ML12352A085
ML12352A085 | |
Person / Time | |
---|---|
Site: | Diablo Canyon ![]() |
Issue date: | 12/17/2012 |
From: | Vincent Gaddy NRC Region 4 |
To: | Pacific Gas & Electric Co |
laura hurley | |
References | |
Download: ML12352A085 (134) | |
Text
U.S. Nuclear Regulatory Commission Diablo Canyon RO Written Examination Applicant Information Name: KEY Date: 21 November, 2012 Facility/Unit: Diablo Canyon Region:
I II III IV Reactor Type: W CE BW GE Start Time:
Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicants Signature Results Examination Value __________ Points Applicants Score
__________ Points Applicants Grade
__________ Percent
U.S. Nuclear Regulatory Commission Diablo Canyon SRO Written Examination Applicant Information Name: KEY Date: 21 November, 2012 Facility/Unit: Diablo Canyon Region:
I II III IV Reactor Type: W CE BW GE Start Time:
Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets.
To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicants Signature Results RO/SRO-Only/Total Examination Values
/ / Points Applicants Scores
/ / Points Applicants Grade
/ / Percent
DCPP L111 NRC Exam 21 November, 2012 Page i Multiple Choice (Fill in your choice)
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DCPP L111 NRC Exam 21 November, 2012 Page ii Multiple Choice (Fill in your choice)
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DCPP L111 NRC Exam 21 November, 2012 P a g e l 1 Examination Outline Cross-Reference Level RO Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: Containment isolation valves affecting RCP operation Tier #
2 Group #
1 K/A #
003 K6.04 Rating 2.8 Question 1 Unit 1 is at 100% and a Phase A containment isolation occurs.
The Phase A containment isolation signal will immediately close:
A. CVCS-8112, RCPs #1 Seal Outlet valve B. FCV-355, CCW Header C Supply valve C. FCV-363, RCP Lube Oil Cooler Return valve D. CVCS-8141A/B/C/D, RCP Seal Leakoff valves Proposed Answer: A. CVCS-8112, RCPs #1 Seal Outlet valve Explanation:
A. Correct - Phase A Containment Isolation closes CVCS-8100 and CVCS-8112, RCPs #1 Seal Outlet valve. This results in RV-8121 lifting and diverting flow to the PRT B. Incorrect - FCV-355 is isolated by a Phase B Containment Isolation signal C. Incorrect - FCV-363 is isolated by a Phase B Containment Isolation signal D. Incorrect - CVCS-8141A/B/C/D fail open on a loss of instrument air following a Phase A Containment Isolation signal.
Technical
References:
LB-6A, Reactor Protection System Supplement LB-6A, Protection Systems LB-1A, Chemical and Volume Control System LF-2, Component Cooling Water References to be provided to applicants during exam: None Learning Objective: 5080 - Analyze automatic features and interlocks associated with the CVCS Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 2 Examination Outline Cross-Reference Level RO Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: Activity levels in primary system Tier #
2 Group #
1 K/A #
004 A1.01 Rating 2.9 Question 2 GIVEN:
Unit 1 is at 100% power Letdown flow decreased to zero Letdown Heat Exchanger Room rose to 175ºF and now is lowering slowly PK04-21, LETDOWN PRESS/FLO/TEMP, alarm is in The operators should:
A. Place Excess Letdown in service and ensure Chemistry is informed to carefully monitor RCS chemistry and activity.
B. Place Excess Letdown in service and begin a continuous makeup to maintain VCT level.
C. Restart a charging pump and place excess letdown in service.
D. Restart a charging pump and initiate continuous makeup to maintain VCT level.
Proposed Answer: A. Place Excess Letdown in service and ensure Chemistry is informed to carefully monitor RCS chemistry and activity.
Explanation:
- Note: 8149A/B/C closed auto on LTDN H/X room. 150°F A. Correct. A Letdown line rupture has occurred downstream of CVCS-1-8152 as evidenced by the High Letdown Heat Exchanger room temperature and initial conditions. Also, Letdown Isolation valves 8149A/B/C failed shut when Letdown Heat Exchanger room temperature reached 150ºF.
AP-18, Letdown Line Failure, directs operators to place Excess Letdown on service and OP B-1A IV, CVCS Excess Letdown Place in Service, cautions operators, When the normal letdown flowpath is out of service and excess letdown is the only letdown flowpath, RCS chemistry and activity should be carefully monitored since the CVCS demins, filters and volume control tank are bypassed during this mode.
B. Incorrect. Makeup to the VCT on a continuous basis should not be necessary because charging is reduced to seals only.
C. Incorrect. Symptoms for Loss of Charging are similar, however, the lowering VCT level differentiates between the two casualties; actions per AP-17, Loss of Charging are incorrect. If Letdown is not operating properly, increasing Charging will not lead to proper resolution of the malfunction.
