction of current required.

A - Incorrect - Density lowers requiring less amps. System configuration does not change so head losses are consistent.

B - Incorrect - System configuration does not change so head losses are consistent.

C - Incorrect - Density lowers requiring less amps.

REFERENCE:

RO-C-GF17, Pumps (GP Components 2) pg. 40, RO-C-00201, Reactor Coolant Pump System pg. 27 LESSON PLAN/OBJ: RO-C-GF17/7b KA - 003000 A4.02 Reactor Coolant Pump System (RCPS)

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

RCP motor parameters RO/SRO Value - (2.9 / 2.9) CFR - 41.7 / 45.5 to 45.8 Original Question Source: Bank 01COMC0207-39 From GFES# 335;BANK: P2023 (B2020)

Original KA - 191004 K1.07 (2.9/2.9)

Exam/Quizzes: RO24GFES04 Question #27 KA# - 003000 A4.02 Exam Level - RO Question Source - DIRECT - 01COMC0207-39 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 34

28. 028 3 Unit 2 Power is 75%.

Which ONE of the following describes the effect on the plant if QRV-251 Charging Flow Control Valve fails closed (assume no operator action)?

Pressurizer level lowers until letdown isolates and PRZ heaters deenergize; pressurizer level...

A. continues to lower and pressure lowers until the Reactor trips on Low Pressurizer Pressure.

B. continues to lower and pressure lowers until the Reactor trips on (OTDT) Over Temperature Delta Temperature.

C. then rises and pressure rises until the Reactor trips on High Pressurizer Pressure.

D. then rises until the Reactor trips on High Pressurizer Level.

Answer: A When the QRV-251 fails closed all Charging line and Seal Injection flow is stopped.

Pressurizer level will lower until letdown isolates at 17%. Pressure Heaters will also trip off. Pressurizer level will continue to lower and pressure will lower (from no heaters and level loss) until the Reactor trips on Low pressure at 1875 psig.

B - Incorrect - Pressure does input to OTDT but the impact is small without a corresponding temperature change. The reactor will trip on Low pressure first.

C - Incorrect - Seal Injection is isolated so level will not rise and heaters remain de-energized. Plausible if this were a level channel failure & Reactor did not trip on High level.

D - Incorrect - Seal Injection is isolated so level will not rise and heaters remain de-energized. Plausible if this were a level channel failure as the Reactor would trip on High level.

REFERENCE:

SOD-00202-003 Pressurizer Level Control System and SOD-00300-001 Charging and Letdown System.

LESSON PLAN/OBJ: RO-C-0300/#8, RO-C-0202/#11 KA - 004000 K3.07 Chemical and Volume Control System (CVCS)

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

PZR level and pressure RO/SRO Value - (3.8 / 4.1) CFR - 41.7 / 45.6 Question #28 KA# - 004000 K3.07 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 35

29. 029 1 The following conditions exist:

- Unit is in Mode 4 during cooldown per OHP-4021-001-004, Plant Cooldown from Hot Standby to Cold Shutdown

- West RHR Pump is operating with the return aligned to the Alternate Cooldown Path to Loops 2 & 3.

- RCS temperature is 300°F and stable

- RCS pressure is 335 psig and stable The air supply line to IRV-320, West RHR Hx Outlet Valve, breaks, causing a complete loss of Instrument Air to the valve.

Which ONE (1) of the following describes the effect on the plant and the action that could be taken to mitigate the transient?

A. RHR Flow through the West HX will be lost. Throttle open IRV-311 RHR HX Bypass to maintain greater than 3000 gpm RHR flow.

B. RHR Flow through the West HX will be lost. Stop the West RHR pump immediately to prevent overpressurizing letdown.

C. RHR Flow through the West HX will rise. Throttle ICM-111, RHR Discharge to Cold Leg 2 &

3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.

D. RHR Flow through the West HX will rise. Throttle ICM -321, West RHR Injection to Loops 2

& 3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.

Answer: D IRV-320 fails open on loss of air. This will raise RHR flow through the HX. The ICM-321 can be throttled closed to reduce total RHR flow and IRV-311 can be throttled open to allow more flow to bypass the HX in order to control RCS cooldown.

A - Incorrect - IRV-320 fails open so flow will raise. Plausible since RHR flow is maintained > 3000 gpm to minimize vibrations & cavitation through the piping.

B - Incorrect - IRV-320 fails open so flow will raise. Plausible since RHR flow to the letdown system taps off before the IRV-320 and if IRV-320 closed it may raise letdown pressure.

C - Incorrect - The ICM-111, RHR Discharge to Cold Leg 2 & 3 is in the NORMAL Cooldown Path and should only be used if the Alternate cooldown path is unavailable.

This would be correct if the Normal Cooldown path was used. (02-OHP-4021-017-002 3.10)

REFERENCE:

02-OHP-4021-017-002, Placing in Service the RHR System; 02-OHP-4022-064-002, Loss of Control Air Recovery (Step 43 & Att. B-9)

LESSON PLAN/OBJ: RO-C-01700/#6 KA - 005000 A2.01 Residual Heat Removal System (RHRS)

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

Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation RO/SRO Value - (2.7 / 2.9) CFR - 41.5 / 43.5 / 45.3 / 45.13 Question #29 KA# - 005000 A2.01 Exam Level - RO

DC Cook 2006 NRC Exam 36 Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 37

30. 030 1 Which ONE of the following describes the freeze protection for the Unit 1 and Unit 2 Refueling Water Storage Tanks (RWST)?

A. Both Units use banks of heat tracing on the outside of the tank.

B. Both Units use a recirculation loop comprised of a pump and electric water heaters.

C. Unit 1 uses a recirculation loop comprised of a pump and electric water heaters, while Unit 2 uses banks of heat tracing on the outside of the tank.

D. Unit 1 uses banks of heat tracing on the outside of the tank, while Unit 2 uses a recirculation loop comprised of a pump and electric water heaters.

Answer: D Unit 1 uses banks of heat tracing on the outside of the tank, while Unit 2 uses a recirculation loop comprised of a pump and electric water heaters.

A - Incorrect - Unit 2 uses a recirculation loop comprised of a pump and electric water heaters.

B - Incorrect - Unit 1 uses banks of heat tracing on the outside of the tank C - Incorrect - This is the opposite of the actual configuration.

REFERENCE:

RO-C-00800, Emergency Core Cooling System pg. 35-36 LESSON PLAN/OBJ: RO-C-00800/#3,4 KA - 006000 K1.10 Emergency Core Cooling System (ECCS)

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

Safety injection tank heating system RO/SRO Value - (2.6 / 2.8) CFR - 41.2 to 41.9 / 45.7 to 45.8 SOURCE: INPO # 0 -

Original Quest. KA - 006000 k1.10 Question #30 KA# - 006000 K1.10 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 38

31. 031 21 Which ONE of the following data sets are within normal PRT limits for at power operations?

A. PRT Temperature - 100°F, Level - 90%, and Pressure - 12 psig B. PRT Temperature - 100°F, Level - 82%, and Pressure - 4 psig C. PRT Temperature - 65°F, Level - 82%, and Pressure - 1.0 psig D. PRT Temperature - 135°F, Level - 70%, and Pressure - 4 psig Answer: B PRT temperature is normally at Containment Temperature of ~100-110°F with level 80-84% and pressure of ~ 2-3 psig. The Alarm Limits are <126°F, Pressure 2.5 to 10 psig, and level 78.5% to 84.5%.

A - Incorrect - The tank level and pressure are too high.

C - Incorrect - The Temperature is too low for normal containment temperature.

Pressure is also too low.

D - Incorrect - The temperature is too high. The level is too low.

REFERENCE:

01-OHP-4021-002-006, Pressurizer Relief Tank Operation, Annunciator

1. 208 Drops 26, 31, and 36.

LESSON PLAN/OBJ: RO-C-AOP-1 / #19 Control Board Parameters are monitored on a ongoing basis 02-OHP-5030-001-003.

KA - 007000 2.1.1 Pressurizer Relief Tank/Quench Tank System (PRTS)

Conduct of Operations Knowledge of conduct of operations requirements.

RO/SRO Value - (3.7 / 3.8) CFR - 41.10 / 45.13 Question #31 KA# - 007000 2.1.1 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 39

32. 032 1 The following conditions exist:

- Unit 2 was in cold shutdown with the RCS drained to mid-loop.

- Filling and venting is in progress.

- Pressurizer level is 100%.

- A nitrogen blanket is on the PRT; PRT level is at 5%.

- The gaseous waste disposal system is aligned to support a bubble.

- The PZR heaters are energized.

Prior to drawing a bubble in the pressurizer which ONE of the following must be accomplished?

A. Pressurize the RCS to 325-350 psig B. Establishing 50% in the pressurizer C. Bumping the RCPs to remove entrapped gases D. Filling the PRT to 80-85%

Answer: D After the Pressurizer has been filled to >100% (level change seen in PRT), the PRT is re-aligned and the Level is raised to the normal operating band of 80-85%.

A - Incorrect - The RCS pressure is maintained around 50-60% while a bubble is drawn.

Plausible since pressure is raised to 325-350 psig after the bubble is drawn.

B - Incorrect - The pressurizer is maintained at 100% until the bubble is drawn. Plausible since level is reduced to 50% after the bubble is drawn.

C - Incorrect - The RCS is pressurized prior to bumping the RCPs. Plausible since the RCS is vented prior to drawing the bubble and the RCPs are bumped after the bubble is drawn.

Deleted "RCS is aligned to vent..." from stem 4th line.

REFERENCE:

02-OHP-4021-002-001 Filling and Venting the RCS (Steps 4.60-4.63)

LESSON PLAN/OBJ: RO-C-NOP04/#25 KA - 007000 K5.02 Pressurizer Relief Tank/Quench Tank System (PRTS)

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

Method of forming a steam bubble in the PZR RO/SRO Value - (3.1 / 3.4) CFR - 41.5 / 45.7 SOURCE: INPO # 19506 Cook 1 - 5/21/2001 Original Quest. KA - 007.k5.02 Question #32 KA# - 007000 K5.02 Exam Level - RO Question Source - DIRECT - COOK 2001 Q#51 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 40

33. 033 1 The following conditions exist:

- U2 is at 100% Power with normal operating temperature and pressure

- The East CCW pump is running

- The water level in the CCW surge tank is 58" and slowly lowering

- The makeup valves to the CCW surge tank are fully open Which one of the following actions is required by 02-OHP-4022-016-001, Malfunction of the CCW System?

A. Start the West CCW pump, split the East and West Headers, when surge tank level is <

48 identify and isolate the leaking header.

B. Start the West CCW pump, split the East and West Headers, and shutdown the East CCW pump.

C. Isolate letdown(or excess letdown if in service), place standby CCW switch in Lock-Out and commence a rapid plant shutdown.

D. If level cannot be maintained > 48", trip the Rx, trip RCPs and go to E-0.

Answer: A The CCW Surge tank has a baffle that extends to 48". The procedure has the operator split the CCW headers and then monitor the CCW surge tank level indicators after surge tank level has lowered to <48" (height of baffle). Two surge tank level indicators monitor both sides of the baffle. When it is identified which side is leaking that header is isolated.

B - Incorrect - The surge tank levels are monitored after level lowers to <48" to determine which train is leaking. The east pump is not stopped until it is determined which header is leaking. Additionally, the Miscellaneous header is initially aligned to the East CCW header. Plausible since the East Pump was initially running.

C - Incorrect - A plant Shutdown is not required with a slow level reduction. Letdown would be an inleakage source. Plausible since placing the standby pump in Lock-Out would be conservative to protect the pump.

D - Incorrect - The surge tank baffle only extends to 48" so level must be allowed to lower to <48" to identify the leaky header. Plausible since these actions are required if CCW pumps must be stopped due to cavitation.

REFERENCE:

02-OHP-4022-016-001, Malfunction of the CCW System Steps 3, 5-11 LESSON PLAN/OBJ: RO-C-AOP11/#11.19 KA - 008000 A1.04 Component Cooling Water System (CCWS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including:

Surge tank level RO/SRO Value - (3.1 / 3.2) CFR - 41.5 / 45.5 Question #33 KA# - 008000 A1.04 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/4

DC Cook 2006 NRC Exam 41

34. 034 2 Which ONE of the following correctly describes the conditions that allow implementation of 02-OHP-4023-ES-0.0, Re-Diagnosis?

A. Entry is based on operator judgement at any time that a Function Restoration (FR) procedure is NOT in effect.

B. Entry is based on operator judgement anytime after 02-OHP-4023-E-0, Reactor Trip/Safety Injection completion provided that SI is actuated and a Function Restoration (FR) procedure is NOT in effect.

C. Anytime after the EOP's have been entered.

D. Anytime that SI is actuated and a Function Restoration (FR) procedure is in effect.

Answer: B An SI must be in service, and E-0 has been completed.

A. Incorrect, ES-0.0 entry requires an SI to be in service and E-0 to be completed.

C. Incorrect, E-0 must be completed.

D. Incorrect, an SI must be present, and E-0 completed.

Added FR NOT in Effect to A & B.

REFERENCE:

OHI-4023, ABNORMAL/EMERGENCY PROCEDURE USER'S GUIDE, Section 4 LESSON PLAN/OBJ: RO-C-EOP01/#24 KA - 00WE01 EA2.2 Rediagnosis Ability to determine and interpret the following as they apply to the Reactor Trip or Safety Injection/Rediagnosis:

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments RO/SRO Value - (3.3 / 3.9) CFR - 43.5 / 45.13 SOURCE: INPO # 25785 Surry 1 - 3/14/2003 Original Quest. KA - WE01GW.45.1 Question #34 KA# - 00WE01 EA2.2 Exam Level - RO Question Source - DIRECT-INPO - SURRY2003-25785 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 42

35. 035 6 What indication is used in OHP-4023-ECA-1.2, LOCA Outside Containment, to verify that the actions taken in the procedure have isolated the break?

A. RCS pressure rising B. Containment Sump level rising C. Auxiliary Building radiation lowering D. RHR pump discharge pressure rising Answer: A OHP-4023-ECA-1.2, LOCA Outside Containment verifies that likely leakage paths are isolated and then sequentially isolates ECCS injection paths looking for a rising RCS pressure to verify the leak is isolated.

B - Incorrect - The procedure checks for rising RCS pressure. The Sump level could rise even if the leak outside containment was not isolated.

C - Incorrect - The procedure checks for rising RCS pressure. Plausible since the procedure is entered based on rising/elevated Auxiliary Building radiation.

D - Incorrect - The procedure checks for rising RCS pressure. Based on the leak size the RHR pumps may be at shutoff head and therefore would not rise even if the leak was isolated.

REFERENCE:

OHP-4023-ECA-1.2, LOCA OUTSIDE CONTAINMENT LESSON PLAN/OBJ: RO-C-EOP09/#34 KA - 00WE04 EA2.2 LOCA Outside Containment Ability to determine and interpret the following as they apply to the LOCA Outside Containment:

Adherance to appropriate procedures and operations within the limitations in the facilities license and amendments RO/SRO Value - (3.6 / 4.2) CFR - 43.5 / 45.13 Original Source Question Prairie Island 8/2002 Original KA WE 04 EA1.3 (3.8 /4.0)

Question #35 KA# - 00WE04 EA2.2 Exam Level - RO Question Source - DIRECT - PRAIRIE ISLAND 2002 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 43

36. 036 5 Response to Loss of Secondary Heat Sink, 01-OHP-4023-FR-H.1, attempts to restore the secondary heat sink capability before initiating a bleed and feed cooling.

If Unit 1 AF pump flow can NOT be restored, which ONE of the following is the correct chronological order of preference for restoring SG feedwater? (In accordance with 01-OHP-4023-FR-H.1,from Highest to Lowest)

A. Main Feedwater flow Condensate flow Cross-tied Auxiliary Feedwater flow (from unaffected unit)

B. Main Feedwater flow Cross-tied Auxiliary Feedwater flow (from unaffected unit)

Condensate flow C. Cross-tied Auxiliary Feedwater flow (from unaffected unit)

Main Feedwater flow Condensate flow D. Cross-tied Auxiliary Feedwater flow (from unaffected unit)

Condensate flow Main Feedwater flow Answer: C 01-OHP-4023-FR-H.1 Step 4 has the operator attempt to restore AFW first. The RNO directs the operator to then use the Opposite Unit AFW. Steps 6-10 aligns Main FW and Steps 11-15 depressurizes the SGs and aligns the condensate System.

A - Incorrect - Crosstied AFW is used before Condensate and Main FW flow.

B - Incorrect - Crosstied AFW is used before Condensate and Main FW flow.

D - Incorrect - Main FW Flow is tried before Condensate (Plausible since Condensate is aligned prior to Main FW)

Changed Stem to Include Unit 1 AFW. Changed Distractors to List sources rather than number.

REFERENCE:

RO-C-EOP11 pg. 45, 01-OHP-4023-FR-H.1,Response to Loss of Secondary Heat Sink Steps 4-15 LESSON PLAN/OBJ: RO-C-EOP11/#10 KA - 00WE05 EK2.2 Loss of Secondary Heat Sink Knowledge of the interrelations between the Loss of Secondary Heat Sink and the following:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility RO/SRO Value - (3.9 / 4.2) CFR - 41.7 / 45.7 Original Source Question Bank 01EOPC1110-6 Original KA 2.4.23 (2.8/3.8)

Exam/Quizzes: RO20ECOMP; RO22EOP7A; RO22EOP7C; RQ2802V; RQ2802C Similar to INPO # 26856 Kewaunee, Unit 1 - 2/2/2004 and INPO # 27275 Ginna 1 -

4/27/2004

DC Cook 2006 NRC Exam 44 Question #36 KA# - 00WE05 EK2.2 Exam Level - RO Question Source - DIRECT - 01EOPC1110-6 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 45

37. 037 3 In OHP-4023-ECA-2-1, Uncontrolled Depressurization Of All Steam Generators, a foldout page item listing the OHP-4023-E-2, Faulted Steam Generator Isolation transition criteria is found.

This foldout page item is continuously monitored while performing the actions in OHP-4023-ECA-2.1 because...

A. if a faulted SG completely depressurizes, the operator is directed back to OHP-4023-E-2 for prompt identification and isolation.

B. if any SG level rises in an uncontrolled manner, the operator is directed to OHP-4023-ECA-3.1 for furthur recovery actions.

C. if complete depressurization of all SGs occurs, the operator is required to make a prompt transition to OHP-4023-FR-H.1 to maintain an adequate Heat Sink.

D. if restoration of any SG pressure boundary occurs, the operator is directed back to OHP-4023-E-2 for further optimal recovery actions.

Answer: D The Foldout page has transition criteria based on any SG pressure rising. When this condition exists, the operator is transitioned back to OHP-4023-E-2, since 1 SG is now isolated.

A - Incorrect - While OHP-4023-E-2, is titled 'Faulted Steam Generator Isolation' attempts at isolation have failed so OHP-4023-ECA is used to stabilize the plant.

B - Incorrect - If any intact SG level rises in an uncontrolled manner, the operator is directed to OHP-4023-E-3, SGTR.

C - Incorrect - Complete depressurization is expected during OHP-4023-ECA-2.1. AFW flow is maintained to keep the tubes wetted and may be raised to stabilize temperatures after the SGs are depressurized.

REFERENCE:

02-OHP 4023 ECA-2.1 LESSON PLAN/OBJ: RO-C-EOP07/#12 KA - 00WE12 EA1.3 Uncontrolled Depressurization of all Steam Generators Ability to operate and/or monitor the following as they apply to the Uncontrolled Depressurization of all Steam Generators:

Desired operating results during abnormal and emergency situations RO/SRO Value - (3.4 / 3.9) CFR - 41.7 / 45.5 / 45.6 Original Question Source: Bank 01EOPC0712-2 Original KA E12EK3.3 3.5/3.7 EXAM/QUIZZES: RO20EOP4; RO21EOP3; RO22EOP3B; RQ2706VA; RQ2804A; RO23EOP3 Question #37 KA# - 00WE12 EA1.3 Exam Level - RO Question Source - DIRECT - 01EOPC0712-2 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 46

38. 038 11 Unit 1 was operating at 100% when a pipe break occurred on the SG 14 steam header.

The following conditions exist :

- Main Steam Line Isolation has occurred.

- PRZ level dropped to 0% and has restored to 20%.

- RCS pressure is 1900 psig.

- Safety Injection (SI) has NOT been reset.

Which ONE of the following are the REQUIRED actions to re-energize the Pressurizer back-up heaters?

A. Reset SI Place the back-up heater control switches in the TRIP position.

Place the back-up heater control switches in the CLOSE position.

B. Reset SI Place the back-up heater control switches in the PULL-TO-RESET position.

Place the back-up heater control switches in the CLOSE position.

C. Place the back-up heater control switches in the PULL-TO-RESET position.

Place the back-up heater control switches in the CLOSE position.

D. Place the back-up heater control switches in the TRIP position.

Place the back-up heater control switches in the CLOSE position.

Answer: D The Backup heaters will trip off on the low pressurizer level. After level has been restored the control switches must be placed in the trip condition to reset the breaker.

Then the switches are placed to close to energize the heaters.

A - Incorrect - SI reset is NOT Required.

B - Incorrect - SI reset is NOT Required. Switch must be taken to TRIP position to reclose.

C - Incorrect - The Heaters must be placed to Trip to reset the seal-in.

(From Original Question - Reset the master Pressure Controller - reason Plausible since many controllers have to be reset after the parameter has exceeded limits.)

REFERENCE:

RO-C-00202, Pressurizer and Pressure Relief System pg. 16-24 LESSON PLAN/OBJ: RO-C-00202/#16 KA - 010000 A4.02 Pressurizer Pressure Control System (PZR PCS)

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

PZR heaters RO/SRO Value - (3.6 / 3.4) CFR - 41.7 / 45.5 to 45.8 SOURCE: AUD02-BOTH47, RO23 Audit RO#008 /SRO#008 Original Quest. KA - SYS 010 G2.1.31 (4.2/3.9)

Question #38 KA# - 010000 A4.02 Exam Level - RO Question Source - DIRECT - AUDIT RO23 Q#8 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 47

39. 039 1 The following conditions exist:

-Unit 1 is at 100% power

-Pressurizer PORV NRV-151 opens and sticks open.

-The associated PORV block valve cannot be closed

-PRT pressure rises to the point that the PRT Rupture Disc ruptures What is the effect of the disc rupturing?

A. Makeup to the PRT initiates.

B. PRZ PORV outlet temperature lowers.

C. Containment N2 header pressure lowers.

D. PRT level drains below the sparging nozzles.

Answer: B When the PRT rupture disc ruptures, The Pressurizer through the PORV & PRT will discharge to containment. This will lower the pressure to which the PORV discharges which will lower the temperature at the outlet.

A - Incorrect - While level may lower (water flash to steam) makeup to the PRT is not automatic.

C - Incorrect - While a N2 cover is supplied to the PRT, The regulator will limit the flow out of the N2 to the PRT.

D - Incorrect - Incorrect - While level may lower (water flash to steam), it will not lower below the sparger @ 5%.

REFERENCE:

RO-C-00202, Pressurizer and Pressure Relief System LESSON PLAN/OBJ: RO-C-00202/#8 KA - 010000 K1.07 Pressurizer Pressure Control System (PZR PCS)

Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems:

Containment RO/SRO Value - (2.9 / 3.1) CFR - 41.2 to 41.9 / 45.7 to 45.8 SOURCE: INPO # 27569 Prairie Island 1 - 4/23/2004 Original Quest. KA - 010 K6.04 Question #39 KA# - 010000 K1.07 Exam Level - RO Question Source - DIRECT-INPO - PRAIRIE ISLAND 2004 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 48

40. 040 1 The following conditions exist:

- Charging, letdown, and PRZ level control system are in automatic. QRV-160 and 2-QRV-161, Letdown Orifice Valves are Open

- Letdown Hx Outlet Flow QFI -301 118 gpm

- Charging Header Flow QFI -200 130 gpm

- Total seal flow to RCPs QFI -210 to 240 32 gpm The controlling PRZ level channel fails high to an indicated 100% level.

Which of the following describes the short term effect on total RCP seal injection flow, assuming NO operator action?

Total seal injection flow...

A. raises to about 50 gpm B. remains about 32 gpm C. lowers to about 12 gpm D. lowers to 0 gpm Answer: C With the Pressurizer level channel failed high the charging flow control valve QRV-251 will close to the minimum flow position of ~ 47 gpm. This will cause seal injection flow to lower accordingly (47/130 =.036 x 32 = 11.6).

A - Incorrect - The charging flowpath is not isolated (QRV-200 is still open). This is the minimum charging flow value.

B - Incorrect - The charging flowpath is not isolated (QRV-200 is still open) so flow to the seals will lower as total charging flow lowers.

D - Incorrect - The minimum flow setting on QRV-251 controller will keep some flow going to the seals.

REFERENCE:

SOD-00202-003 Pressurizer Level Control System and SOD-00300-001 Charging and Letdown System.

LESSON PLAN/OBJ: RO-C-0202/#5 KA - 011000 K3.02 Pressurizer Level Control System (PZR LCS)

Knowledge of the effect that a loss or malfunction of the PZR LCS will have on the following:

RCS RO/SRO Value - (3.5 / 3.7) CFR - 41.7 / 45.6 SOURCE: INPO # 26398 Byron 1 - 12/10/2003 Original Quest. KA - 011000.K3.02 Question #40 KA# - 011000 K3.02 Exam Level - RO Question Source - DIRECT-INPO - BYRON 2003 - 26398 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 49

41. 041 5 The following plant conditions exist:

- A Unit 1 startup is in progress.

- The reactor is critical in the source range.

- N41 Power Range channel is removed from service for zero power physics testing.

- A loss of power to the CRID 2 bus occurs.

Which ONE of the following actions will occur?

A. Reactor trips and N32 Source Range channel is de-energized.

N31 Source Range channel is still in operation.

B. The reactor is critical and BOTH source range channels are de-energized.

C. The reactor is critical and N32 Source Range channel is de-energized.

N31 Source Range channel is still in operation.

D. Reactor trips and BOTH source range channels are de-energized.

Answer: D A loss of CRID 2 causes a loss of power to N42. This loss also causes a loss of power to RPS channel 2. This will cause a trip condition for Power range trips for channel 2.

Since N41 is already removed from service its bistable are in the tripped condition. This meets the 2/4 logic to cause a reactor trip. Additionally the signal for 2/4 power range channels above P-10 will cause the SR channels to deenergize.

A - Incorrect - P-10 will be met, both SR's will de energize.

B - Incorrect - Reactor trips on a number of PR/SR trip setpoints.

C - Incorrect - Reactor trips on a number of PR/SR trip setpoints. Also, P-10 will turn off both SR's.

REFERENCE:

RO-C-01101, Solid State Protection System LESSON PLAN/OBJ: RO-C-01101/#6 KA - 012000 K2.01 Reactor Protection System Knowledge of bus power supplies to the following:

RPS channels, components, and interconnections RO/SRO Value - (3.3 / 3.7) CFR - 41.7 Original Source Question AUDIT RO22-BOTH-23 Q#20 Original KA SYS 015K4.01 Question #41 KA# - 012000 K2.01 Exam Level - RO Question Source - DIRECT - AUDIT RO22-BOTH-23 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 50

42. 042 10 The following conditions exist:

- Containment pressure instrument Channel #1, 2-PPP-303 (PT-937) declared inoperable.

- Required actions per 02-OHP-4022-013-011 Containment Instrumentation Malfunction have been completed.

- Required Technical Specification Actions have been taken for Channel #1, 2-PPP-303 (PT-937)

Which ONE of the following describes the REMAINING coincidence for the SAFETY INJECTION ACTUATION and CTS ACTUATION?

Remaining Channels to cause actuation Remaining Channels with INPUT to this function SAFETY INJECTION CTS ACTUATION ACTUATION A.

2/3 2/3 B.

1/3 2/3 C.

1/2 1/3 D.

2/3 1/3 Answer: A The CTS Actuation Bistable is placed in the BYPASSED condition to prevent inadvertent actuation. This changes the remaining channel coincidence to 2/3 instead of the previous 2/4. Only 3 channels (Channels 2, 3, & 4) feed the SI Actuation This channel does NOT feed SI so the SI coincidence remains at 2/3.

B - Incorrect - This channel does NOT feed SI. (True if candidate assumes 4 channels feed SI)

C - Incorrect - This channel does NOT feed SI Actuation. CTS is bypassed. (True if candidate assumes this channel does feed SI and that CTS is tripped)

D - Incorrect - The CTS is placed in BYPASS. (True if candidate knows this channel does NOT feed SI but assumes CTS is tripped)

Modified Stem to Fail Channel #1, 2-PPP-303 (PT-937) instead of Channel 3. Changed answer to A. Also changed distractors B, C, & D (Note that channels feed different actuations)

REFERENCE:

RO-C-AOP-2, Abnormal Operating Procedures - Day 2 pg. 13 and 14 02-OHP-4022-013-011 Containment Instrumentation Malfunction pg. 2 and Att. D LESSON PLAN/OBJ: RO-C-01100/#9 KA - 013000 K6.01 Engineered Safety Features Actuation System (ESFAS)

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

Sensors and detectors RO/SRO Value - (2.7 / 3.1) CFR - 41.7 / 45.7

DC Cook 2006 NRC Exam 51 SOURCE: INPO # 27671 Cook 1 - 4/29/2004(RO#037/SRO#037)

From 1021-CALLAWAY97 Original Quest. KA - 013000 K6.01 Engineered Safety Features Actuation System (ESFAS)

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

Sensors and detectors RO-2.7 SRO-3.1 Question #42 KA# - 013000 K6.01 Exam Level - RO Question Source - MODIFIED - COOK 2004 #37 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 52

43. 043 3 Fire in the Unit #1 Control Room Cable Vault has resulted in loss of equipment control and normal habitability. The Control Room has been evacuated.

Which ONE of the following would be used to monitor Reactor Power from outside the control room?

A. Source Range - N21 B. Source Range - N23 C. Wide Range - N21 D. Wide Range - N23 Answer: B Source Range N-23 may be monitored from OHP-4025-LS-1.

A - Incorrect - This gamametrics monitor only indicates on the NIS I panel.

C - Incorrect - This gamametrics monitor only indicates on the NIS I panel and the Rod and Flux panel.

D - Incorrect - This gamametrics monitor only indicates on the NIS III panel and the Rod and Flux panel.

REFERENCE:

RO-C-01300, Excore Nuclear Instrumentation System pg. 29 LESSON PLAN/OBJ: RO-C-01300/#5, 17 KA - 015000 K4.03 Nuclear Instrumentation System Knowledge of NIS design feature(s) and/or interlock(s) which provide for the following:

Reading of source range/intermediate range/power range outside control room RO/SRO Value - (3.9 / 4.0) CFR - 41.7 Question #43 KA# - 015000 K4.03 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 53

44. 044 11 OHP-4021-028-001 Containment Ventilation contains a caution concerning the Loss of Lower Containment Ventilation.

Which ONE of the following describes the concern and the basis for this concern?

The Loss of Lower Containment Ventilation may...

A. cause NESW Containment Isolation Valves to become pressure bound due to temperature changes.

B. cause Phase B isolation signal in about 15 minutes due to Reactor Coolant Pump motors heating the Containment atmosphere.

C. cause the Ice Condenser Doors to open due to pressure differential if the Upper Containment Ventilation units are not also stopped.

D. cause Ice Condenser temperatures to raise to unacceptable levels if additional Coolers are not started.

Answer: B The RCPs continue to provide heat input to the containment atmosphere during a loss of containment cooling. The heat input would cause a rapid rise in containment pressure, resulting in an SI and CTS actuation based solely on a loss of containment cooling.

A - Incorrect - This concern is based on isolation of a segment of the NESW system.

C - Incorrect - The ice condenser door differential pressure caution is contained in the Containment Purge procedure and requires Upper Containment pressure to be less than lower Containment.

D - Incorrect - Ice condenser cooling is provided by the Glycol cooling system which uses NESW to Cool its chillers. A loss of lower containment ventilation should not significantly impact this cooling.

REFERENCE:

OHP-4021-028-001 Containment Ventilation Step 3.5, RO-C-02800, Containment Ventilation System LESSON PLAN/OBJ: RO-C-02800/#6 KA - 022000 2.1.32 Containment Cooling System (CCS)

Conduct of Operations Ability to explain and apply all system limits and precautions.

RO/SRO Value - (3.4 / 3.8) CFR - 41.10 / 43.2 / 45.12 Question #44 KA# - 022000 2.1.32 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 54

45. 045 2 Containment temperature has risen from 100oF to 160oF due to a containment cooling malfunction. If the plant is stable at 100% power and there are negligible RCS or containment pressure changes, which one of the following describes the effect of the increase in containment temperature on the pressurizer level indicated by the pressurizer level control channels?

A. Indicated level will be HIGHER than actual level because the reference leg fluid density lowers.

B. Indicated level will be LOWER than actual level because of the elevated containment temperature causes flashing in the reference leg.

C. Indicated level will be HIGHER than actual level because the elevated containment temperature causes flashing in the reference leg.

D. Indicated level will be LOWER than actual level because the reference leg fluid density lowers.

Answer: A Pressurizer Level uses a wet reference leg DP level indicator. This compares the pressure of the full reference leg with the pressure of the actual water in the pressurizer.

When these are equal the level indicates 100%. As the temperature in Containment and therefore the reference leg raises the density & weight of the reference leg lowers. This means that the level in the pressurizer will indicate higher for the same initial actual level.

B - Incorrect - The reference leg will not flash since it is exposed to the high pressure of the pressurizer.

C - Incorrect - The reference leg will not flash since it is exposed to the high pressure of the pressurizer.

D - Incorrect - The density does lower but this causes a higher indication.

REFERENCE:

RO-C-GF27, Sensors and Detectors pg. 51 & 52 LESSON PLAN/OBJ: RO-C-GF27/9d KA - 022000 K3.02 Containment Cooling System (CCS)

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

Containment instrumentation readings RO/SRO Value - (3.0 / 3.3) CFR - 41.7 / 45.6 SOURCE: INPO # 26772 Kewaunee, Unit 1 - 2/2/2004 Original Quest. KA - 022000K302 Question #45 KA# - 022000 K3.02 Exam Level - RO Question Source - DIRECT-INPO - KEWNEE 2004 - 26772 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 55

46. 046 1 Unit 2 was operating at 95% power, when an operator identified that one of the Ice Condenser Intermediate Deck Doors was slightly open and incapable of being fully closed due to Ice Buildup.

Which ONE of the following actions, if any, are required to continue operations at power?(Technical Specifications 3.6.11 and 3.6.12 provided)

A. Monitor the ice bed temperature less than or equal to 27oF every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for a maximum of 2 days.

B. Monitor the ice bed temperature less than or equal to 27oF every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for a maximum of 14 days.

C. No action is required since the Ice Condenser Inlet Doors are all OPERABLE.

D. Restore the Door to Operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Answer: B If the Ice Condenser Intermediate deck doors are inoperable then ITS 3.6.12 Action B applies. This requires temperature to be monitored every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for up to 14 days.