D. Incorrect. AP-17, Loss of Charging actions are not correct. Pumping the Containment Sump to the Floor Drain Receivers might be beneficial but does not address trends, initial conditions or system operating conditions associated with alarm PK4-21 LETDOWN PRESS/FLO/TEMP.
DCPP L111 NRC Exam 21 November, 2012 P a g e l 3 Technical
References:
AP-17, Loss of Charging; AP-18, Letdown Line Failure; OP B-1A IV, CVCS Excess Letdown Place in Service; PK4-21 LETDOWN PRESS/FLO/TEMP; Operations Lesson Chemical and Volume Control System, LB-1A.
References to be provided to applicants during exam: None Learning Objective: Discuss significant precautions and limitations associated with the CVCS. (5093)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.5 / 45.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 4 Examination Outline Cross-Reference Level RO Knowledge of the effect of a loss or malfunction on the RHRS will have on the following: ECCS Tier #
2 Group #
1 K/A #
005 K3.05 Rating 3.7 Question 3 GIVEN:
- A LOCA occurs
- Offsite power is available
- The plant is aligned for Cold Leg Recirculation
- SI has been reset A loss of Start-Up power occurs. The Diesel Emergency Generators start and load their respective vital bus.
Which of the following actions will be taken by the operator?
A. Trip the ECCS CCPs if an RHR pump does not automatically start.
B. Manually start the RHR pumps, then manually start the ECCS CCPs.
C. Hold the switches for the ECCS CCPs in STOP/RESET until RHR is in service.
D. Manually restart the ECCS CCPs after RHR pumps have automatically started.
Proposed Answer:
Explanation:
The loss of startup power results in a loss of RHR because SI has been reset, so there is no signal to auto start the RHR (or SI) pumps. However, the transfer to diesel starts the charging pumps, which are aligned to take suction from the discharge of the RHR pumps. Because there is no RHR flow, suction to the ECCS CCPs is lost. To prevent the charging pumps from running without suction, the operator must hold the switches for the charging pumps in stop until an RHR pump can be started and restore suction to the pumps.
Answer A incorrect - the RHR pumps must be started by the operator.
Answer B incorrect - the CCPs will automatically start.
Answer C incorrect - the switches are held in STOP/RESET to prevent the CCPs from running until an RHR pump can be started. Otherwise the pumps would be running without a suction source.
Answer D incorrect - CCPs automatically start. Because SI is reset, the RHR do not.
Technical Reference(s):
E-1, Foldout page Proposed references to be provided to applicants during examination:
NONE C. Hold the switches for the ECCS CCPs in STOP/RESET until RHR is in service.
DCPP L111 NRC Exam 21 November, 2012 P a g e l 5 Learning Objective:
42458 Explain basis of emergency steps of E-1.3 Question Source:
Bank #
X P-50918 Modified Bank #
(Note changes or attach parent)
New Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 7
55.43
DCPP L111 NRC Exam 21 November, 2012 P a g e l 6 Examination Outline Cross-Reference Level RO Ability to monitor automatic operation of the ECCS, including: Pumps Tier #
2 Group #
1 K/A #
006 A3.02 Rating 4.1 Question 4 Following an SI actuation, the Safety Injection Pump 1-1 will trip if:
A. RWST level lowers to 30%.
B. valve 8925, RHR 1-1 heat exchanger to Safety Injection pumps suction isolation valve, is closed.
C. power to 4 kV bus F is transferred to the Emergency Diesel Generators.
D. the motor stator temperature reaches 219°F.
Proposed Answer: C. power to 4 kV bus F is transferred to the Emergency Diesel Generators Explanation:
A. Incorrect. An RWST level less than 33 percent will trip the RHR pumps but not the SI pumps.
B. Incorrect. SI pump 1 will trip if this valve is closed unless it is in the SI mode of operation.
C. Correct. If electrical power is transferred to the EDG, the SI pump will automatically trip.
D. Incorrect. This will cause an alarm but not a trip of the motor.
Technical
References:
System Lesson Emergency Core Coolings System LB-3, E-0 Reactor Trip or Safety Injection References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Emergency Core Cooling System. (8045)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.7 / 45.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 7 Examination Outline Cross-Reference Level RO Knowledge of the physical connections and / or cause effect relationships between the PRTS and the following systems: RCS Tier #
2 Group #
1 K/A #
007 K1.03 Rating 3.0 Question 5 GIVEN:
Unit 1 experienced a faulted steam generator outside containment resulting in a Reactor Trip and Safety Injection The faulted steam generator has been isolated RCS pressure is 1800 psig and RISING slowly PK05-25, PRT Pressure Hi/Low, is alarming PRT level and pressure are RISING slowly The reason the PRT parameters are rising is:
A. A Pressurizer Power Operated Relief Valve is open.
B. Reactor Coolant Pump Seal Water Return flow is going to the PRT.
C. Normal letdown flow is diverted to the PRT through RV-8117, Letdown Relief Valve.
D. RCS-8030, Primary Water valve to the PRT, has failed open due to instrument air isolation to containment.
Proposed Answer: B. Reactor Coolant Pump Seal Water Return flow is going to the PRT.
Explanation:
A. Incorrect - The PORVs will discharge to the PRT but RCS pressure is 1800 psig below PORV lift setting and RCS pressure is increasing.
B. Correct - Seal Water Return is isolated by the Safety Injection Signal which diverts water through 8121, Seal Water Return Relief Valve to the PRT.
C. Incorrect - Letdown was isolated by the Safety Injection Signal. RV-8117 will not lift to relieve to the PRT because letdown is isolated.
D. Incorrect - RCS-8030 is isolated from the primary water supply by RCS-8029 following Phase A Containment Isolation preventing water to flow to the PRT.
Technical
References:
LA-4B, Pressurizer Relief Tank, Revision 13 LB-1A, Chemical and Volume Control System, Revision 18 LA-4A, Pressurizer, Pressure and Level Control, Revision 17 AR PK05-25, PRT Press/LVL Temp, Revision 15 References to be provided to applicants during exam: None Learning Objective: 4945 - Describe system interrelationships between the PRT system and other plant systems Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.2-9
DCPP L111 NRC Exam 21 November, 2012 P a g e l 8 Examination Outline Cross-Reference Level RO Ability to locate and operate components, including local controls. 008 Component Cooling Water Tier #
2 Group #
1 K/A #
008 2.1.30 Rating 4.4 Question 6 GIVEN:
Unit 1 is in MODE 4 and preparing to enter MODE 3 PK01-06, CCW Vital Header A/B, comes in due to input number 459, RHR HX 1-2 CCW FLO Hi-Lo The Aux Watch has been tasked to verify Component Cooling Water (CCW) flow on the outlet of the affected RHR HX 1-2.
This would be performed by locally checking CCW flow on the outlet of RHR HX 1-2:
A. In RHR heat exchanger 1-2 room at the 115 foot elevation of the Auxiliary Building. If flow through the RHR HX is low, then throttle the RHR HX Number 1-2 CCW outlet valve, CCW-1-151.
B. In RHR heat exchanger 1-2 room at the 115 foot elevation of the Auxiliary Building. If flow through the RHR HX is low, then throttle RHR heat exchanger 1-1 CCW outlet valve, CCW-1-459.
C. At the 100 foot elevation of the GE-GW Penetration passageway. If flow through the RHR HX is low, then throttle the RHR HX Number 1-2 CCW outlet valve, CCW-1-151.
D. At the 100 foot elevation of the GE-GW Penetration passageway. If flow through the RHR HX is low, then throttle RHR heat exchanger 1-1 CCW outlet valve, CCW-1-459.
Proposed Answer: C. At the 100 foot elevation of the GE-GW Penetration passageway. If flow through the RHR HX is low, then throttle the RHR HX Number 1-2 CCW outlet valve, CCW-1-151.
Explanation:
A. Incorrect. The RHR #2 HX is at this location, but not the outlet valves for CCW.
B. Incorrect. The CCW HXs are located at the 85 foot elevation of the turbine building. Also operating CVCU 1-3 outlet valve is the action to take if flow is too high.
C. Correct. The correct location to observe RHR HX CCW flow is the 100 foot elevation of the GE-GW Penetration passageway. If flow is too low, the correct action is to throttle the RHR HX Number 2 CCW outlet valve, CCW-1-151.
D. Incorrect. Throttling flow to RHR HX Number 1 CCW outlet valve would not correct the given condition since the alarm came in on RHR HX 1-2 (Unit 1, heat exchanger number 2).
Technical
References:
Operations Lesson Component Cooling Water LF-2, AR PK 01-06.
References to be provided to applicants during exam: None Learning Objective: Identify the location of components associated with the CCW System. (8128)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.7 / 45.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 9 Examination Outline Cross-Reference Level RO Knowledge of the bus power supplies to the following: Controller for PZR spray valve Tier #
2 Group #
1 K/A #
K2.02 Rating 2.5 Question 7 The TRICON for the non-safety related Unit 1 Pressurizer Spray valve controllers, PCV-455A and 455B, is powered from:
A. Normal - PY-11; Backup - PY-14 B. Normal - PY-16; Backup - PY-17N C. Normal - PY-14; Backup - PY-16 D. Normal - PY-17N; Backup - PY-11 Proposed Answer: B. Normal - PY-16; Backup - PY17N Explanation:
NOTE: Unit difference.
A. Incorrect. The only controllers with vital 120 VAC PYs are the AFW controllers. Also, PY-11 and 14 are power supplies to SSPS slave relays.