A - Incorrect - The ice Condenser temperature is monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per ITS 3.6.11 and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> are allowed per ITS 3.6.11 if the ice condenser is Inoperable.

C - Incorrect - The Intermediate deck doors are still required to be operable.

D - Incorrect - This is the requirement if a INLET door can not be fully opened.

REFERENCE:

ITS 3.6.12 pg. 3.6.12-1, 2,&3 OHP-4030-STP-030 Rev 35 LESSON PLAN/OBJ: RO-C-01000/#12 Attachment Provided - Technical Specifications 3.6.11 and 3.6.12 KA - 025000 K6.01 Ice Condenser System Knowledge of the effect of a loss or malfunction of the following will have on the Ice Condenser System:

Upper and lower doors of the ice condenser RO/SRO Value - (3.4 / 3.6) CFR - 41.7 / 45.7 Question #46 KA# - 025000 K6.01 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 56

47. 047 11 Following a LOCA inside containment in which pressure reached 3.4 psig, the operator notes the following indications:

- Containment pressure is 2.8 psig and lowering slowly;

- Refueling Water Storage Tank (RWST) level is 23 percent and lowering slowly.

- Panel 105 Drop 2 - "SPRAY ADDITIVE TANK LEVEL AT 4000 Gal" alarm has just actuated.

Assuming NO operator actions, the operator would observe:

- IMO 215/225 CTS Pump Suction Valves __(1)__,

- IMO-212/222 Eductor Supply Valves __(2)__,

- and the CTS pumps __(3)__.

(1) Pump Suction (2) Eductor (3) CTS IMO 215/225 IMO-212/222 Pumps A.

Closed Closed Tripped B.

Open Open Running C.

Closed Open Tripped D.

Open Closed Running Answer: B Containment Spray will actuate at 2.9 psig. This will start the pumps and open the Spray additive tank valves. The eductor and suction valves are normally open. The Spray additive tank alarm at 4000 gal is an expected alarm after the actuation and shows that NaOH is being injected. If the tank reached the Lo-2 alarm the spray additive tank valves would have isolated (Valves close 1" above lo-2). The RWST level of <25% is below the switchover setpoint but the actions to close the suction to the RWST must be performed manually.

A - Incorrect - Pump suction and inductor will not close. Pumps should still be running.

C - Incorrect - Pump suction will not close. Spray Additive tank would still be open.

Pumps should still be running.

D - Incorrect - Spray Additive tank and eductor would still be open.

REFERENCE:

1-OHP-4024-105, Annunciator #105 Response: Containment Spray, Drop 3 and 23; SOD -00901-001, Containment Spray and Hydrogen Recombiners, RO-C-00900 pg 17.

LESSON PLAN/OBJ: RO-C-00900/#11 KA - 026000 K4.04 Containment Spray System (CSS)

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

Reduction of temperature and pressure in containment after a LOCA by condensing steam, to reduce radiological hazard, and protect equipment from corrosion damage (spray)

RO/SRO Value - (3.7 / 4.1) CFR - 41.7 SOURCE: AUDIT RO22 Both 80 (067/72)

Original Quest. KA - SYS 026 A3.01 Containment Spray System (CSS)

- Ability to monitor automatic operation of the CSS, including: Pump starts and correct

DC Cook 2006 NRC Exam 57 MOV positioning Question #47 KA# - 026000 K4.04 Exam Level - RO Question Source - DIRECT - AUDIT RO22-BOTH-80 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 58

48. 048 10 Unit 1 has experienced a LOCA and Loss of Offsite power.

The following conditions exist:

- Emergency Diesel Generator 1AB failed to start.

- Emergency Diesel Generator 1CD has started and loaded as designed.

- Power has been restored to the Reserve Aux Transformers.

- No buses have been energized from the RATs.

The Unit Supervisor directs you to verify or restore power so a hydrogen recombiner may be run.

Which ONE of the following actions is required to enable the associated Hydrogen Recombiner to be operated?

A. Verify that bus T11C has energized 600V Bus 11C and MCC-1-EZC-C for (Train A)

Hydrogen Recombiner Number 2.

B. Verify that bus T11C has energized 600V Bus 11C and Close the 11AC crosstie to supply power to Bus 11A and MCC-1-EZC-A for (Train B) Hydrogen Recombiner Number 1.

C. Energize RCP Bus 1B from the RAT to supply power to 600V Bus 11BMC for (Train B)

Hydrogen Recombiner Number 1.

D. Energize RCP Bus 1C from the RAT to supply power to 600V Bus 11CMC for (Train A)

Hydrogen Recombiner Number 2.

Answer: A Hydrogen Recombiner #1 is powered form MCC-1-EZC-B and Hydrogen Recombiner

1. 2 is powered form MCC-1-EZC-C.

B - Incorrect - Hydrogen Recombiner #1 is powered form MCC-1-EZC-B C - Incorrect - Hydrogen Recombiner #1 is powered form MCC-1-EZC-B D - Incorrect - Hydrogen Recombiner #2 is powered form MCC-1-EZC-C

REFERENCE:

SD-00900 Containment Spray and Hydrogen Recombiner System Description pg. 32 LESSON PLAN/OBJ: RO-C-00900/#9 KA - 028000 K2.01 Hydrogen Recombiner and Purge Control System (HRPS)

Knowledge of bus power supplies to the following:

Hydrogen recombiners RO/SRO Value - (2.5 / 2.8) CFR - 41.7 SOURCE: INPO # 27678 Cook 1 - 4/29/2004(RO#044/SRO#044)

Original Quest. KA - 028000 K2.01 Question #48 KA# - 028000 K2.01 Exam Level - RO Question Source - DIRECT - COOK 2004 #44 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 59

49. 049 31 The following conditions exist:

- Unit 1 was in Mode 1.

- A Containment Pressure Relief was in progress with the following lineup:

1-VCR-107, Cntmt Press Relief Valve IC - OPEN 1-VCR-207, Cntmt Press Relief Valve OC - OPEN 1-HV-CPR-1, CNTMT Press Relief Fan - RUNNING

- An external failure alarm on VRS-1201, Upper Containment Normal Range Monitor, occurs.

Which ONE of the following describes the required operator response for the Containment Pressure Relief System due to the failure alarm?

A. Verify 1-VCR-207 has automatically closed, 1-HV-CPR-1 has automatically tripped, and manually close 1-VCR-107.

B. Verify 1-VCR-107 has automatically closed, manually trip 1-HV-CPR-1, and manually close 1-VCR-207.

C. Verify 1-VCR-207 and 1-VCR-107 have automatically closed and 1-HV-CPR-1 has automatically tripped.

D. Initiate an eSAT to address the VRS-1201 failure. The release may be continued as long as VRS-1101 is operable.

Answer: A VRS-1201 closes the Outside Containment Isolation Valve VCR-201 and trips the 1-HV-CPR-1, CNTMT Press Relief Fan. VCR-107 is closed by the VRS-1101 channel.

B - Incorrect - Actions for failure of the VRS-1101 channel.

C - Incorrect - VCR-107 is closed by the VRS-1101 channel. This would be the response if both channels sensed a high radiation condition.

D - Incorrect - The release will automatically terminate (VCR-207 & HV-CPR-1). The release may be restarted after appropriate actions have been taken.

REFERENCE:

SOD-1350-001, Radiation Monitoring System; 12-OHP-4021-139 #2 LESSON PLAN/OBJ: RO-C-02800 / #14 KA - 029000 2.4.31 Containment Purge System (CPS)

Emergency Procedures/Plan Knowledge of annunciators alarms and indications, and use of the response instructions.

RO/SRO Value - (3.3 / 3.4) CFR - 41.10 / 45.3 Question #49 KA# - 029000 2.4.31 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 60

50. 050 11 The following conditions exist:

- Unit 1 is at 100% power and stable.

- Steam Generator Level Controls are in AUTOMATIC.

- Steam Generator #12 Steam Flow Channel 1, 1-MFC-121, is selected to the Steam Generator Level Control System.

A blown fuse causes 1-MFC-121 to fail offscale low.

Which ONE of the following describes the expected plant response?

(Assuming no operator action)

The Steam Generator Level Control system will...

A. initially lower feed flow and then slowly return #12 SG level to approximately program level.

B. automatically transfer the # 12 FW Regulating Valve Controller to Manual to maintain the current valve position.

C. initially raise feed flow and then slowly return #12 SG level to approximately program level.

D. lower feed flow to #12 SG to 0 pph, resulting in a Reactor Trip.

Answer: B and D Depending on the fuse location, a sudden failure offscale will either: 1) cause the Taylor Controller to shift to manual at the current position, or 2) lower feed flow to #12 SG to 0 pph, resulting in a Reactor Trip.

A - Incorrect - The FRV Controller will shift to manual. This is the response to a slow Steam Flow failure.

C - Incorrect - The FRV Controller will shift to manual. This is the response to a slow FW Flow failure.

REFERENCE:

SD-05100, Steam Generator System Description pg. 20-21, RO-L-ES07, Control Room Controllers and Recorders LESSON PLAN/OBJ: RO-C-05100 / #9 KA - 035000 A4.01 Steam Generator System (S/GS)

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

Shift of S/G controls between manual and automatic control, by bumpless transfer RO/SRO Value - (3.7 / 3.6) CFR - 41.7 / 45.5 to 45.8 Question #50 KA# - 035000 A4.01 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 61

51. 051 3 Which ONE (1) of the following describes the location and use of MRA-1600/1700 Steam Generator PORV Monitors?

A. Positioned downstream of SG PORVs. Reliable under non-accident conditions ONLY.

B. Positioned upstream of SG PORVs. Reliable under non-accident conditions ONLY.

C. Positioned upstream of SG PORVs. Reliable under both accident and non-accident conditions.

D. Positioned downstream of the SG PORVs. Reliable under both accident and non-accident conditions.

Answer: D The SG PORVs are Positioned downstream for the SG PORVs and are Reliable under both accident and non-accident conditions.

A - Incorrect. Monitors are Reliable under both accident and non-accident conditions.

B - Incorrect. Positioned downstream of PORVs, Monitors are Reliable under both accident and non-accident conditions.

C - Incorrect. Positioned downstream of PORVs.

REFERENCE:

RO-C-01350 slide 51, UFSAR Table 7.8-1 pg. 22 of 26 LESSON PLAN/OBJ: RO-C-05103/#4 KA - 039000 K1.09 Main and Reheat Steam System (MRSS)

Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems:

RMS RO/SRO Value - (2.7 / 2.7) CFR - 41.2 to 41.9 / 45.7 to 45.8 SOURCE: INPO # 28083 Robinson 2 - 9/27/2004 Original Quest. KA - 039 K1.09 1 Question #51 KA# - 039000 K1.09 Exam Level - RO Question Source - DIRECT-INPO - ROBINSON2004-28083 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 62

52. 052 2 The following plant conditions exist on Unit 1:

- The reactor is critical at 7% power just prior to rolling the turbine.

- Manual FW control is maintaining SG Levels.

- RCS temperature is being controlled by steam dumps in automatic in the Steam Pressure Mode.

UPC-101 Steam Header Pressure transmitter, then fails to 0 psig.

With NO operator action, which of the following statements describes the resulting SG Level response?

Steam Generator levels will initially...

A. lower due to shrink and then raise as RCS temperature raises.

B. raise due to swell and then lower due to the higher steam release rate.

C. lower due to shrink and then raise due to higher FW flow at the lower pressure.

D. raise due to swell and then lower due to lower FW flow at the high pressure.

Answer: A The Steam Dumps will be trying to maintain Steam Line pressure. When the actual pressure fails low the dumps will close, causing SG pressure to be higher. This will cause an initial shrink followed by level rising as the RCS temperature raises.

B - Incorrect - The failure will cause steam dumps to close.

C - Incorrect - The SG will shrink but due to higher pressure.

D - Incorrect - The SG will shrink from the higher pressure.

REFERENCE:

RO-C-05200 Steam Dump System LESSON PLAN/OBJ: RO-C-05200/#7 KA - 041000 K1.02 Steam Dump System (SDS) and Turbine Bypass Control Knowledge of the physical connections and/or cause-effect relationships between the SDS and the following systems:

S/G level RO/SRO Value - (2.7 / 2.9) CFR - 41.2 to 41.9 / 45.7 to 45.8 Question #52 KA# - 041000 K1.02 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 63

53. 053 5 The following plant conditions exist:

- Unit 2 is at 9% power with the Main Turbine rolling at 1800 rpm.

- East Main Feedwater Pump is supplying main feedwater to the SGs.

- AFW pumps are aligned in Auto.

- All operating condensate booster pumps trip.

Which ONE of the following describes the system response and required operator action?

A. MFPs--Immediate Trip MDAFW Pumps Start from MFP trip Reactor Trip--On low low SG level Enter OHP-4023-E-0, Reactor Trip or Safety Injection B. MFPs--Immediate Trip MDAFW Pumps Start & Turbine Trips from AMSAC Enter OHP-4022-001-002, Loss of Load (Load Rejection)

C. MFPs--Trip after 5 sec. Delay MDAFW Pumps Start from MFP trip Reactor Trip--On low low SG level Enter OHP-4023-E-0, Reactor Trip or Safety Injection D. MFPs--Trip after 5 sec. Delay MDAFW Pumps Start & Turbine Trips from AMSAC Enter OHP-4022-001-002, Loss of Load (Load Rejection)

Answer: C The MFPs will trip after FW suction pressure has lowered to <180 psig for 5 seconds.

The trip of the Main FW pumps will cause the MDAFW pumps to start. The delayed trip of the FW pumps and the current power level will cause SG levels to Lower to the low low SG level reactor trip setpoint. After the reactor is tripped E-0 is entered.

A - Incorrect - The FW pumps will not trip immediately.

B - Incorrect - The FW pumps will not trip immediately. AMSAC will not actuate since power is not >40%.

D - Incorrect - AMSAC will not actuate since power is not >40%.

Modified to address Required Operator Action in stem. Added AMSAC to distracters and added procedure choices.

REFERENCE:

2-OHP 4021.001.006, Power Escalation LESSON PLAN/OBJ: RO-C-05500/#6 KA - 056000 A2.04 Condensate System Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of condensate pumps RO/SRO Value - (2.6 / 2.8) CFR - 41.5 / 43.5 / 45.3 / 45.13 Original Question Source : Master Bank 01055C0007-2 Original KA 035 A4.01 2.7/2.9 EXAM/QUIZZES: RO18PRAC-1; RO15 NRC EXAM; RO20NOP1M; RO24SYS03

DC Cook 2006 NRC Exam 64 LESSON PLAN/OBJ: RO-C-05500/#6;

REFERENCES:

2-OHP 4021.001.006, "Power Escalation Question #53 KA# - 056000 A2.04 Exam Level - RO Question Source - MODIFIED - 01055C0007-2 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 65

54. 054 3 If Turbine Bypass Header Pressure Transmitter UPC-101 fails HIGH during normal plant operation the MFP Speed Control System will generate a Delta-P signal ____(1)_____ than required, causing the main feed pump(s) to ______(2)_____.

(Assume the failover circuit does NOT function)

(1)

(2)

A.

larger speed up B.

larger slow down C.

smaller speed up D.

smaller slow down Answer: C The Main FW Pump Speed control compares the UPC-101 steam header pressure to the FW pump Discharge pressure. The speed control attempts to maintain the Main FW Pump speed such that the FW header to Steam Header DP is on Program. When the Steam Pressure fails high, it will appear that a smaller DP exists which will raise FW pump speed to try to raise FW pump Discharge header pressure.

A - Incorrect - Steam to FW discharge pressure DP will be lower.

B - Incorrect - Steam to FW discharge pressure DP will be lower. Response if Steam pressure failed low.

D - Incorrect - The controller will raise FW pump Speed.

REFERENCE:

SOD-05100-003 LESSON PLAN/OBJ: RO-C-05500/#8 KA - 059000 A1.07 Main Feedwater (MFW) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW System controls including:

Feed Pump speed, including normal control speed for ICS RO/SRO Value - (2.5 / 2.6) CFR - 41.5 / 45.5 SOURCE: Master Bank 01055C0008-5 Original Quest. KA - 059 K4.05 (2.5/2.8)

EXAM/QUIZZES: Q2404V; R2324B-A3B; RO23SYS3; RO24SYS03 Question #54 KA# - 059000 A1.07 Exam Level - RO Question Source - DIRECT - 01055C0008-5 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 66

55. 055 1 Following a reactor trip with NO SI required, the operators are performing the immediate actions of OHP-4023-E-0, Reactor Trip or Safety Injection Which of the following describes the basis for the following Step 4 RNO actions?

(1) Lower AFW flow to < 450,000 pph (2) Maintain AFW flow >240,000 pph until minimum S/G water level is obtained A. (1) Limit runout of the Aux Feed pumps.

(2) Ensure enough feedwater flow for decay heat removal.

B. (1) Limit runout of the Aux Feed pumps.

(2) Ensure enough flow for Aux Feed Pump protection.

C. (1) Limit overcooling of the RCS.

(2) Ensure enough feedwater flow for decay heat removal.

D. (1) Limit overcooling of the RCS.

(2) Ensure enough flow for Aux Feed Pump protection.

Answer: C

- If Si is not actuated and it is determined that an Si is not required AFW flow is reduced to limit the RCS cooldown due to excessive feed flow while ensuring that adequate AFW flow is maintained above the required value for decay heat removal until the minimum required SG narrow range level is restored. The intent of this action is to reduce excessive AFW flow to near the discharge flow of two MDAFW pumps at SG design pressure in a prompt manner prior to transitioning to ES-0.1 to allow for stabilizing RCS temperature near its normal no-load value in a timely manner. After transitioning to ES-0.1 the operators will follow the appropriate guidance in ES-0.1 for controlling AFW flow following a reactor trip.