B. Correct - The controllers are powered by PCS, which is powered by PY-16 and 17N (non-vital supplies).
C. Incorrect. Although spray valves are on a primary system, they are not powered from vital 120 VAC supply.
D. Incorrect. Both power supplies are non-vital.
Technical
References:
OP O-2, R116C5 References to be provided to applicants during exam: None Learning Objective: 9990 - State the power supplies to the Pzr, Pzr Pressure and Level Control System components Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 10 Examination Outline Cross-Reference Level RO Knowledge of RPS design feature(s) and/or interlock(s) which provide for the following: Automatic or manual enable/disable of RPS trips Tier #
2 Group #
1 K/A #
012 K4 06 Rating 3.2 Question 8 During a plant startup, when reactor power exceeds 10% on at least two power range nuclear instruments,
_____________ is automatically blocked.
A. the IR C-1 rod stop B. manual reinstatement of the SR high voltage C. the PR High Flux Low Set point reactor trip D. the IR reactor trip Proposed Answer: B. manual reinstatement of the SR high voltage Explanation:
A. Incorrect. The C-1 rod stop is not automatic, but can be manually blocked when above the P-10 interlock.
B. Correct. When P-10 is energized, the SR high voltage may not be reinstated.
C. Incorrect. When P-10 is energized, the PR High Flux Low Set point reactor trip may be manually blocked.
D. Incorrect. When P-10 is energized, it allows the IR trip to be manually blocked.
Technical
References:
System Lesson Reactor Protection System, LB-6A References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Reactor Protection System. (37048)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.7 Added nuclear after power range in question
DCPP L111 NRC Exam 21 November, 2012 P a g e l 11 Examination Outline Cross-Reference Level RO Ability to manually operate and/or monitor in the control room:
ESFAS initiated equipment which fails to actuate Tier #
2 Group #
1 K/A #
013 A4.01 Rating 4.5 Question 9 GIVEN:
Monitor Light Box D o
Phase B Train A and Train B Red Activated Lights - On o
Two Phase B White Status Light - On E-0, Reactor Trip or Safety Injection, is in use Which of the following valves could be open and causing the Phase B white lights to be lit?
A. Main Steam isolation valve, FCV-41 and RCP Thermal Barrier Return valve, FCV-357 B. RHR Heat Exchanger 1-1 Outlet Header to Spray Header "A" valve, CS-9003A and CCW Header C Supply valve, FCV-355 C. Excess Letdown Heat Exchanger Outlet valve, FCV-361 and RCP Lube Oil Cooler valve, FCV-363 D. CCW Supply to RCP and Reactor Vessel Support Coolers valve, FCV-356 and RCP Thermal Barrier Return valve, FCV-750 Proposed Answer:
D. CCW Supply to RCP and Reactor Vessel Support Coolers valve, FCV-356 and RCP Thermal Barrier Return valve, FCV-750 Explanation:
A. Incorrect - MSIVs are closed by MSI when containment pressure reaches the Phase B setpoint but are not on the Phase B light box portion.
B. Incorrect - The CS isolation valve should be open closed, but is not a Phase B valve.
C. Incorrect - Excess letdown valve do not automatically close and the light would be lit, however, the valve is not on the phase B light box.
D. Correct - CCW to and from the RCPs isolate, if either of these valves are open, the lights would be lit.
Technical
References:
E-0, Reactor Trip or Safety Injection, Revision 40 LPE-0, Reactor Trip and Safety Injection Response, Revision 11, LB-6A References to be provided to applicants during exam: None Learning Objective: 3798 - Explain the means to verify ECCS injection valve alignment Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.7 Added valve to distractors
DCPP L111 NRC Exam 21 November, 2012 P a g e l 12 Examination Outline Cross-Reference Level RO Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of service water Tier #
2 Group #
1 K/A #
022 A2.04 Rating 2.9 Question 10 GIVEN:
A Steam Line Rupture has occurred in Unit 1 RCS Pressure is 1712 psig and slowly rising Containment Pressure is 29 psig and slowly rising All Emergency Core Cooling Systems and Safety Injection systems actuated as designed Auxiliary Saltwater (ASW) Pumps #1 and #2 have tripped and cannot be restarted The Refueling Water Storage Tank (RWST) level is 2%
In order to continue the cooldown of Unit 1; A. RHR must be aligned to take a suction from the Containment Structure Sump.
B. The Containment Spray pumps will use water from the Boric Acid Storage Tank (BAST) to continue providing flow to the CS system.
C. Fire water must be used to cool the CCW heat exchangers.
D. Open the Unit 1 to Unit 2 ASW cross-tie valve, FCV-601, to restore ASW cooling.
Proposed Answer: D. Open the Unit 1 to Unit 2 ASW cross-tie valve, FCV-601, to restore ASW cooling.
Explanation:
A. Incorrect. RHR must be manually aligned to take a suction from the Containment Recirculation Sump.
B. Incorrect. The Containment Spray Pumps will provide forced flow for the CS system from the RWST, not the BAST.
C. Incorrect. The site was designed so that ASW could be cross-connected between Unit 1 and Unit 2 (ie Unit 2 may supply ASW to Unit 1 and vice versa).
D. Correct. The RHR heat exchangers are cooled by CCW, and the CCW heat exchangers are cooled by ASW. AP-10 directs operators to open the Unit 1 and Unit 2 ASW cross-tie valve.
Technical
References:
AP-10, Loss of Auxiliary Saltwater; AP-11, Malfunction of Component Cooling Water System including Appendix D, Loss of the Ultimate Heat Sink; systems lesson Containment Spray; Operations Lesson Emergency Core Cooling Systems.
References to be provided to applicants during exam: None Learning Objective: Describe system interrelationships between the ASW System and other plant systems. (3785)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.5 / 43.5 / 45.3 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 13 Examination Outline Cross-Reference Level RO Knowledge of operational implications of the following concepts as they apply to the Hydrogen Recombiner and Purge System and flammable hydrogen concentration.
Tier #
2 Group #
1 K/A #
028 K5.02 Rating 3.4 Question 11 Procedurally, which of the following is the MAXIMUM hydrogen concentration for operating the hydrogen recombiners due to possible hydrogen ignition?
A. 4.0%
B. 4.5%
C. 5.0%
D. 7.0%
Proposed Answer:
A. 4.0%
Explanation:
A. Correct -OP H in service above 0.5%. P&L 5.2 states ignition could occur at 4%.
B. Incorrect - OP H in service above 0.5%. P&L 5.2 states ignition could occur at 4%.
C. Incorrect - OP H in service above 0.5%. P&L 5.2 states ignition could occur at 4%.
D. Incorrect - OP H in service above 0.5%. P&L 5.2 states ignition could occur at 4%.
Technical
References:
LH-9, Containment Hydrogen Recombiners, Revision 11A OP H-9, Inside Containment H2 Recombination System, Revision 10 References to be provided to applicants during exam: None Learning Objective: Discuss significant precautions and limitations associated with the Containment Hydrogen Recombiner System (7062)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 14 Examination Outline Cross-Reference Level RO Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment pressure Tier #
2 Group #
1 K/A #
026 A1.01 Rating 3.9 Question 12 GIVEN:
A LOCA has occurred on Unit 1 4 kV bus G is de-energized Containment Spray pump 1-1 is inoperable Safety Injection has actuated Containment Pressure is 23 psig and rising Only 2 Containment Fan Cooling Units (CFCUs) are operating The crew is about to transition from E-0, Reactor Trip or Safety Injection Based on these conditions, Containment Spray pump 1-2 __________ started and containment pressure A. has NOT; will exceed 57 psig.
B. has; will NOT exceed 47 psig.
C. has; will exceed 47 psig.
D. has NOT; will NOT exceed 57 psig.
Proposed Answer: B. has; will NOT exceed 47 psig.
Explanation:
A. Incorrect. CS Train B will have actuated on Hi Hi Containment Pressure at 22 psig. The 1-2 pump is powered from Bus H. Bus G powers pump 1-1 (already inoperable). The design limit for Containment Pressure is 47 psig; pressure will not exceed 57 psig.
B. Correct. CS Train B will have actuated on Hi Hi Containment Pressure at 22 psig. One train of CS and two CFCUs in operation will prevent Containment Pressure from exceeding the design limit of 47 psig.
C. Incorrect. Containment Pressure will NOT exceed 47 psig.
D. Incorrect. CS Train B will have actuated on Hi Hi Containment Pressure at 22 psig. Containment Pressure will not exceed 47 psig.
Technical
References:
LB-6A, LI-2 References to be provided to applicants during exam:
None Learning Objective: Explain significant Containment Spray System design features and the importance to nuclear safety. (40802)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.5 / 45.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 15 Examination Outline Cross-Reference Level RO Knowledge of the physical connections and/or cause effect relationships between the MRSS and the following systems: AFW Tier #
2 Group #
1 K/A #
039 K1.07 Rating 3.4 Question 13 The TDAFW Pump steam supply is from steam generators:
A. 1-1 and 1-2 upstream of the Main Steam Isolation Valves.
B. 1-1 and 1-2 downstream of the Main Steam Isolation Valves.
C. 1-2 and 1-3 upstream of the Main Steam Isolation Valves.
D. 1-2 and 1-3 downstream of the Main Steam Isolation Valves.
Proposed Answer:
C. 1-2 and 1-3 upstream of the Main Steam Isolation Valves.
Explanation:
A. Incorrect - Supplied from SG 2 and 3.
B. Incorrect - Supplied from SG 2 and 3 and upstream of the MSIVs.
C. Correct. Supplied from SG 2 and 3 and upstream of the MSIVs D. Incorrect - Supplied from upstream of the MSIVs.
Technical
References:
L C-2A, Main Steam System, Revision 14 References to be provided to applicants during exam: None Learning Objective: 8418 - Describe the basic flow path of the AFW system Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.2-9
DCPP L111 NRC Exam 21 November, 2012 P a g e l 16 Examination Outline Cross-Reference Level RO Ability to monitor automatic operation of the MFW, including:
Feedwater isolation Tier #
2 Group #
1 K/A #
059 A3.06 Rating 3.2 Question 14 GIVEN:
The plant trips from 100% power due to P-14 RCS temperature is currently 558°F What is the expected status of the main feedwater system for the current plant conditions?