A - Incorrect - AFW pumps are protected from runout through the flow retention system.

B - Incorrect - AFW pumps are protected from runout through the flow retention system.

The ELO valves cycle automatically on signals from their respective flow controllers to ensure AFW pump minimum flow.

D - Incorrect - The ELO valves cycle automatically on signals from their respective flow controllers to ensure AFW pump minimum flow.

REFERENCE:

12-OHP-4023-E-0, Reactor Trip or Safety Injection Background Step 4 pg. 16 LESSON PLAN/OBJ: RO-C-EOP03/#19 KA - 061000 K5.01 Auxiliary / Emergency Feedwater (AFW) System Knowledge of the operational implications of the following concepts as they apply to the AFW System:

Relationship between AFW flow and RCS heat transfer RO/SRO Value - (3.6 / 3.9) CFR - 41.5 / 45.7 SOURCE: INPO # 26387 Byron 1 - 12/10/2003 Original Quest. KA - 061000.A3.02 Question #55 KA# - 061000 K5.01

DC Cook 2006 NRC Exam 67 Exam Level - RO Question Source - DIRECT-INPO - BYRON 2003 - 26387 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 68

56. 056 21 The following conditions exist:

- Unit 1 is at 38% power.

- RCP busses 1A and 1B are energized from Reserve power

- Bus T11A voltage drops to 112 V indicated for 2 minutes Which ONE of the following describes the plant response and the required operator actions?

A. RCP bus-tie breakers to buses T11A and T11B open. Verify Reactor Trip and go to 01-OHP-4023-E-0, Reactor Trip or Safety Injection.

B. Reserve Feeder breakers to RCP buses 1A and 1B open. Verify Reactor Trip and go to 01-OHP-4023-E-0, Reactor Trip or Safety Injection.

C. RCP bus-tie breakers to buses T11A and T11B open. Verify 1AB DG energizes T11A and T11B and raise RAT feed voltage per 01-OHP-4021-082-026, Operation of the Load Tap Changer.

D. RCP buses 1A and 1B fast transfer to the Auxiliary Transformer 1AB. Verify T11A and T11B remain energized and raise RAT feed voltage per 01-OHP-4021-082-026, Operation of the Load Tap Changer.

Answer: C A Vital bus Undervoltage condition of 113 V will energize 62-1 T11A. After a 111 Second delay it will open T11A9 and T11B1 causing T11 A and T11B to lose power. This will cause the EDG to start and energize T11A and T11B. The RAT voltage should be raised using the Load Tap changers (this should have occurred automatically) as directed per 01-4021-082-026.

A - Incorrect - This does Not cause a Reactor trip.

B - Incorrect - The RCP Busses will remain energized and will only send an undervoltage signal (2/4 cause RX trip) when indicated voltage on the RCP bus is <88 volts.

D - Incorrect - The RCP buses will not transfer to the UATs. (They do auto transfer from UAT to RAT).

Modified by adding vital bus voltage and requesting actions required in the stem.

Distracters all changed.

REFERENCE:

01-OHP-4024-121, Annunciator #121 Response, Drop 78 Train B Aux Buses Undervoltage pg. 168 LESSON PLAN/OBJ: RO-C-08201/#6 KA - 062000 A2.08 A.C. Electrical Distribution System Ability to (a) predict the impacts of the following malfunctions or operations on the A.C.

Distribution System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Consequences of exceeding voltage limitations RO/SRO Value - (2.7 / 3.0) CFR - 41.5 / 43.5 / 45.3 / 45.13 SOURCE: Cook Master Bank 01082C0108-1 Original Quest. KA - 062 K1.04 A.C. Electrical Distribution System

DC Cook 2006 NRC Exam 69 Knowledge of the physical connections and/or cause-effect relationships between the A.C. Distribution System and the following systems:

Offsite Sources RO-3.7 SRO-4.2 EXAM/QUIZZES: R911516; RO1825; 99SEIRO7; 99SEIR20; RO2002; RO20SRO1; RQ2605C; RQ2605D; RO22SYSREVIEW; RO23SYS1R; RO23SYS5; RO23SYSCOMP; RQ2807E; RQ2807A; RO24SYS04 Question #56 KA# - 062000 A2.0 Exam Level - RO Question Source - MODIFIED - 01082C0108-1 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 70

[ NOTE: During administration of the written examination, the facility provided a clarification to distractor D. The words that are shown underlined were added to distractor D as a clarification, and the words that were deleted are shown as a strikeout. ]

57. 057 2 The operator incorrectly opens the breaker labeled "7.5 KVA Static Inverter Channel IV" on 250 VDC distribution panel "MCAB". The operator realizes the mistake and immediately recloses the breaker.

Which ONE of the following describes the effect of these actions, if any?

A. The alternate power source to the CRID Inverter will be lost when the breaker is reclosed.

The CRID will transfer to the 120 VAC from the Regulating Transformer.

B. The alternate power source to the CRID Inverter will be lost. No automatic action will occur when the breaker is reclosed. The auto transfer lockout must be reset at the inverter.

C. The normal power source to the CRID Inverter will be lost so it will auto transfer to the alternate source. When the breaker is reclosed, it will auto transfer to the normal source.

D. The normal power source to the CRID Inverter will be lost so it will auto transfer to the alternate source. No automatic action will occur When when the breaker is reclosed.,the The auto transfer lockout must be reset at the inverter.

Answer: C The static transfer switch provides a virtual zero time transfer to the alternate source in case of inverter failure. Thirty seconds after the static switch transfer event ceases and all system parameters are normal, the static switch automatically re-transfers the load to the inverter, without power interruption.

A - Incorrect - The Alternate source will not be lost. The normal DC supply will be restored and the Inverter will re-transfer to the normal source.

B - Incorrect - The Alternate source will not be lost. The normal DC supply will be restored and the Inverter will re-transfer to the normal source.

D - Incorrect - The normal DC supply will be restored when the breaker is closed and the Inverter will re-transfer to the normal source.

REFERENCE:

RO-C-08203, Instrumentation Electrical System pg. 19-20 LESSON PLAN/OBJ: RO-C-08203/#3e, #3g KA - 062000 A3.04 A.C. Electrical Distribution System Ability to monitor automatic operation of the A.C. Distribution System, including:

Operation of inverter (e.g., precharging synchronizing light, static transfer)

RO/SRO Value - (2.7 / 2.9) CFR - 41.7 / 45.5 Original Question Source: Bank 01082C0303-2 Original KA - 063 K4.01 (2.7 / 3.0)

EXAM/QUIZZES: ROC19C2; 99SEIR08; RO2001;RO20SRO1; RO22EOP8A ;

RO23SYS2; RO23AOP2; RO24SYSAUX

REFERENCES:

DC Aux One-Line Drawing OP2-1112060-8 Question #57 KA# - 062000 A3.04

DC Cook 2006 NRC Exam 71 Exam Level - RO Question Source - DIRECT - 01082C0303-2 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 72

58. 058 3 If the 250 VDC CD bus deenergizes with the Unit at 100% power, the initiating signal for the reactor trip is due to a loss of power to:

A. CRIDS I and II.

B. CRIDS III and IV.

C. two RCP Bus underfrequency relays.

D. all four Feed Reg Valves.

Answer: C The Loss of 250 VDC will cause the loss of Power to the RCP Bus Underfrequency relays. This will generate a direct reactor trip.

A - Incorrect - CD does supply power to CRID I & II, but the backup AC source will continue to supply the CRIDs.

B - Incorrect - CD does NOT supply power to CRID III & IV, and the backup AC source will continue to supply the CRIDs.

D - Incorrect - The Reactor trips from the RCP Underfrequency (initiating signal is requested in stem). The FW Reg valves will fail closed which leads to a trip but this is not the initiating signal.

REFERENCE:

RO-C-AOP10, Abnormal Operating Procedures - Day 10, OHP-4022-082-002CD, Loss of Power to 250VDC Bus 1CD LESSON PLAN/OBJ: RO-C-AOP10/#10 KA - 063000 K4.04 D.C. Electrical Distribution System Knowledge of D.C. Electrical System design feature(s) and/or interlock(s) which provide for the following:

Trips RO/SRO Value - (2.6 / 2.9) CFR - 41.7 ORIGINAL Question Source : Bank AOP1CAOP10B-2 Original Question KA - SYS 045 K1.20 (3.4/3.6)

EXAM/QUIZZES: RO22AOPCOMP; RO22EOP8A Question #58 KA# - 063000 K4.04 Exam Level - RO Question Source - DIRECT - AOP1CAOP10B-2 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 73

59. 059 2 The following conditions exist:

- The Plant is operating at 100% power.

- 2CD EDG is started for surveillance testing.

Which ONE (1) of the following describes how a reverse power condition is prevented when closing the EDG output breaker?

A. Ensure incoming voltage is slightly higher than running prior to closing the output breaker.

B. Ensure synchroscope is at 12 o'clock position prior to closing the output breaker.

C. Ensure synchroscope is rotating slowly in the 'FAST' (clockwise) direction prior to closing the breaker.

D. Ensure running and incoming frequencies are matched prior to closing the output breaker Answer: C The procedure requires the Operator to ensure that the synchroscope is rotating slowly in the 'FAST' (clockwise) direction prior to closing the breaker. This causes the DG to be slightly faster than the Grid causing it to pick up some load thus preventing a reverse power trip.

A. - Incorrect. Ensuring voltage is matched ensures no large reactive current flow on breaker closure.

B. - Incorrect. The synchroscope is required to be at 12 o'clock but this does not prevent reverse power.

D. - Incorrect. Matching frequency may actually increase the chance of reverse power because the incoming machine will take less load on parallel

REFERENCE:

02-OHP-4021-032-001 Attachment 2 DG2AB Operation on Safeguards Buses Section 4.2 LESSON PLAN/OBJ: RO-C-03200/#10 KA - 064000 A1.08 Emergency Diesel Generator (ED/G) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G System controls including:

Maintaining minimum load on ED/G (to prevent reverse power)

RO/SRO Value - (3.1 / 3.4) CFR - 41.5 / 45.5 SOURCE: INPO # 23362 Indian Point 3 (Unit) - 3/10/2003 Original Quest. KA - 064.A1.08 Question #59 KA# - 064000 A1.08 Exam Level - RO Question Source - DIRECT-INPO - IP3 2003 - 23362 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 74

60. 060 21 Which ONE of the following actions, if any, will occur due to a HIGH rad alarm on monitor RRS-1001, Liquid Radwaste Effluent Line?

A. Alarms at WDS and Control Rooms only - no automatic actions.

B. RRV-285, liquid waste disposal effluent discharge header shutoff valve closes.

C. RRV-286 or RRV-287 liquid waste effluent to header CIRC water discharge header shut off valve closes.

D. RRV-284, liquid release control valve closes.

Answer: B The high alarm on RRS-1001 will cause RRV-285 to close and also trip the WECT and monitor tank pumps.

A - Incorrect - The High Level causes the isolation. The alert level just gives an alarm.

C - Incorrect - RRV-286 or 287 are used to select which unit to discharge to and will close on loss of CW flow. These valves do not auto close on High radiation.

D - Incorrect - RRV-284 controls the release flow rate but does not isolate on High Radiation.

REFERENCE:

12-OHP 4024.139 Drop 16 LESSON PLAN/OBJ: RO-C-01350/#4 KA - 068000 A3.02 Liquid Radwaste System (LRS)

Ability to monitor automatic operation of the Liquid Radwaste System, including:

Automatic isolation RO/SRO Value - (3.6 / 3.6) CFR - 41.7 / 45.5 SOURCE: RO23 Audit (RO# 099/SRO# N/A)

From Bank AS06-22 Original Quest. KA - SYS 068 A3.02 (3.6/3.6)

EXAM/QUIZZES: AE26AE1; AE26AE2; AE26AE3; AE26AE4; AE26AE5; UO22FNL; AE2703V; AE2703A; AE2703C; AE2709AC Question #60 KA# - 068000 A3.02 Exam Level - RO Question Source - DIRECT - AUDIT RO23 Q#99 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 75

61. 061 10 Unit 1 is in Mode 5 following a refueling outage. The Containment Purge System was operating in the VENTILATION MODE with the following lineup:

Purge Supply Fan HV-CPS RUNNING Purge Exhaust Fan HV-CPX RUNNING Purge Supply to Upper Containment VCR-105 and VCR-205 OPEN Purge Exhaust from Upper Containment VCR-106 and VCR-206 OPEN Following a HIGH alarm on ERS-1401, Lower Containment Radiation Monitor, the Containment Purge System is aligned as follows:

Purge Supply Fan HV-CPS RUNNING Purge Exhaust Fan HV-CPX RUNNING Purge Supply to Upper Containment VCR-105 and VCR-205 OPEN Purge Exhaust from Upper Containment VCR-106 and VCR-206 OPEN Which ONE of the following describes the required operator actions?

A. Stop the Containment Purge and declare Containment Ventilation Isolation inoperable.

B. Stop the Containment Purge and notify Radiation Protection.

C. Continue the Purge as long as VRS-1101, Containment Normal Range Area Radiation Monitor is still indicating as expected.

D. Continue the Purge as long as VRS-1505, Auxiliary Building Ventilation Noble Gas Activity Monitor is still indicating as expected.

Answer: B When the Containment Purge system is operating in the Ventilation Mode, the automatic isolation signals are blocked. The procedure requires the Purge to be stopped and radiation protection notified.

A - Incorrect - When the Containment Purge system is operating in the Ventilation Mode, the automatic isolation signals are blocked.

C - Incorrect - The procedure requires the Purge to be stopped and radiation protection notified.

D - Incorrect - The procedure requires the Purge to be stopped and radiation protection notified. (This monitor is required per ODCM)

REFERENCE:

01-OHP-4021-028-005, Operation of the Containment Purge System (Ventilation Mode) pg. 29 LESSON PLAN/OBJ:RO-C-02800 / #9 KA - 073000 K3.01 Process Radiation Monitoring (PRM) System Knowledge of the effect that a loss or malfunction of the PRM System will have on the following:

Radioactive effluent releases RO/SRO Value - (3.6 / 4.2) CFR - 41.7 / 45.6 SOURCE: INPO # 27721 Cook 1 - 4/29/2004(RO#087/SRO#NA)

Modified from COOK02-086-1 Original Quest. KA - 029000 2.1.23 Containment Purge System (CPS)

DC Cook 2006 NRC Exam 76 Conduct of Operations Ability to perform specific system and integrated plant procedures during all modes of plant operation.

RO-3.9 SRO-4.0 Question #61 KA# - 073000 K3.01 Exam Level - RO Question Source - DIRECT - COOK 2004 #87 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 77

62. 062 2 During a surveillance test on the Unit 1 train A SI, the East Essential Service Water (ESW) pump failed to receive an auto-start signal. The pump started normally in manual.

Which ONE of the following describes the operability and Technical Specification (TS) applicability associated with the Unit 1 East ESW pump?

The Unit 1 East ESW Pump is...

A. still operable because it can still be manually started and a service water TS LCO action statement would not be entered.

B. inoperable and a service water TS LCO action statement would be entered because the auto start is required to be operable.

C. still operable because it will start automatically if the pump's discharge pressure falls below 40 psig and a service water TS LCO action statement would not be entered.

D. inoperable and a service water TS LCO action statement would be entered only if the associated Unit 2 ESW pump is also inoperable or the crossties are closed.

Answer: B Technical Specification 3.7.8 Essential Service Water Systems, SR 3.7.8.3 requires the auto start function of the ESW pump for operability.

A - Incorrect - Auto start is required for operability.

C - Incorrect - While the pump may start at 40 psig the SI auto start would also be required for operability.

D - Incorrect - If the crossties are open then the LCO action statement also applies to Unit 2.

REFERENCE:

Technical Specification 3.7.8 Essential Service Water Systems, SR 3.7.8.3 LESSON PLAN/OBJ: RO-C-01900/#15 KA - 076000 2.1.33 Service Water System (SWS)

Conduct of Operations Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

RO/SRO Value - (3.4 / 4.0) CFR - 43.2 / 43.3 / 45.3 SOURCE: INPO # 2836 Point Beach 1 - 8/2/1999 Original Quest. KA - 076000.G2.2 Question #62 KA# - 076000 2.1.33 Exam Level - RO Question Source - DIRECT-INPO - PNT BEACH 1999-2836 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 78

63. 063 3 Unit 2 is operating at power with the South NESW Pump supplying the Unit 2 and Miscellaneous Header Heat Loads. An electrical fault on Transformer TR21B causes the loss of Bus 21B. The NESW System response to this event would be to:

A. start the Unit 2 North NESW Pump due to low header pressure.

B. start the Unit 1 South NESW Pump due to load shed.

C. open the Miscellaneous Header cross connect valves on load shed.

D. continue operating as before the loss, except that the North NESW pump would not be available.

Answer: A The South NESW pump is supplied from Bus 21B. On a loss of power the Unit 2 North NESW pump will auto start on Low NESW Header Pressure.

B - Incorrect - The Unit 1 NESW pump will not start due to a loss on Unit 2.

C - Incorrect - The Miscellaneous header Cross connect valves do not automatically reposition.

D - Incorrect - The south NESW pump is supplied from bus 21B. The North pump doe not lose power.

REFERENCE:

RO-C-02000, Non-Essential Service Water LESSON PLAN/OBJ: RO-C-02000/#12 KA - 076000 K2.04 Service Water System (SWS)

Knowledge of bus power supplies to the following:

Reactor building closed cooling water RO/SRO Value - (2.5 / 2.6) CFR - 41.7 SOURCE: Master Bank 01020C0012-2 Original Quest. KA - 076 K6.04 (2.1/2.2)

EXAM/QUIZZES: R921717; RO19C1; RQ2804E LESSON PLAN/OBJ: ;

REFERENCES:

SD-02000 Question #63 KA# - 076000 K2.04 Exam Level - RO Question Source - DIRECT - 01020C0012-2 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 79

64. 064 11 Which ONE of the following describes the normal operation of the Plant and Control Air systems?