A. Main Feedwater pumps tripped, feedwater isolation valves open, main feedwater regulating and bypassed valves closed.
B. Main Feedwater pump running, feedwater isolation valves open, main feedwater regulating and bypassed valves closed.
C. Main Feedwater pumps tripped, feedwater isolation valves closed, main feedwater regulating and bypassed valves closed.
D. Main Feedwater pumps tripped, feedwater isolation valves closed, main feedwater regulating and bypassed valves open.
Proposed Answer: C. Main Feedwater pumps tripped, feedwater isolation valves closed, main feedwater regulating and bypassed valves closed Explanation:
A. Incorrect. P-14 will close the FWIV and trip the MFPs, P-14 (and/or P-4 below 554°F) will close the reg and bypass valves.
B.
Incorrect. This is the expected state if feedwater is isolated due to low Tave w/P-4.
C.
Correct. P-14 actuates and trips the pumps and closes the FWIV, reg and bypass valves.
D. Incorrect. Although above the low Tave setpoint of 554°F setpoint the reg and bypass valves are closed by P-14.
Technical
References:
System Lessons Main Feedwater System and Auxiliary Feedwater System; Reactor Protection System Supplement References to be provided to applicants during exam: None Learning Objective: Describe system interrelationships between the Main Feedwater System and other plant systems. (37615)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.7 / 45.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 17 Examination Outline Cross-Reference Level RO Knowledge of the operational implications of the following concepts as they apply to AFW: Relationship between AFW flow and RCS heat transfer Tier #
2 Group #
1 K/A #
K5.01 Rating 3.6 Question 15 E-0, Reactor Trip or Safety Injection, requires the operator to ensure AFW flow is at least 435 gpm.
The basis for this minimum flow is to:
A. make up for the initial shrink in SG water level.
B. remove RCS decay heat.
C. maintain SG water level in the narrow range.
D. ensure a sufficient heat sink to initiate natural circulation.
Proposed Answer:
B. remove RCS decay heat.
Explanation:
A. Incorrect - While the SG level will initially shrink, it is not the basis for the AFW flow requirement.
B. Correct - The basis document states this minimum flow is the minimum for heat removal. The design basis for the AFW minimum flow on a loss of feedwater is to prevent over pressurization of the primary system due to a loss of secondary heat sink if there is not sufficient flow.
C. Incorrect - SG level may shrink out of the narrow range but this is not the reason for the minimum AFW flow rate.
D. Incorrect - E-0 is not used for natural circulation.
Technical
References:
DCPP Units 1 & 2 FSAR Update, Revision 19 E-0 Background Documents LD-1, Auxiliary Feed Water System, Revision 14 References to be provided to applicants during exam:
None Learning Objective: 8430 - Explain significant Auxiliary Feed Water System design features and the importance to nuclear safety.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 18 Examination Outline Cross-Reference Level RO Knowledge of bus power supplies to the following: Major system loads Tier #
2 Group #
1 K/A #
062 K2.01 Rating 3.3 Question 16 A loss of the Unit 1 12 kV Bus D would result in a loss of power to which of the following components?
- 1. RCP 1-1
- 2. RCP 1-2
- 3. Circulating Water Pump 1-1
- 4. Circulating Water Pump 1-2 A. 1 and 3 B. 1 and 4 C. 2 and 3 D. 2 and 4 Proposed Answer: C. 2 and 3 Explanation:
A. Incorrect. RCP 1-1 is powered from bus E.
B. Incorrect. RCP 1-1 is powered from bus E.
C. Correct.
D. Incorrect. Circulating Water Pump 1-2 is powered from bus E.
Technical
References:
Systems Lessons Reactor Coolant Pump and Circulating Water References to be provided to applicants during exam: None Learning Objective: State the power supplies to Circulating Water System components. (8346)
State the power supplies to RCP components. (6080)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 19 Examination Outline Cross-Reference Level RO Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following: EDG Tier #
2 Group #
1 K/A #
K3.01 Rating 3.7*
Question 17 Unit 2 is at 100% power and EDG 2-2 has been running in parallel with off-site power for the last two hours.
Then, DC Control power is lost to EDG 2-2.
The loss of DC Control Power to the diesel will result in which of the following?
A. The Air Start solenoid valves for the diesel closing.
B. Inability to shutdown the diesel from the control room.
C. Losing excitation voltage to the generator.
D. Frequency rising to 63.3 Hz.
Proposed Answer:
B. Inability to shutdown the diesel from the control room.
Explanation:
A. Incorrect - Air start solenoids close during the starting sequence. Therefore the valves will already be closed at the time of the loss of DC.