A. The standby Plant Air Compressor will automatically start at 90 psig in the ring header.

B. Control Air pressure lowering to 90 psig in the wet air receiver will start the Control Air Compressor.

C. The operating Plant Air Compressor cycles on and off between 93 and 100 psig when operating in automatic.

D. The Unit 1 Control Air header will automatically cross connect to the Unit 2 Control Air header if the Unit 1 Control Air Compressor fails to maintain header pressure above 85 psig.

Answer: B At 90 psig in the Control air systems the Control air compressors start to supply their respective units.

A - Incorrect - The PAC will auto start at 95 psig.

C - Incorrect - The CAC will cycle between 93 and 100 psig.

D - Incorrect - The Plant Air Header Isolates at 85 psig.

REFERENCE:

SD-06400 LESSON PLAN/OBJ: RO-C-06401/#4 KA - 078000 A3.01 Instrument Air System (IAS)

Ability to monitor automatic operation of the IAS, including:

Air pressure RO/SRO Value - (3.1 / 3.2) CFR - 41.7 / 45.5 SOURCE: Audit RO22 #18 Original Quest. KA - SYS 078 A3.01 Question #64 KA# - 078000 A3.01 Exam Level - RO Question Source - DIRECT - AUDIT RO22 #18 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 80

65. 065 1 With the plant at 100% power, an inadvertent Containment Isolation Phase A occurs. The reactor has NOT tripped and Safety Injection has NOT actuated.

How has the Phase A signal affected the Chemical and Volume Control System Containment paths?

A. ONLY the Letdown Containment Isolation Valves (QCR-300/301), Letdown Orifice Isolation Valves (QRV-160/161/162) and the Seal Water Return Containment Isolations (QCM-250/350) have closed.

B. ONLY the Letdown Containment Isolation Valves (QCR-300/301), Seal Water Return Containment Isolations (QCM-250/350) and Charging Header Isolation Valves (QMO-200/201) have closed.

C. ONLY the Letdown Containment Isolation Valves (QCR-300/301) and the Letdown Orifice Isolation Valves (QRV-160/161/162) have closed.

D. None of the CVCS System valves have closed, since the SI signal is what isolates the CVCS Containment paths.

Answer: A Containment Phase A isolates the Letdown Containment Isolation Valves (QCR-300/301), Letdown Orifice Isolation Valves (QRV-160/161/162) and the Seal Water Return Containment Isolations (QCM-250/350).

B - Incorrect - The Charging header Isolation Valves are closed by the SI signal.

C - Incorrect - The Seal water return valves will also isolate.

D - Incorrect - Phase A will isolate the Letdown & Seal Return Paths.

REFERENCE:

SOD-00300-001 Charging and Letdown System LESSON PLAN/OBJ: RO-C-01100/#6 KA - 103000 K4.06 Containment System Knowledge of Containment System design feature(s) and/or interlock(s) which provide for the following:

Containment isolation system RO/SRO Value - (3.1 / 3.7) CFR - 41.7 SOURCE: INPO # 27098 Millstone 3 - 7/16/2004 Original Quest. KA - 103.K4.06 Question #65 KA# - 103000 K4.06 Exam Level - RO Question Source - DIRECT-INPO - MILLSTONE2004-27098 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 81

66. 066 1 A plant heatup was in progress, with the RCS at 300oF and 350 psig, when the South SI pump failed its surveillance and was declared inoperable. Which ONE of the below limits are placed on the plant heatup?

A. ECCS LCO for modes 3 and 4 satisfied but may not proceed into mode 2.

B. ECCS LCO for this mode satisfied but may not proceed into mode 3.

C. ECCS LCO satisfied for this mode but enter a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock when proceeding into mode 3.

D. ECCS LCO not satisfied, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore South SI pump, then 24 hrs to mode 5.

Answer: B Mode 4 requires only 1 train of ECCS (with only RHR & CCP) per TS 3.5.3. The SI pumps are not required to be operable until Mode 3 (>350oF). Entry into Mode 3 is not allowed since TS 3.0.4 prohibits changing modes if TS 3.5.2 is not met (2 full ECCS trains are required).

A - Incorrect since the LCO requirement for modes 1-3 are not met.

C - Incorrect since the LCO for the current mode is met.

D - Incorrect since TS 3.0.4 prohibits changing mode 4 to 3.

REFERENCE:

Technical Specification 3.5.2, 3.5.3, 3.0.4 LESSON PLAN/OBJ: RO-C-TS01/#11 KA - Generic 2.1.12 Generic Conduct of Operations Ability to apply technical specifications for a system.

RO/SRO Value - (2.9 / 4.0) CFR - 43.2 / 43.5 / 45.3 SOURCE: INPO # 28264 Ginna 1 - 11/1/2004 Original Quest. KA - G2.1.12 Question #66 KA# - GENERIC 2.1.12 Exam Level - RO Question Source - DIRECT-INPO - GINNA 2004-28264 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 82

67. 067 3 During steady state 100% power operation, you are performing a Plant Computer Online Calorimetric Validation in accordance with 01-OHP-4030-114-029 Reactor Thermal Power.

The computer points for Steam Generator (S/G) Blowdown are reading as follows:

SG #

11 12 13 14 PPC Value(gpm) 98 100 95 96 PPC Color Green Magenta Green Green Local Value(gpm) 98 105 95 95 Which ONE of the following would be a correct response to these conditions?

A. Use the online calorimetric since the computer value for the #12 S/G is less than the actual blowdown.

B. Manually enter a value of 105 gpm for the #12 S/G Blowdown and insure the online calorimetric remains below 100%.

C. Isolate SG Blowdown. A Manual calorimetric may be used after the plant has been stable for 10 minutes.

D. Use the online calorimetric since the computer value for the #12 S/G is within allowed tolerance. The magenta color indicates a deviation from the other SGs.

Answer: C The computer input for #12 SG blowdown is inaccurate. The magenta reading means that it is inaccurate and should not be used. The SG Blowown Flow should be Isolated.

The manual calorimetric may then be used.

A - Incorrect - The Computer value is inaccurate and should not be used if it is blue or magenta. Using a lower value is non conservative.

B - Incorrect - The program does not have provisions for manually entering SG blowdown flows.

D - Incorrect - The Computer value is inaccurate and should not be used if it is blue or magenta.

REFERENCE:

01-OHP-4030-114-029 Reactor Thermal Power (Attachment 4 Prereq.

2.2 & step 4.4)

LESSON PLAN/OBJ: RO-C-NOP07/#7.37 KA - Generic 2.1.19 Generic Conduct of Operations Ability to use plant computer to obtain and evaluate parametric information on system or component status.

RO/SRO Value - (3.0 / 3.0) CFR - 45.12 SOURCE: INPO # 24112 Salem Unit 1 - 5/5/2003 Original Quest. KA - 194001.G1.19 Question #67 KA# - GENERIC 2.1.19 Exam Level - RO Question Source - DIRECT-INPO - SALEM 2003 - 24112 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 83

68. 068 2 During an independent verification a valve is found out of position.

Which of the following is the way the component out of position shall be handled?

The component shall be repositioned...

A. by the person who did the initial lineup and then verified B. and verified by the person performing the verification C. and the supervisor notified of the discrepancy D. after the supervisor gives the approval Answer: D PMP-4043-ICV-001 requires notification of the supervisor for resolution of the discrepancy.

A - Incorrect - Procedure does not specify who realigns valve, but supervisor approval is required prior to repositioning.

B - Incorrect - Valve is repositioned and then independently verified after supervisor approval.

C - Incorrect - Valve is repositioned and then independently verified after supervisor approval.

REFERENCE:

PMP-4043-ICV-001, Independent and Concurrent Verification Section 3.3.9 LESSON PLAN/OBJ: RO-C-ADM02/#5 KA - Generic 2.1.29 Generic Conduct of Operations Knowledge of how to conduct and verify valve lineups.

RO/SRO Value - (3.4 / 3.3) CFR - 41.10 / 45.1 / 45.12 SOURCE: INPO # 19510 Cook 1 - 5/21/2001 Original Quest. KA - g.2.1.29 Question #68 KA# - GENERIC 2.1.29 Exam Level - RO Question Source - DIRECT - COOK 2001 Q#56 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 84

69. 069 4 When during the review/approval process does a On-The-Spot Change (OTSC) become available for immediate use?

A. Once the applicable 10CFR50.59 review is complete.

B. Following SRO review of the change and signature.

C. After the Qualified Technical Reviewer reviews the change.

D. When the Department Manager signs the change.

Answer: B The procedure is ready for use after the SRO has approved the changes.

A - Incorrect - The 10CFR 50.59 review may be done after the procedure is used.

C - Incorrect - The SRO must review and approve the changes after the QTR.

D - Incorrect - The department manager does not need to sign.

REFERENCE:

PMP-2010-PRC-003 Procedure Use and Adherence Section 3.5.2 LESSON PLAN/OBJ: RO-C-ADM12/#3.2 KA - Generic 2.2.11 Generic Equipment Control Knowledge of the process for controlling temporary changes.

RO/SRO Value - (2.5 / 3.4) CFR - 41.10 / 43.3 / 45.13 SOURCE: Master Bank 01ADMC12-3 Original Quest. KA - 2.2.6 (2.3/3.3)

EXAM/QUIZZES: RO23ADM1 Question #69 KA# - GENERIC 2.2.11 Exam Level - RO Question Source - DIRECT - 01ADMC12-3 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 85

70. 070 2 Unit 1 is in mode 6 and in the process of core offload.

Which ONE (1) of the following limiting conditions requires immediate suspension of all CORE ALTERATIONS?

A. Source Range Channel N32 is inoperable with Gamma-Metrics channels N21 and N23 both operable.

B. Loss of direct communications between control room and personnel at the refueling station.

C. Loss of the Fuel Handling Ventilation System.

D. Time since entering mode 3 is 155 hours0.00179 days <br />0.0431 hours <br />2.562831e-4 weeks <br />5.89775e-5 months <br />.

Answer: B 01-OHP-4030-127-037, Refueling Surveillance & TRO 8.91 require that core alterations be suspended if direct communications are lost.

A-Incorrect-Plausible since Tech Spec 3.9.2 requires 2 operable source range channels. Wide range flux monitors (gamma metrics) are allowed to substitute for NIS source flux monitors. BUT the gamma metrics do NOT provide an audible function.(B 3.9.2)

C - Incorrect-Plausible since Fuel Handling Ventilation is Required for Movement in SFP area not containment (Unlatch requires no SFP crane movements) LCO 3.7.13 D - Incorrect - Plausible since time limits apply to refueling except times are - Shutdown

>100 hours Sept-June and 148 Hours June-Sept (TRO 8.9.2)

Updated to ITS (SWP 7-28-05)

REFERENCE:

Admin TRO 8.9.1, 01-OHP-4030-127-037, Refueling Surveillance Data Sheet 2 pg. 19 LESSON PLAN/OBJ: RO-C-ADM13/#ADM13.3.0 KA - Generic 2.2.27 Generic Equipment Control Knowledge of the refueling process.

RO/SRO Value - (2.6 / 3.5) CFR - 43.6 / 45.13 SOURCE: RO23 Audit RO23-012-5 (#12)

Original Quest. KA - 2.2.27 Question #70 KA# - GENERIC 2.2.27 Exam Level - RO Question Source - DIRECT - RO23-012-5 (#12)

Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 86

71. 071 2 Prior to entering the Auxiliary Building for a tour, you identify a limited access area as having the following readings:

Surface Contamination Level of 80 dpm/100 cm2 (Alpha)

General Area Radiation Level of 1200 mRem/hr Airborne Radioactivity Level of 7 DAC-hours/week Which ONE of the following postings is required at the entrance to this area?

A. High Radiation and Contamination Area B. Very High Radiation and Airborne Area C. Locked High Radiation and Contamination Area D. Very High Radiation, Contamination and Airborne Area Answer: C a Locked High Radiation area is defined as an Area accessible in which radiation levels could result in a dose equivalent in excess of 1000 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm. A Contamination area is defined as An area where loose surface contamination may exceed 1000 dpm/100 cm2 smear or 20 dpm/100 cm2 of.

A - Incorrect - The area is required to be posted as a LOCKED high rad area.

B - Incorrect - The area does not exceed the Very High limit of 500 rads in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter nor the Airborne level of 12 DAC-hours/week.

D -Incorrect - The area does not exceed the Very High limit of 500 rads in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter nor the Airborne level of 12 DAC-hours/week.

REFERENCE:

PMI-6010 Rev 15 section 4.7 LESSON PLAN/OBJ: RO-C-RP02/#7 KA - Generic 2.3.4 Generic Radiological Controls Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

RO/SRO Value - (2.5 / 3.1) CFR - 43.4 / 45.10 SOURCE: INPO # 20261 Cook 1 - 9/10/2001 Original Quest. KA - g2.3.4 Question #71 KA# - GENERIC 2.3.4 Exam Level - RO Question Source - DIRECT - COOK 2001 RT Q#97 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 87

72. 072 21 Which ONE of the following describes the Operation of the Containment Purge System (in Ventilation Mode) while the Containment equipment Hatch is open?

A. Air flow must be OUT of Containment to prevent to minimize radiation levels.

B. Air flow must be INTO Containment to prevent the spread of contamination.

C. Containment Purge Exhaust and Supply flows must be matched to ensure the Containment and Aux Building are maintained at the same pressure.

D. Containment Purge Exhaust and Supply flows must be balanced to prevent Ice Condenser doors from opening.

Answer: B A lower pressure in containment with respect to the Aux Building will cause an airflow into containment and help to minimize the spread of contamination.

A - Incorrect - Air Flow is maintained into Containment.

C - Incorrect - Aux Building and Containment Pressures are not maintained equal.

D - Incorrect - This concern is addressed by the Upper/Lower Containment Pressure balance.

REFERENCE:

01-OHP-4021-028-005, Operation Of The Containment Purge System,, step 3.7 and Figure 2 LESSON PLAN/OBJ: RO-C-02800 /#4 KA - Generic 2.3.9 Generic Radiological Controls Knowledge of the process for performing a containment purge.

RO/SRO Value - (2.5 / 3.4) CFR - 43.4 / 45.10 Question #72 KA# - GENERIC 2.3.9 Exam Level - RO Question Source - NEW -

Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 88

73. 073 10 Unit 1 has experienced a Large Break LOCA. All safeguards equipment functioned properly following the event initiation. You are the BOP assigned to perform 01-OHP-4023-E-0, Reactor Trip or Safety Injection Attachment A.

Which ONE of the following describes the action required for the Control Room Pressurization fans and why?

A. Manually start both pressurization fans to ensure that enough pressure exists to ensure adequate filter flow.

B. Verify that both pressurization fans automatically start to ensure that enough pressure exists to ensure adequate filter flow.

C. Manually stop one pressurization fan to ensure that control room dose remains within analyzed limits.

D. Notify Unit 2 control room to start both pressurization fans if one Unit 1 fan is NOT running to ensure that control room dose remains within analyzed limits.

Answer: C Attachment A Step 4 provides direction to stop 1 pressurization fan to limit the filter flow rates to ensure the dose remains within limits.

A - Incorrect - Both fans are expected to auto start and one fan must be stopped.

B - Incorrect - One fan must be stopped to limit the filter flow rate.

D - Incorrect - One pressurization fan for each Unit through its respective (independent) filter train is required.

REFERENCE:

01-OHP-4023-E-0, Reactor Trip or Safety Injection Attachment A pg. 35 PSBD 12-OHP-4023-E-0 background document pg. 75 LESSON PLAN/OBJ: RO-C-EOP03/#22 KA - Generic 2.3.10 Generic Radiological Controls Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

RO/SRO Value - (2.9 / 3.3) CFR - 43.4 / 45.10 SOURCE: INPO # 27700 Cook 1 - 4/29/2004 (RO#066/SRO#066)

Original Quest. KA - 2.3.10 Generic Radiological Controls Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

RO-2.9 SRO-3.3 Question #73 KA# - GENERIC 2.3.10 Exam Level - RO Question Source - DIRECT - COOK 2004 Q#66 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 89

74. 074 1 FOLLOWING the Initial Notifications for an Emergency Plan accident classified as an ALERT, the SEC must FAX an updated EMD-32 form to the State of Michigan every ____ minutes until relieved by the EOF.

A. 15 B. 30 C. 60 D. 120 Answer: B PMP-2080-EPP-100 Attachment 8 Section 3.1 requires EMD32 followup notifications every 30 minutes.

A - Incorrect - 15 minutes is required for initial notifications.

C - Incorrect - 60 Minutes is the notification time for the NRC D - Incorrect - Followup notifications are required every 30 minutes.

Changed stem to Following initial notification and to Michigan vs sheriff. Changed Distracters from 5,10,15,20 minutes. Answer Changed to 30 Minutes from 15 minutes.

REFERENCE:

PMP-2080-EPP-100 Attachment 8 Section 3.1 LESSON PLAN/OBJ: ST-C-EP04/#5 KA - Generic 2.4.29 Generic Emergency Procedures/Plan Knowledge of the emergency plan.

RO/SRO Value - (2.6 / 4.0) CFR - 43.5 / 45.11 SOURCE: INPO # 20238 Cook 1 - 9/10/2001 from 12EPPC0303 2 Original Quest. KA - g2.4.29 Question #74 KA# - GENERIC 2.4.29 Exam Level - RO Question Source - MODIFIED - COOK 2001 RT Q#73 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 90

75. 075 1 The plant is in Mode 4 on RHR. NPS-121, Wide Range RCS pressure instrument is de-energized for maintenance. During a control board walkdown you discovered that Panel 206 Drop 36, RHR OPEN TO HI RCS HOT LEG PRESSURE, was illuminated.