B. Correct - Loss of DC control power will result in the loss of the 125VDC power supply and will result in the loss of the EDGs Tachometer Pack, Voltage Regulator (Manual), Alarms, Indications and the Shutdown Relay.
C. Incorrect - Excitation voltage is provided by self excitation from the EDG output while running and loss of DC Control Power will result in the loss of the Field Flash used at startup.
D. Incorrect - Frequency can not increase when paralleled to off-site power.
Technical
References:
LJ6B, Diesel Generator System LJ9, DC Power References to be provided to applicants during exam: None Learning Objective: 37724 - Describe controls, indications, and alarms associated with the Diesel Generator System.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7 Added to the diesel in question
DCPP L111 NRC Exam 21 November, 2012 P a g e l 20 Examination Outline Cross-Reference Level RO Process Rad Monitors. Ability to manually operate and/or monitor in the control room: Effluent release Tier #
2 Group #
1 K/A #
073 A4.01 Rating 2.7 Question 18 While conducting a planned effluent release from Gas Decay Tank 1-1, Unit 1 received alarm PK 11-21, High Radiation, due to RM-22, Gas Decay Tank to Plant Vent radiation monitor in high alarm.
As a result of this valid alarm, which of the following valves will close?
- 1. Plant Vent Valve, RCV-17
- 2. Unit1/Unit 2 Crosstie valve, FCV-417
- 3. Gas Decay Tank Outlet Header to Plant Vent Valve, FCV-410 A. 1 and 2 B. 1 and 3 C. 2 and 3 D. 1 only Proposed Answer: D. 1 only Explanation:
A. Incorrect. PK 11-21, High Radiation was received due to RM-22, Gas Decay Tank to Plant Vent Radiation Monitor reaching the high alarm set point. Plant Vent Valve, RCV-17, automatically shuts on a high radiation alarm from RM-22. However, FCV-417, Unit 1/Unit 2 Crosstie valve is open. FCV-417 shuts when Gas Decay Tank 1-3 or 2-3 is selected for venting or purging.
B. Incorrect. Gas Decay Tank Outlet Header to Plant Vent Valve, FCV-410, is key operated and administratively controlled. It would be open during an effluent release.
C. Incorrect. FCV-417 shuts when Gas Decay Tank 1-3 or 2-3 is selected for venting or purging.
FCV-410, is key operated and administratively controlled and would be open during an effluent release.
D. Correct. Plant Vent Valve, RCV-17, automatically shuts on a high radiation alarm from RM-22.
RE-22 must be reset for RCV-17 to open.
Technical
References:
LG-2, Gaseous Radwaste; AR PK 11-21 High Radiation References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Gaseous Radwaste System. (37707)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.7 / 45.7 Reworded question to add valves to question and remove from each of the listed valves
DCPP L111 NRC Exam 21 November, 2012 P a g e l 21 Examination Outline Cross-Reference Level RO Knowledge of the effect of a loss or malfunction of the following will have on the EDG system: Air Receivers Tier #
2 Group #
1 K/A #
064 K6.07 Rating 2.7 Question 19 GIVEN:
- Unit 1 is at 100% power.
- A relief valve has failed open on Starting Air Receiver A for Diesel Generator 1-1.
- The leakage exceeds the capacity of the starting air compressor.
- The NO is responding to the associated alarm.
Which of the following describes the response of Diesel Generator 1-1 if a start signal occurs before any operator action is taken?
A. It will start in the normal time from Starting Air Receiver B via all four starting air solenoids.
B. It will start in the normal time via the two starting air solenoids associated with the Starting Air Receiver B.
C. It will take longer to start or may NOT start because the starting air system will be depressurizing/depressurized.
D. It will take longer to start or may NOT start because the fuel rack fails to no fuel position and may not reset.
Proposed Answer: B. It will start in the normal time via the two starting air solenoids associated with the Starting Air Receiver B.
Explanation:
A incorrect. Only two solenoids associated with B receiver.
B correct. The B receiver is capable of starting the diesel by opening the two solenoid valves for the air motors.
C incorrect. Receivers are not cross tied.
D incorrect. Loss of one receiver will not affect operation.
Technical
References:
L6B, Diesel Generators References to be provided to applicants during exam: None Learning Objective: 6431 - State the purpose of D/G subsystems and components Question Source:
Bank # DCPP 2007 X
(note changes; attach parent)
Modified Bank #
New Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.8
DCPP L111 NRC Exam 21 November, 2012 P a g e l 22 Examination Outline Cross-Reference Level RO Service Water. Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: Automatic opening features associated with SWS isolation valves to CCW heat exchanges Tier #
2 Group #
1 K/A #
076 K4.03 Rating 2.9 Question 20 GIVEN:
Unit 1 is at 100% power Auxiliary Saltwater Pump (ASW) 1-2 is running ASW Pump 1-1 is in standby ASW Pump Discharge Cross-Connect Valves, FCV-495 and FCV-496 are open Unit 1 and Unit 2 Cross-Tie Valve, FCV-601, is closed FCV-602, ASW/CCW Heat Exchanger #1 Inlet Valve was cleared, jacked closed and the air supply isolated An operator when removing the clearance on FCV-602, unjacks the valve but does not re-establish the air supply. Shortly after, PK 01-01 ASW SYS HX DELTA P/HDR PRESS, and PK 01-03, Aux Salt Water Pumps, alarms.