What would cause this condition?

A. RHR valve 2-ICM-129 (RHR pump suction from Loop 2 hot leg) is open, reactor coolant system pressure is 500 psig.

B. RHR valve 2-ICM-129 (RHR pump suction from Loop 2 hot leg) is closed, reactor coolant system pressure is 450 psig.

C. RHR valve 2-IMO-128 (RHR pump suction from Loop 2 hot leg) is open, reactor coolant system pressure is 500 psig.

D. RHR valve 2-IMO-128 (RHR pump suction from Loop 2 hot leg) is closed, reactor coolant system pressure is 450 psig.

Answer: C Panel 206 Drop 36 will alarm with either NPS-121 at >491.25 psig and ICM-129 open or with NPS-122 at >491.25 psig and IMO-128 open. Since NPS-121 is deenergized the alarm is from IMO-128 and rising pressure.

A - Incorrect - ICM-129 is alarmed from NPS-121 which is De-energized per the stem.

B - Incorrect - ICM-129 not FULLY closed and rising pressure causes the alarm from NPS-121 which is De-energized per the stem.

D - Incorrect - IMO-128 not FULLY closed and rising pressure causes the alarm.

REFERENCE:

02-OHP 4024.206 Drop 36, SOD-01700-002 LESSON PLAN/OBJ: RO-C-01700/#7 KA - Generic 2.4.50 Generic Emergency Procedures/Plan Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

RO/SRO Value - (3.3 / 3.3) CFR - 45.3 SOURCE: INPO # 19318 Cook 1 - 5/21/2001 Original Quest. KA - 000025.g2.4 Question #75 KA# - GENERIC 2.4.50 Exam Level - RO Question Source - DIRECT - COOK 2001 Q#10 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 91

76. 076 11 Technical Specification LCO 3.5.5 Seal Injection Flow places limits on RCP seal injection flow resistance.

Which ONE of the following describes the reason for maintaining RCP seal injection flow resistance?

A. If the resistance is too small, Auxiliary Spray flow may be less than required for RCS depressurization during a Steam Generator Tube Rupture.

B. If the resistance is too large, RCP seals may overheat during a large break LOCA.

C. If the resistance is too large, Charging Pump minimum flow may be less than required for pump cooling during a small break LOCA.

D. If the resistance is too small, ECCS injection flow may be less than required for core cooling during a small break LOCA.

Answer: D The limitation on seal line resistance is to ensure that the minimum safeguards flow to the RCS will be sufficient for core cooling. The analysis assumes that flow diverted from the BIT line to seal injection is lost to core cooling.

A - Incorrect - Auxiliary Spray flow varies depending on plant conditions. it is not required to be any specific value during accident conditions.

B - Incorrect - In the event of a Large Break LOCA the CCP will inject less flow to the RCP seals but the limit requires a minimum seal resistance not maximum.

C - Incorrect - Charging flow minimum flow will come off prior to seal injection. (Pump is only concerned with total flow for cooling)

REFERENCE:

Technical Specification 3.5.5 and Bases LESSON PLAN/OBJ: RO-C-00200/#14 KA - 000009 2.2.25 Small Break LOCA Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

RO/SRO Value - (2.5 / 3.7) CFR - 43.2 Question #76 KA# - 000009 2.2.25 Exam Level - SRO Question Source - DIRECT - COOK 2004 SRO#074 Cognitive/Difficulty Level - F/4

DC Cook 2006 NRC Exam 92

77. 077 5 Six hours ago, Unit 2 experienced a Large break LOCA. All equipment operated as designed.

The following plant conditions exist:

z ECCS pumps have been aligned per 02-OHP-4023-ES-1.3,Cold Leg Recirculation.

z RCS Pressure is 95 psig z RCS temperature is 215oF z 02-OHP-4023-E-1, Loss of Reactor or Secondary Coolant is in progress.

The Reactor Operator has just informed you that Containment Pressure has lowered to 1.8 psig.

Which ONE of the following describes the required action(s) concerning Containment Spray operation?

A. Containment Spray may be secured.

B. The Containment Spray Pump must continue to operate until the Spray Additive Tank is drained.

C. Containment Spray must continue to operate for 1 more hour.

D. Containment Spray must continue to operate until Containment pressure lowers to less than 1.1 psig.

Answer: A CTS is reset and secured after it has operated for at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressure is <2 psig.

B - Incorrect - The Spray additive tank should already be drained.(4000 gal/18.5gpm)

C - Incorrect - CTS operation is only required for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. After 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> ES-1.4 Hot leg recirc is implemented.

D - Incorrect - CTS may be reset and secured when pressure lowers to <2 psig Modified Stem by changing operating time to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (from 4) and removed failure of West RHR. This changed answer to A (from C). Removed references fto RHR in Distractors.

REFERENCE:

02-OHP-4023-E-1 Loss of Reactor or Secondary Coolant Step 7 pg. 8-10 PSBD 12-OHP-4023-E-1 Background Document Step 7 Basis pg. 16-18 LESSON PLAN/OBJ: RO-C-EOP09/#36 KA - 026000 A2.08 Containment Spray System (CSS)

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

Safe securing of containment spray (when it can be done)

RO/SRO Value - (3.2 / 3.7) CFR - 41.5 / 43.5 / 45.3 / 45.13 Original Question Source COOK 2004 NRC Exam SRO#075 Large Break LOCA

DC Cook 2006 NRC Exam 93 Ability to determine and interpret the following as they apply to a Large Break LOCA:

Conditions necessary for recovery when accident reaches stable phase RO-3.4 SRO-3.9 Question #77 KA# - 026000 A2.08 Exam Level - SRO Question Source - MODIFIED - COOK 2004 SRO#075 Cognitive/Difficulty Level - H/4

DC Cook 2006 NRC Exam 94

78. 078 3 The unit has suffered a large break LOCA. The operators are currently performing OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation, step 12, realigning SI pump Recirculation valves IMO-262 and IMO-263. While attempting to close IMO-263, SI Pump Recirc to RWST, the MCC breaker trips open on overload and will not reset. The crew is expected to...

A. not continue with the procedure until the AEO has locally closed IMO-263, SI Pump Recirc to RWST, to prevent contaminating the RWST.

B. not continue with the procedure until the AEO has locally closed IMO-263, SI Pump Recirc to RWST, since it is an interlock to open IMO-340, CCP Suction from East RHR Hx, and IMO-350, SI Pump Suction from West RHR Hx.

C. verify IMO-262, SI Pump Recirc to RWST, is closed and continue with the procedure since adequate isolation exists and required interlocks are met.

D. leave the SI pumps aligned to the RWST until the operability of IMO-263, SI Pump Recirc to RWST, is resolved.

Answer: C OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation, step 12, requires that 1 of the SI recirc valves be closed. The valves are in series to isolate flow back to the RWST.

Closing at least one valve satisfies the interlock and prevents the return of radioactive water to the RWST.

A - Incorrect - Only 1 valve is required to be isolated.

B - Incorrect - Only 1 valve is required to be isolated.

D - Incorrect - Only 1 valve is required to be isolated. The SI pumps can not be left aligned to the RWST since its level is lowering.

REFERENCE:

RO-C-EOP09 Study Guide, ES-1.3 Background Step 12 LESSON PLAN/OBJ: RO-C-EOP09/#36 KA - 000011 2.3.10 Large Break LOCA Radiological Controls Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

RO/SRO Value - (2.9 / 3.3) CFR - 43.4 / 45.10 SOURCE: Bank 01EOPC0922-11 Original Quest. KA - 000011 EA1.13 (4.1/4.2)

EXAM/QUIZZES: RO20ECOMP; RO21EOP5; RQ2604A; RQ2604C; RQ2604E; RO22EOP5A; RO22EOP5B Question #78 KA# - 000011 2.3.10 Exam Level - SRO Question Source - DIRECT - 01EOPC0922-11 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 95

79. 079 6 Unit 2 is performing a shutdown.

The following conditions exist :

- Reactor is critical at 8% power.

- Annunciator Panel 110 Drop 8, INTMED RANGE COMPENSATE VOLT FAILURE alarms.

How does this affect the Nuclear Instrumentation?

A. N36 reading would immediately rise about 1 decade. The channel will be INOPERABLE for the P-6 Function once power is reduced.

B. N36 reading would immediately drop about 1 decade. The channel will be OPERABLE for the P-6 Function once power is reduced.

C. N36 reading would NOT immediately change. The channel will be OPERABLE for the P-6 Function once power is reduced.

D. N36 reading would NOT immediately change. The channel will be INOPERABLE for the P-6 Function once power is reduced.

Answer: D When the compensating voltage is lost, while in the power range the effect will not be noticed since the impact from the gamma radiation is a small percentage of the total current. At lower power levels, excessive gamma current will cause the IR detector to indicate excessively HIGH. This may result in 1 of 2 channels remaining above the P-6 setpoint, which would keep all P-6 blocking features active.

A - Incorrect - While in the power range the impact from gamma and the gamma compensation is a small percentage of the actual current. The channel will indicate high as power is reduced and so will not actuate P-6 at the correct time.

B - Incorrect - While in the power range the impact from gamma and the gamma compensation is a small percentage of the actual current. The channel will indicate high as power is reduced and so will not actuate P-6 at the correct time.

C - Incorrect - Both IR channels must be below P-6 to energize the SR. So the P-6 Function of the IR channel would be Inoperable.

REFERENCE:

RO-C-01300 Excore Nuclear Instrumentation System Handout #3, TS 3.3.1 and Bases.

LESSON PLAN/OBJ: RO-C-01300/#9 KA - 000033 2.4.46 Loss of Intermediate Range Nuclear Instrumentation Emergency Procedures/Plan Ability to verify that the alarms are consistent with the plant conditions.

RO/SRO Value - (3.5 / 3.6) CFR - 43.5 / 45.3 / 45.12 Question #79 KA# - 000033 2.4.46 Exam Level - SRO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 96

80. 080 20 The following alarms are received on Unit 2:

Steam Gen 1/2/3/4 Steam Line Flow High Alarms Steam Gen 1/2/3/4 SF>FWF Flow Mismatch Alarms The following conditions exist:

- RCS Tavg is 561oF and lowering

- Turbine load is lowering

- Rods are stepping out

- Steam flows are - 3.6 x 106 lbm/hr

- FW flows are - 2.1 x 106 lbm/hr Which one of the following correctly describes the cause and required action to be taken for the above conditions?

A. A steam line break exists. Direct the operators to perform a Reactor Trip and Main Steamline Isolation.

B. A feed line break exists. Direct the operators to perform a Reactor Trip and Main Feedwater Isolation.

C. Feedwater Pump delta P is too Low. Direct the operator to raise FW Pump Speed and FW pump flow.

D. MPC-253 has failed LOW. Direct the operators to perform actions for failed First Stage Turbine Impulse Pressure Transmitter.

Answer: A Based on the conditions presented a steam line break has occurred. Steam flow is indicating at the 97 to 98% power range. Tavg is 13oF Low for 98% power. A reactor trip and Steam Line isolation is warranted.

B - Incorrect - If A FW break existed RCS temperature would be rising.

C - Incorrect - If FW Flow was low RCS Temperature would be rising.

D - Incorrect - If MPC-253 failed low the alarms would come in (Steam flow higher than calculated power) but rods would step out and SF/FWF mismatch would not be this high.

Changed stem Steam Flow. Added FW flow to Stem. added SF>FWF Alarm.

Changed distractors B & C to include FW flow. Changed all distractors to include Operator actions.

REFERENCE:

RO-C-05103 Main Steam Systems pg. 9 SD-01100 RPS/ESFAS Signals System Description pg. 56 LESSON PLAN/OBJ: RO-C-05103/#9 KA - 000040 AA2.01 Steam Line Rupture Ability to determine and interpret the following as they apply to the Steam Line Rupture:

Occurrence and location of a steam line rupture from pressure and flow indications RO/SRO Value - (4.2 / 4.7) CFR - 43.5 / 45.13 SOURCE: INPO # 27707 Cook 1 - 4/29/2004(RO#073/SRO#NA)

Original Quest. KA - 000040 AA2.02

DC Cook 2006 NRC Exam 97 Steam Line Rupture Ability to determine and interpret the following as they apply to the Steam Line Rupture:

Conditions requiring a reactor trip RO-4.6 SRO-4.7 Question #80 KA# - 000040 AA2.01 Exam Level - SRO Question Source - MODIFIED - COOK 2004 #73 Cognitive/Difficulty Level - H/4

DC Cook 2006 NRC Exam 98

81. 081 3 The following plant conditions exist on Unit 2:

The Plant is in Mode 3 preparing for Reactor Startup.

Gamma Metrics Detector N21 is Inoperable.

A Loss of CRID III has occurred and operators have taken actions to stabilize the plant.

What ONE of the following actions are required for the Neutron Flux Function of Technical Specification 3.3.3, Post Accident Monitoring Instrumentation based on these events?

(Technical Specification 3.3.3, Post Accident Monitoring Instrumentation attached)

A. Restore Either Neutron Flux Detector to Operable status immediately or be in Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. Restore Either Neutron Flux Detector to Operable status within 7 Days from the Loss of N23.

C. Restore Neutron Flux Detector N21 to Operable status within 30 Days from the Loss of N21. N23 was not impacted by the loss of CRID III.

D. Restore Both Neutron Flux Detectors to Operable status within 30 Days from the Loss of N21.

Answer: C CRID III supplies power to the Unit 1 - N23. TS 3.3.3 requires 2 channels operable for the Neutron Flux Function. Loss of one channel allows 30 Days for repair. A loss of 2 Channels requires that 1 be restored within 7 Days.

A - Incorrect - This is the Action required by Condition G which is referenced in the Table if the other actions are not completed within the allowable time.

B - Incorrect - The N23 channels on both Units are powered from the Opposite Unit. The N21 Channels are powered from the respective unit.

D - Incorrect - This is the Action required for ONE Channel. Plausible if the candidate confused function with channel (Action A).

REFERENCE:

Technical Specification 3.3.3 Post Accident Monitoring Instrumentation LESSON PLAN/OBJ: RO-C-01300/#21 Attachment Provided - Technical Specification 3.3.3 Post Accident Monitoring Instrumentation pg. 3.3.3-1 to 3.3.3-5 KA - 000057 AA2.08 Loss of Vital AC Electrical Instrument Bus Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

Reactor power digital display and remote flux meter RO/SRO Value - (3.4 / 3.5) CFR - 43.5 / 45.13 Question #81 KA# - 000057 AA2.08 Exam Level - SRO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 99

82. 082 1 02-OHP-4022-064-001, Control Air Malfunction, directs the operator to Check 100 PSIG Control Air Header Pressure on XPI-100.

Which ONE of the following should the SRO direct with respect to 02-OHP-4022-064-001, if air pressure cannot be maintained?

A. If the plant is experiencing symptoms of a loss of Control Air AND system pressure reaches 95 psig then close the Plant Air Header Unit Crosstie valves. If pressure is still lowering then trip the reactor and go to 02-OHP-4023-E-0, Reactor Trip or Safety Injection.

B. If the plant is experiencing symptoms of a loss of Control Air AND system pressure reaches 80 psig and is still lowering, trip the reactor and go to 02-OHP-4023-E-0, Reactor Trip or Safety Injection.

C. If the plant is experiencing symptoms of a loss of Control Air AND system pressure reaches 90 psig and is still lowering, then close the Plant Air Header Unit Crosstie valves and start both Control Air compressors.

D. If the plant is experiencing symptoms of a loss of Control Air AND system pressure reaches 80 psig and is still lowering, then close the Containment Air Header Isolation valves and start a unit shutdown.

Answer: B 02-OHP-4022-064-001, Control Air Malfunction directs a reactor trip if Control Air Pressure is <80 psig.

A - Incorrect - The plant is not tripped until pressure is <80 psig. The PAC will auto start at 95 psig.

C - Incorrect - The crossties are closed at <85 psig. The Control Air compressors will start at 90 psig.

D - Incorrect - The plant should be tripped when pressure reaches <80 psig.

Containment is isolated if it appears to be the source of the leak.

REFERENCE:

02-OHP-4022-064-001, Control Air Malfunction Step 1 LESSON PLAN/OBJ: RO-C-06401/#8 KA - 000065 AA2.06 Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

When to trip reactor if instrument air pressure is decreasing RO/SRO Value - (3.6 / 4.2) CFR - 43.5 / 45.13 SOURCE: INPO # 26557 Turkey Point 3 - 12/15/2003 Original Quest. KA - 065AG2.1.6 Question #82 KA# - 000065 AA2.06 Exam Level - SRO Question Source - DIRECT-INPO - TURKEY PT 2003-26557 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 100

[ NOTE: This question was deleted from the examination. ]

83. 083 3 The plant is at 95% power.

- 1200 on April 3, the 1S SI pump is declared inoperable.

- 1430 on April 4, the 1W Centrifugal Charging pump is declared inoperable.

- 1430 on April 5, the 1S SI pump is restored to OPERABLE status.

Including any extensions that are permitted by TS, which one of the following describes the LATEST time and date to restore the 1W Centrifugal Charging pump to OPERABLE status without requiring a unit shutdown?

A. 1200 on April 6 B. 1200 on April 7 C. 1430 on April 7 D. 1430 on April 8 Answer: Deleted Since the S SI pump and W Centrifugal Charging pump are associated with the same ECCS train, the inoperability of the W Centrifugal Charging pump does NOT constitute a subsequent failure expressed in the Condition. Condition A specifies one ECCS train.

The subsequent inoperability above is in the same train. Therefore, a Completion Time extension is NOT allowed.

B - Incorrect - The LCO is for the ECCS train with a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the first failure.

C - Incorrect - The LCO is for the ECCS train with a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the first failure.

D - Incorrect - The LCO is for the ECCS train with a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the first failure.

REFERENCE:

TS 3.5.2, Condition A LESSON PLAN/OBJ: RO-C-TS01/#11 KA - 006000 2.2.24 Emergency Core Cooling System (ECCS)

Equipment Control Ability to analyze the affect of maintenance activities on LCO status.

RO/SRO Value - (2.6 / 3.8) CFR - 43.2 / 45.13 SOURCE: ITS Exam Question #89 Question #83 KA# - 006000 2.2.24 Exam Level - SRO Question Source - DIRECT - ITS #89 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 101

84. 084 11 The following plant conditions exist on Unit 2:

- The East CCW HX is in service with the West CCW Pump running.

- CCW Surge Tank level is stable.

- CRS-4301, East CCW HX Radiation Monitor, generates an External Failure Alarm due to a faulty flow switch Which ONE of the following describes the response of the CCW system and the required actions, if any, for this condition?

A. No automatic actions will occur since the West CCW pump is running. No Lineup changes are required, operation in this condition is allowed indefinitely. The CCW system remains operable.

B. No automatic actions will occur since the CRS-4401, West CCW HX Radiation Monitor is still functioning. The West CCW HX must be aligned so the 2-CRV-412 Vent Valve will automatically close on a high radiation signal. The East CCW HX must be declared Inoperable.

C. 2-CRV-412, CCW Surge Tank Vent Valve, will automatically close. The West CCW HX must be aligned so the 2-CRV-412 Vent Valve may be reopened. The East CCW HX must be declared Inoperable.

D. 2-CRV-412, CCW Surge Tank Vent Valve, will automatically close. No Lineup changes are required, operation in this condition is allowed indefinitely. The CCW system remains operable.

Answer: D 2-CRV-412, CCW Surge Tank Vent Shutoff Valve closes on High Rad Level Alarm, Low Sample Flow, External Failure on CRS-4300/4400, Channel 4301 - East and/or Channel 4401-West. The automatic action is the closure of the Vent valve. The closure of the vent and monitor failure don't impact system operability, so operation may continue with this lineup.

A - Incorrect - Plausible since the East CCW HX monitor may be associated with the East pump but either radiation monitor will cause the surge tank vent to isolate.

B - Incorrect - Plausible since the east monitor has failed and the west is still operational but an alarm or failure of either will cause the Surge tank vent to isolate.

C - Incorrect - Plausible since the Surge tank vent is isolated. Alignment to the West HX will not allow the vent to be reopened.

Changed stem to ask required actions. Included actions & operability in distracters and changed distractor C to include Automatic vent closure.

REFERENCE:

12-OHP-4024-139 #26 LESSON PLAN/OBJ:RO-C-01350/#4 KA - 008000 A2.04 Component Cooling Water System (CCWS)

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

PRMS alarm RO/SRO Value - (3.3 / 3.5) CFR - 41.5 / 43.5 / 45.3 / 45.13 SOURCE: Audit RO23-074-3 (#068)

Original Quest. KA - SYS 073 A1.01 (3.2/3.5)

DC Cook 2006 NRC Exam 102 Process or Effluent High Radiation

-Ability to predict and/or monitor changes in parameters associated with operating the PRM system controls including radiation levels.

CRS-4300/4400 Channel 4301 - East Channel 4401-West Rad Monitor Flow Switch has high and low setpoints which cause an external failure when reached. 2-CRV-412, CCW Surge Tank Vent Shutoff Valve closes on High Rad Level Alarm, Low Sample Flow, External Failure.

Question #84 KA# - 008000 A2.04 Exam Level - SRO Question Source - MODIFIED - RO23-074-3 (#068)

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 103

85. 085 10 The control room operators are conducting a cooldown following a LOCA.

The following plant conditions exist:

- Core Exit TC's are 410oF.

- RCS Pressure is 260 psig.

- RVLIS Narrow Range is 36%.

- RVLIS Wide Range is 17%.

- RCP #23 is the only RCP Operating.

Which ONE of the following describes current core conditions and operational requirements?(02-OHP-4023-F-0.2, Core Cooling status tree attached)

A. Saturated. The operators are required to immediately enter 02-OHP-4023-FR-C.3, Response to Saturated Core Cooling to restore subcooled core cooling.

B. Saturated. At their discretion, the operators may perform 02-OHP-4023-FR-C.3, Response to Saturated Core Cooling to restore subcooled core cooling.

C. Degraded. The operators are required to immediately enter 02-OHP-4023-FR-C.2, Response to Degraded Core Cooling to prevent conditions from degrading to an inadequate core cooling condition.

D. Degraded. Prompt action must be taken to trip the RCP and enter 02-OHP-4023-FR-C.2, Response to Degraded Core Cooling or conditions could degrade to an inadequate core cooling condition.

Answer: B 260 psig = 274.7 psia = 409oF indicating that Subcooling is <36oF. With 1 RCP running and WR RVLIS > 15% the correct procedure would be 02-OHP-4023-FR-C.3, Response to Saturated Core Cooling. This is a yellow path procedure so discretion is allowed.

A - Incorrect - This is a yellow path procedure so discretion is allowed.

C - Incorrect - Temperature is low enough and there is enough inventory that a degraded condition does not exist.

D - Incorrect - Temperature is low enough and there is enough inventory that a degraded condition does not exist. If the RCP were tripped with these conditions FR-C.2 may be required.

Changed stem to include operating RCP. Changed RVLIS and temperatures. Changed distractors A & D from Subcooled & Inadequate to balance distracters.

REFERENCE:

02-OHP-4023-F-0.2, Critical Safety Functions Status Trees, Core Cooling LESSON PLAN/OBJ: RO-C-EOP10/#21 Attachment Provided - 02-OHP-4023-F-0.2, Core Cooling status tree KA - 00WE07 EA2.1 Saturated Core Cooling Ability to determine and interpret the following as they apply to the Saturated Core Cooling:

Facility conditions and selection of appropriate procedures during abnormal and emergency operations RO/SRO Value - (3.2 / 4.0) CFR - 43.5 / 45.13 SOURCE: INPO # 27714 Cook 1 - 4/29/2004 (RO#080/SRO# N/A)

Modified from Bank 12EOPC1003-1

DC Cook 2006 NRC Exam 104 Original Quest. KA - 00WE07 EA2.1 Saturated Core Cooling Ability to determine and interpret the following as they apply to the Saturated Core Cooling:

Facility conditions and selection of appropriate procedures during abnormal and emergency operations RO-3.2 SRO-4.0 Question #85 KA# - 00WE07 EA2.1 Exam Level - SRO Question Source - MODIFIED - COOK 2004 #80 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 105

86. 086 4 Unit 1 is responding to a LOCA.

Upon entering 01-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation, the following conditions existed:

- 1W RHR pump - tagged out

- 1E RHR stopped when 1-ICM-305, Recirc Sump to East RHR/CTS Pumps, could NOT be opened

- Containment pressure - 11 psig

- RCS pressure - 1200 psig

- RWST level - 16%

- Both SI pumps are shutdown, but available

- Both CCPs are running

- Both CTS Pumps are running

- No RCPs are running The crew has now reached step 12, Check if an RCP should be started.

Which ONE of the following list the pump(s) that are required to be running at this point in the procedure? (01-OHP-4023-ECA 1-1, Steps 1-12 are attached.)

A. An CCP only.

B. An CCP and an CTS pump only.

C. An CCP, an SI pump and one CTS pump only.

D. An CCP, an SI pump and both CTS pumps.

Answer: C One CTS pump is stopped at step 5 when according to the table, one CTS pump is required. One CCP is stopped in step 11a. One SI pump is started at step 11b.

A - Incorrect - Candidate believes both CTS pumps are stopped and misses the SI pump start.

B - Incorrect - Candidate misses SI pump start.

D - Incorrect - Candidate misses step to stop CTS pump.

REFERENCE:

01-OHP-4023-ECA-1-1 LESSON PLAN/OBJ: RO-C-EOP09/#45 Attachment Provided - 01-OHP-4023-ECA 1-1, Steps 1-12.

KA - 00WE11 EA2.2 Loss of Emergency Coolant Recirculation Ability to determine and interpret the following as they apply to the Loss of Emergency Coolant Recirculation:

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments RO/SRO Value - (3.4 / 4.2) CFR - 43.5 / 45.13 SOURCE: AUDIT RO22 Q#25 Original Quest. KA - WE 11 G2.1.7 Question #86

DC Cook 2006 NRC Exam 106 KA# - 00WE11 EA2.2 Exam Level - SRO Question Source - DIRECT - AUDIT RO22 Q#25 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 107

87. 087 4 nit 1 reactor has been manually tripped due to a secondary system malfunction.

01-OHP-4023-E-0 has been performed and a transition made to 01-OHP-4023-ES-0.1, Reactor Trip Response. The STA has identified a YELLOW path on the Heat Sink Status Tree for steam generator pressure.

The crew has entered 01-OHP-4023-FR-H.2, Response to Steam Generator Overpressure.

The following conditions exist:

- Steam Generator #13 Pressure - 1100 psig

- Steam Generator #13 NR Level - 96%

What is the appropriate action relative to steam release from S/G #13?

A. Steam release should NOT occur because it may result in excessive RCS cooldown and depressurization, potentially causing a SI.

B. Steam release should NOT occur because it may result in two phase flow and water hammer, potentially damaging pipes and valves.

C. Steam may be released via the SG PORV ONLY since narrow range level is greater than 67%.

D. Steam may be released without restriction since narrow range level has been adequately established.

Answer: B Per 01-OHP-4023-FR-H.2 Step 3, if the SG level is >92% a transition is made to 01-OHP-4023-FR-H.3 to address the high level that may be causing the pressure concern.

Also as discussed in 01-OHP-4023-FR-H.3 Step 1, with a high SG level steam should not be released until the steam lines can be evaluated.

A - Incorrect - The higher level in the SG should have little effect on the RCS pressure drop. An SI would not be expected from a single SG steam release.

C - Incorrect - Steam release through the SG PORV could still lead to water hammer.

D - Incorrect - At this high of level Steam should not be released.

Modified stem to remove statement that a transition to H.3 is required and changed question to ask correct action vs. reason for transition.

changed distractors to include Steam Release or NOT and changed reasons for A, C, &

D.

REFERENCE:

01-OHP-4023-FR-H.2, Response to Steam Generator Overpressure pg.

2 01-OHP-4023-FR-H.3, Response to Steam Generator High Level pg. 2 PSBD 12-OHP-4023-FR-H.2 Background Document pg. 7 PSBD 12-OHP-4023-FR-H.3 Background Document pg. 5-6 LESSON PLAN/OBJ:RO-C-EOP11/#10 KA - 00WE13 2.1.7 Steam Generator Overpressure Conduct of Operations Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

RO/SRO Value - (3.7 / 4.4) CFR - 43.5 / 45.12 / 45.13

DC Cook 2006 NRC Exam 108 SOURCE: Cook 2004 Question #82 From INPO 21508-BRAIDWOOD02 Original Quest. KA - 00WE13 EK3.4 Steam Generator Overpressure Knowledge of the reasons for the following responses as they apply to the Steam Generator Overpressure:

RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated RO-3.1 SRO-3.3 Question #87 KA# - 00WE13 2.1.7 Exam Level - SRO Question Source - MODIFIED - COOK 2004 #82 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 109

88. 088 3 The following conditions exist :

-Large Break LOCA is in progress

-Containment pressure is 1.9 psig and stable

-You notice an ORANGE condition on CONTAINMENT CSF ST due to the "FLOOD LEVEL" lights being Lit.

What action will be directed by OHP-4023-FR-FR Z.2, Response to Containment Flooding, and what is the concern if these actions are not successful?

A. Divert RHR flow from the Containment Sump to the RWST to lower Containment Level.

High water levels could result in critical components needed for plant recovery being damaged and rendered inoperable.

B. Identify and isolate the source of excess water using control board indications and Containment Sump samples. Water levels could reach the bottom of the reactor vessel resulting in thermal shock and vessel failure.

C. Identify and isolate the source of excess water using control board indications and Containment Sump samples. High water levels could result in critical components needed for plant recovery being damaged and rendered inoperable.

D. Stop both containment spray pumps. Water levels could reach the bottom of the reactor vessel resulting in thermal shock and vessel failure.

Answer: C Containment design basis flood level takes into account the entire water contents of the RCS, RWST, Ice condenser ice bed melt, and SI accumulators, plus the added mass of a LOCA and a steam line or feedline break inside containment. NESW and CCW may be major contributors to exceeding "flood" level and causing a loss of equipment required for long term cooling.

A - Incorrect - Water is not pumped out of containment using the RHR pumps.

B - Incorrect - Water reaching the Reactor vessel would not cause thermal shock or vessel failure.

D - Incorrect - The CTS pumps are not stopped due to high level. Water reaching the Reactor vessel would not cause thermal shock or vessel failure.

REFERENCE:

12-OHP-4023-FR-Z-2, Response to Containment Flooding Background; RO-C-EOP13, Containment CSFST, FR-Z Series EOPs and Background Information LESSON PLAN/OBJ: RO-C-EOP13/#6 KA - 00WE15 EA2.2 Containment Flooding Ability to determine and interpret the following as they apply to the Containment Flooding:

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments RO/SRO Value - (2.9 / 3.3) CFR - 43.5 / 45.13 SOURCE: INPO # 27561 Prairie Island - 4/23/2004 Original Quest. KA - E15 EK1.3 Question #88 KA# - 00WE15 EA2.2 Exam Level - SRO

DC Cook 2006 NRC Exam 110 Question Source - DIRECT-INPO - PRAIRIEIS-2004-27561 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 111

89. 089 4 Unit 1 is in MODE 3 preparing for a reactor startup. All shutdown bank rods are withdrawn.

I&C informs you that MPC-254 (Main Turbine 1st stage pressure transmitter) must be placed in the test (failed high) position in order to perform work on the transmitter.

Which ONE of the following correctly describes your response to this request and the reason?

A. Allow Testing. MPC-253 is still indicating properly, so P-13 and P-7 will indicate correctly.

B. Do NOT Allow Testing. The reactor trips associated with permissive P-7 would be unblocked.

C. Do NOT Allow Testing. The Steam Dumps will close when MPC-254 is placed in the Failed Condition.

D. Allow Testing, after placing AMSAC to the Bypass Condition.

Answer: B P-13 is actuated by 2 of 2 Impulse channels below 10. Failing 1 channel high will remove P-13 which will remove the P-7 Blocks.

A - Incorrect - 2/2 channels below 10% are required for P-13 & P-7.

C - Incorrect - Steam dumps are in steam Pressure Mode during these conditions. The dumps would close if in Tave Mode.

D - Incorrect - AMSAC requires 2/2 Channels >40% to enable.

Changed Stem from Why NOT place in Trip to ask what is correct response (Allow or NOT). Changed answers to include Allow of NOT, A & C - new distractors. Changed D to allow after AMSAC in Bypass (Like original C Distractor)

REFERENCE:

OHP 4022.013.006 LESSON PLAN/OBJ: RO-C-NS11/#21; RQ-C-1543/#1; 4b RQ-C-2323/#15; RO-C-01100/#4c KA - 016000 2.2.18 Non-Nuclear Instrumentation System (NNIS)

Equipment Control Knowledge of the process for managing maintenance activities during shutdown RO/SRO Value - (2.3 / 3.6) CFR - 43.5 / 45.13 Original Question Source - Master Bank 01011C0004-3 Original KA - 016 A2.01 (3.0/3.1)

EXAM/QUIZZES: R911916; RO1822; Q2302C; RO1822MU2; R2324V-A1B; 2904Q; R2324C-A4B; STA EXAM 5; RO22SYSREVIEW; RO22SYSREVA; RQ2705V1;RQ2705A; RQ2705D; RO23SYSCOMP; RO23EOP3; RQ2907B Question #89 KA# - 016000 2.2.18 Exam Level - SRO Question Source - DIRECT - 01011C0004-3 Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 112

90. 090 11 You are the Unit Supervisor responding to a Unit 2 LOCA and are currently implementing 02-OHP-4023-FR-Z.1, Response to High Containment Pressure. Step 6 of this procedure requires you to place DIS in Service.

Which ONE of the following correctly describes required actions to place the Distributed Ignition System (DIS) in operation?

Direct the RO to turn on the Hydrogen Igniters after verifying that the...

A. AEO has locally stopped the Ice Condenser Air Handling Units.

B. AEO has locally started the Ice Condenser Air Handling Units.

C. Ice Condenser Air Handling Units have automatically tripped.

D. Ice Condenser Air Handling Units have automatically started.

Answer: A The Ice Condenser AHUs are locally stopped when DIS is placed in service to reduce potential ignition sources (defrost units). The Hydrogen Igniters are started in the MCR.

B - Incorrect - The AHUs are shutdown prior to turning on the igniters.

C - Incorrect - The AHUs must be locally shutdown, they do not automatically trip.

D - Incorrect - The AHUs must be locally shutdown, they do not automatically start.

REFERENCE:

RO-C-01000, Ice Condenser System LESSON PLAN/OBJ: RO-C-01000 / #12 KA - 025000 2.4.35 Ice Condenser System Emergency Procedures/Plan Knowledge of local auxiliary operator tasks during emergency operations including system geography and system implications.

RO/SRO Value - (3.3 / 3.5) CFR - 43.5 / 45.13 Question #90 KA# - 025000 2.4.35 Exam Level - SRO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 113

91. 091 3 Unit 1 is in Mode 6.

The following conditions exist:

- Fuel Handling Area Ventilation System is aligned with flow through the Charcoal Filter

- Fuel Handling Building Radiation Monitor, R-5 alarm/actuation circuitry has just failed. - VRS-5000 Fuel Handling Area Monitor is available

- MTI has been called for troubleshooting the R-5 failure.

Which of the following describes the required ACTION, if any, to be taken in order to allow core alterations and fuel movement to continue?

A. Core Alterations and Fuel Movement in the Auxiliary Building must be stopped until the R-5 monitor is repaired.

B. Core Alterations and Fuel Movement in the Auxiliary Building may continue uninterrupted, since the Fuel Handling Area Ventilation System is already aligned with flow through the Charcoal Filter.

C. Core Alterations and Fuel Movement in the Auxiliary Building may continue uninterrupted, since the VRS-5000 Fuel Handling Area Monitor is available.

D. Core Alterations may continue but Fuel Movement in the Auxiliary Building must be stopped until the R-5 monitor is repaired.

Answer: D The Fuel Handling Ventilation System must be operable to conduct Fuel Movement in the Auxiliary Building. Upon receipt of a Fuel Handling Area Radiation - High signal the fuel handling area supply fans are tripped, thus ensuring a negative pressure within the space. The charcoal filter section bypass dampers also receive a close signal upon receipt of Fuel Handling Area Radiation - High signal (however, these dampers are maintained closed when the required FHAEV train is in operation). Thus failure of R-5 alarm/actuation circuitry makes the FHAEV inoperable (Supply fans won't trip - SR 3.7.13.4)

A - Incorrect - Core alterations (In Containment) may continue.

B - Incorrect - The Supply fans will not trip.

C - Incorrect - The VRS-5000 provides indications only so the Supply fans will not trip.

REFERENCE:

Technical Specifications 3.7.13 and bases, 01-OHP-4030-127-037 Data Sheet 3 Step 7.

LESSON PLAN/OBJ: RO-C-ADM13/#ADM13.3 KA - 034000 A4.01 Fuel Handling Equipment System (FHES)

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

Radiation levels RO/SRO Value - (3.3 / 3.7) CFR - 41.7 / 45.5 to 45.8 Question #91 KA# - 034000 A4.01 Exam Level - SRO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 114

92. 092 1 You are the Unit Supervisor. Unit 2 is at 100% power.

Panel 215 Drop 48 - BATTERY N UNDERVOLTAGE has just alarmed.

Investigation revealed that a metal plate has shorted the battery terminals.

Which ONE of the following identifies the effects on the operability and capability of the Auxiliary Feedwater System?

A. The TDAFW Pump will NOT start and the FMO-211, 221, 231, & 241 TDAFW to SG Isolation valves are failed in the open position. Declare the TDAFW train inoperable.

B. The TDAFW Pump will start and the FMO-211, 221, 231, & 241 TDAFW to SG Isolation valves are failed in the open position. Declare the N Train battery inoperable.

C. The TDAFW Pump will start but the MCM-221 SG Steam supply to TDAFW Pump Isolation valve is failed in the closed position. Declare the TDAFW Pump inoperable.

D. The TDAFW Pump will NOT start and the FMO-211, 221, 231, & 241 TDAFW to SG Isolation valves are failed in the closed position. Declare the TDAFW train inoperable.

Answer: A The Train N battery supplies power to the TDAFW pump start circuitry, Trip & Throttle Valve, SG FW Valves and Test Valves. The TDAFW pump valves are normally open and so they will fail in the open position. TS 3.7.5 Condition B allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B - Incorrect - The TDAFW Pump will not start. If Train N battery is inoperable the TDAFW pump must be declared inoperable.

C - Incorrect - The TDAFW pump will not start and SG Steam supply valves are AC powered.

D - Incorrect - The valves are normally open and so fail open.

REFERENCE:

RO-C-05600 Auxiliary Feedwater System pg. 24, TS 3.7.5 AFW & 3.8.4 DC-Operating LESSON PLAN/OBJ: RO-C-05600/#4 KA - 061000 A2.03 Auxiliary / Emergency Feedwater (AFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the AFW System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of dc power RO/SRO Value - (3.1 / 3.4) CFR - 41.5 / 43.5 / 45.3 / 45.13 Question #92 KA# - 061000 A2.03 Exam Level - SRO Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 115

93. 093 2 Unit 1 is 100% power and Unit 2 is in Mode 3.

All Unit 1 CW pumps are operating.

Only 2 Unit 2 CW pumps are operating.

The following conditions exist :

- Ann. 223, Drop 44, Travelling Screen DP High - LIT

- Ann. 223, Drop 45, Travelling Screen DP High-High - LIT

- Screen DP is 43".

- Lake Inlet temperature is 75oF Which ONE of the following describes the required actions and the plant impact if these actions are not taken? (12-OHP-4022-057-001 Attached)

A. Reduce Power to <78% and Shutdown 2 Unit 1 CW pumps. Continued operation at high DP could cause loss of screens and subsequent clogging of the Main & FW Pump condensers.

B. Reduce Power to <91% and Shutdown 1 Unit 1 CW pump. Continued operation at high DP could cause loss of screens and subsequent failure of the Essential Service Water System.

C. Immediately Trip Unit 1 and Shutdown All CW pumps on Both units. Continued operation at high DP could cause loss of screens and subsequent failure of the Essential Service Water System.

D. Immediately Trip both Unit 2 CW pumps. Continued operation at high DP could cause loss of screens and subsequent clogging of the Main & FW Pump condensers.

Answer: B 12-OHP-4022-057-001 requires that power be reduced to < 91 % on unit 1 and 1 CW pump stopped. (Step 5 & Attachment A)

A - Incorrect - This would be the required action if Unit 2 was at 100% with 4 CW pumps running. (Unit 1 has only 3 pumps) (Attachment B)

C - Incorrect - This action is not required unless level is > 50" or until both units have been reduced to only 2 CW pumps running. (Foldout Page)

D - Incorrect - This action is not procedurally directed. In Mode 3, Stopping all CW pumps would lead to a loss of vacuum.

REFERENCE:

12-OHP-4022-057,Screen House Degraded Forebay Condition LESSON PLAN/OBJ: RO-C-AOP11/#AOP11.16 Attachment Provided OHP-4022-057,Screen House Degraded Forebay Condition KA - 075000 A2.02 Circulating Water System Ability to (a) predict the impacts of the following malfunctions or operations on the Circulating Water System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of circulating water pumps RO/SRO Value - (2.5 / 2.7) CFR - 41.5 / 43.5 / 45.3 / 45.13 Question #93 KA# - 075000 A2.02 Exam Level - SRO

DC Cook 2006 NRC Exam 116 Question Source - NEW -

Cognitive/Difficulty Level - H/3

DC Cook 2006 NRC Exam 117

94. 094 12 The following conditions exist :

- Unit 2 is in Mode 1.

- A Reactor Operator and the Unit Supervisor are in the Unit 2 Control Room.

- A high vibration alarm is received on the HD pump requiring someone to go behind the panel to check the indications.

Which ONE of the following describes the procedurally accepted method of checking the indications?

A. The Unit Supervisor can go behind the panel to check the vibration.

B. The Reactor Operator can go behind the panel to check the vibration.

C. Both the Reactor Operator and the Unit Supervisor are allowed to go behind the panel to check the vibration as long as all controls are in automatic.

D. Neither the Reactor Operator or the Unit Supervisor can go behind the panels. They must get another operator to check the vibration.

Answer: A The Unit Supervisor must be in the Control Room but may go behind the panels. The RO must remain in the view of the panels.

B - Incorrect - The RO must remain in the view of the panels.

C - Incorrect - The RO must remain in the view of the panels.

D - Incorrect - The SRO may go behind the panels.

REFERENCE:

OHI-4000, Conduct of Operations Attachment 23 (Shift Staffing)

LESSON PLAN/OBJ: RO-C-ADM01/#1 KA - Generic 2.1.4 Generic Conduct of Operations Knowledge of shift staffing requirements.

RO/SRO Value - (2.3 / 3.4) CFR - 41.10 / 43.2 SOURCE: AUDIT RO22-BOTH-25 Original Quest. KA - 2.1.4 Question #94 KA# - GENERIC 2.1.4 Exam Level - SRO Question Source - DIRECT - AUDIT RO22-BOTH-25 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 118 NOTE:

Question 95 contained security-related information that was not publicly available.

DC Cook 2006 NRC Exam 119

96. 096 3 Which ONE of the following describes a responsibility of the SRO when reviewing an On-The-Spot Change (OTSC)?

A. Verifies that the change is consistent with appropriate source documentation.

B. Determines that the change is technically correct.

C. Ensures that the change meets equipment engineering requirements.

D. Ensures that the change does not adversely affect plant safety or operation.

Answer: D The SRO is responsible for ensuring that the change doesn't affect safety or plant operation.

A - Incorrect - This is not a function of the SRO review.

B - Incorrect - This is covered by the QTR signoff.

C - Incorrect - This is not a function of the SRO review.

REFERENCE:

PMP-2010-PRC-003 LESSON PLAN/OBJ: RO-C-ADM12/#3.3 KA - Generic 2.2.6 Generic Equipment Control Knowledge of the process for making changes in procedures as described in the safety analysis report.

RO/SRO Value - (2.3 / 3.3) CFR - 43.3 / 45.13 SOURCE: Direct 01ADMC12-2 Original Quest. KA - G2.2.6 EXAM/QUIZZES: RO23ADM1 Question #96 KA# - GENERIC 2.2.6 Exam Level - SRO Question Source - DIRECT - 01ADMC12-2 Cognitive/Difficulty Level - F/2

DC Cook 2006 NRC Exam 120

97. 097 2 Unit 2 is in Mode 1 at 100% power. Maintenance activities require that the North SI pump discharge cross-tie valve must be CLOSED for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

What actions must be taken to ensure Technical Specifications AND LOCA analysis requirements protection is provided?

A. Reduce thermal Power to < 3304 MWt.

No equipment alignment changes are required.

B. Open both RHR discharge x-tie valves, IMO-314 and IMO-324 Place the E RHR pump switch in the pull to lock position.

C. Open the E RHR discharge x-tie valve, IMO-314 Place the W RHR pump switch in the pull to lock position.

D. Reduce thermal Power to < 3304 MWt.

Open both RHR discharge x-tie valves, IMO-314 and IMO-324 Place the E RHR pump switch in the pull to lock position.

Answer: D Technical specifications require that both SI Discharge crosstie valves be open. If one valve must be closed then power must be reduced to < 3304 MWt within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

additionally the RHR Crossties must be opened to ensure 4 loop injection. The East RHR pump is placed in PTL to prevent the RHR pumps from overpowering each other.

A - Incorrect - The RHR crossties must be opened.

B - Incorrect - Power must be reduced on Unit 2 to < 3304MWt C - Incorrect - Both crosstie valves must be opened.

REFERENCE:

02-OHP 4021.008.002, ATT. 1, Technical Specification 3.5.2 Condition D and bases.

LESSON PLAN/OBJ: RO-C-00800/#13 KA - Generic 2.2.24 Generic Equipment Control Ability to analyze the affect of maintenance activities on LCO status.

RO/SRO Value - (2.6 / 3.8) CFR - 43.2 / 45.13 SOURCE: Master Bank 01008C0014-2 Original Quest. KA - 006 K6.02 EXAM/QUIZZES: RO1821; Q2703V; RQ2526C-R/S; RO22EOP5A Question #97 KA# - GENERIC 2.2.24 Exam Level - SRO Question Source - DIRECT - 01008C0014-2 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 121

98. 098 3 Rad monitor VRS-1505, Unit Vent Effluent Low Range Noble Gas Channel, has failed HIGH.

Repairs will take at least 3 days.

Which ONE of the following actions is required by PMP-6010-OSD-001, Off-site Dose Calculation Manual, regarding a release from the waste gas decay tank?

References Attached:

PMP-6010-OSD-001, Off-site Dose Calculation Manual, Attachment 3.4 (pages 50-52)

The release...

A. may NOT be started until the repairs on VRS-1505 are completed.

B. may be started provided RP recalculates the trip set points using VRS-2505 as the release path alarm and termination monitor.

C. may be started for up to 30 days provided grab samples are taken shiftly and analyzed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. may be started for up to 14 days provided 2 independent samples are analyzed and the calculations and valve lineups are dual verified.

Answer: D PMP-6010-OSD-001 Attachment 3.4 Action 9 allows releases to continue for 14 days provided 2 independent samples are analyzed and the calculations and valve lineups are dual verified.

A - Incorrect - Action 9 allows releases to continue.

B - Incorrect - Additional Actions are required to allow releases. The VRS-2505 monitors the Unit 2 vent stack.

C - Incorrect - Additional Actions are required to allow waste gas releases. These actions are for normal operation.

REFERENCE:

PMP-6010-OSD-001, Off-site Dose Calculation Manual, Attachment 3.4 (pages 50-52)

LESSON PLAN/OBJ: RO-C-ADM10/#ADM10.5 Attachment Provided - PMP-6010-OSD-001, Off-site Dose Calculation Manual,.4 (pages 50-52)

KA - Generic 2.3.6 Generic Radiological Controls Knowledge of the requirements for reviewing and approving release permits.

RO/SRO Value - (2.1 / 3.1) CFR - 43.4 / 45.10 SOURCE: Direct RO23-93-3 (RO# N/A /SRO#088)

From AUDIT02-SRO25 Original Quest. KA - APE 060 G 2.3.8 (2.3/3.2)

Question #98 KA# - GENERIC 2.3.6 Exam Level - SRO Question Source - DIRECT - AUDIT RO23 Q#88

DC Cook 2006 NRC Exam 122 Cognitive/Difficulty Level - H/4

DC Cook 2006 NRC Exam 123

99. 099 1 The maximum time that can elapse between a gas decay tank being approved for release and the start of the release is...

A. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Answer: C 12-OHP-4021-023-002 Release of Radioactive Waste from Gas Decay Tanks, Data Sheet 1 requires the SRO to ensure that not more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> have elapsed since the Release was approved.

A - Incorrect - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the required time for estimated sample flow if rad flow monitors are inoperable.

B - Incorrect - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the typical shiftly surveillance time.

D - Incorrect - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is the time allowed to reduce the GDT concentration if it exceeds 43,800 Ci.

REFERENCE:

12-OHP-4021-023-002 Release of Radioactive Waste from Gas Decay Tanks, Data Sheet 1 LESSON PLAN/OBJ: RO-C-ADM01/#21 KA - Generic 2.3.8 Generic Radiological Controls Knowledge of the process for performing a planned gaseous radioactive release.

RO/SRO Value - (2.3 / 3.2) CFR - 43.4 / 45.10 SOURCE: INPO # 27356 Ginna 1 - 4/27/2004 Original Quest. KA - 2.3.6 Question #99 KA# - GENERIC 2.3.8 Exam Level - SRO Question Source - DIRECT-INPO - GINNA 2004-27356 Cognitive/Difficulty Level - F/3

DC Cook 2006 NRC Exam 124 100. 100 2 Unit 1 was in MODE 4 with a Containment Inspection in progress.

An explosive device was found in close proximity to the Main Steam Leads.

How should the this event be classified per the Emergency Plan? (PMP-2080-EPP-101 is attached)

A. General Emergency B. Site Area Emergency C. Alert D. Unusual Event Answer: B Security event H-2 requires a Site Area Emergency if a bomb is found within the Vital Area.

A - Incorrect - This applies if the ability to reach Mode 5 is lost.

C - Incorrect - This applies to a security event in the protected area - Intrusion or civil disturbance.

D - Incorrect - This applies to an explosive device in the protected but not vital area.

Containment is a Vital area

REFERENCE:

PMP-2080-EPP-101 Attachment 1 pg. 14 & Attachment 3 pg. 48 LESSON PLAN/OBJ: ST-C-EP03/#2 Attachment Provided - PMP-2080-EPP-101 Attachment 1 NOTE - PMP -2080-EPP-101 should not be placed in Public record. NRC EAL Security Advisory.

KA - Generic 2.4.28 Generic Emergency Procedures/Plan Knowledge of procedures relating to emergency response to sabotage.

RO/SRO Value - (2.3 / 3.3) CFR - 41.10 / 43.5 / 45.13 SOURCE: INPO # 22510 Diablo Canyon 1 - 10/1/2002 Original Quest. KA - 007.2.4.28 Question #100 KA# - GENERIC 2.4.28 Exam Level - SRO Question Source - DIRECT-INPO - DIABLO CYN2002-22510 Cognitive/Difficulty Level - H/2

DC Cook 2006 NRC Exam 125

• Questions 1 through 75 are RO.

• Questions 76 through 100 are SRO.

1.B 2.B 3.C 4.B 5.C 6.B 7.B 8.C 9.A 10.B 11.C 12.D 13.D 14.B 15.A 16.C 17.A 18.A 19.A 20.D 21.D 22.C 23.C 24.D 25.A 26.B 27.D 28.A 29.D 30.D 31.B 32.D 33.A 34.B 35.A 36.C 37.D 38.D 39.B 40.C 41.D 42.A 43.B 44.B 45.A 46.B 47.B 48.A 49.A 50.

B and D 51.D 52.A 53.C 54.C 55.C 56.C 57.C 58.C 59.C 60.B 61.B 62.B 63.A 64.B 65.A 66.B 67.C 68.D 69.B 70.B 71.C 72.B 73.C 74.B 75.C 76.D 77.A 78.C 79.D 80.A 81.C 82.B 83.

Deleted 84.D 85.B 86.C 87.B 88.C 89.B 90.A 91.D 92.A 93.B 94.A 95.D 96.D 97.D 98.D 99.C 100.B