With these conditions, FCV-602; A. is shut and ONLY ASW pump 1-2 is running.
B. is open and BOTH ASW pumps are running.
C. must be in the same position as FCV-603 with only 1 pump running.
D. will shut after a 10 second time delay and the operating pump will switch to Startup Transformer power Proposed Answer: B. is open and BOTH ASW pumps are running.
Explanation:
A. Incorrect. The ASW/CCW HX inlet valves are air operated and fail open. Without Instrument air or the backup air supply, the valves will fail open. It is true that the operating ASW pump remains running but the standby pump will also start on low pressure at 40.5 psig.
B. Correct. FCV-602 will be open; without Instrument air or the backup air supply, the valves will fail open. The standby pump will have started on low system pressure.
C. Incorrect. During normal operations, only one of the ASW/CCW HX Inlet valves will be open. Both ASW pumps will be running.
D. Incorrect. There is no time delay associated with the operation of the ASW/CCW HX Inlet valves.
The operating pump will auto transfer to Startup power after a Safety Injection signal.
Technical
References:
LE-5, Auxiliary Saltwater System; DCPP-AR A0662367 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the ASW System. (5365)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
DCPP L111 NRC Exam 21 November, 2012 P a g e l 23 New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 24 Examination Outline Cross-Reference Level RO Ability to explain and apply system limits and precautions associated with Instrument Air Tier #
2 Group #
1 K/A #
078 2.1.32 Rating 3.8 Question 21 Air compressors _______________ are controlled from the Unit 2 Control Room and they will automatically start at _________ instrument air pressure if the Compressor Control switch is in AUTO.
A. 0-1 and 0-2; 98 psig and decreasing B. 0-3 and 0-4; 98 psig and decreasing C. 0-1 and 0-2; 93 psig and decreasing D. 0-3 and 0-4; 93 psig and decreasing Proposed Answer:
D. 0-3 and 0-4; 93 psig and decreasing.
Explanation:
A. Incorrect - Compressors 0-1 and 0-2 are controlled from Unit 1 and 98 psig is the pressure at which the compressor will load if the compressor Control switch is in the ON position and the Master Unloader ON.
B. Incorrect - Compressors 0-3 and 0-4 are controlled from Unit 2 but 98 psig is the pressure at which the compressor will load if the compressor Control switch is in the ON position and the Master Unloader ON.
C. Incorrect - Compressors 0-1 and 0-2 are controlled from Unit 1 but the compressor will start if the Compressor Control switch is in AUTO at 93 psig.
D. Correct - Compressors 0-3 and 0-4 are controlled from Unit 2 and the compressor will start if the Compressor Control switch is in AUTO at 93 psig Technical
References:
LK-1, Compressed Air System References to be provided to applicants during exam: None Learning Objective: 7226 - Describe controls, indications, and alarms associated with the Compressed Air System Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 25 Examination Outline Cross-Reference Level RO Containment:
Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Containment evacuation (including recognition of the alarm)
Tier #
2 Group #
1 K/A #
103 A2 04 Rating 3.5 Question 22 Unit 1 is in MODE 6. Core offload is in progress.
The following alarms are received:
PK02-06, CONTMT VENT ISOLATION PK11-19, CONTMT RADIATION PK11-21, HIGH RADIATION The SRO in Containment reports it appears a fuel assembly has been damaged during transit to the upender. The crew enters AP-21, Irradiated Fuel Damage.
In accordance with AP-21, what will be the first action performed by the operator in the Control Room?
A. Contact RP to verify alarm and dose rate B. Activate the Containment Evacuation alarm C. Verify Containment Ventilation Isolation D. Start fan E-15 and E-16, CONTMT Iodine Removal Units Proposed Answer: B. Activate the Containment Evacuation alarm Explanation:
A incorrect. this would have to be done in Containment by the refueling SRO B correct. Necessary to protect the safety of those on site. Done before once the alarms are validated (in setup). The operator should understand the urgent need to alert site personnel.
C incorrect. Done after the alarm is activated.
D incorrect. Done after the alarm is activated.
Technical
References:
AP-21 References to be provided to applicants during exam: None Learning Objective: Explain the actions for fuel damage, actual or suspected. (6619)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.5 / 43.5 / 45.3 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 26 Examination Outline Cross-Reference Level RO Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: Closed cooling water flow rate and temperature Tier #
2 Group #
1 K/A #
005 A1.03 Rating 2.5 Question 23 GIVEN:
Unit 2 is in MODE 6 RHR pump 2-1 and both RHR Heat Exchangers are in service for decay heat removal CCW pumps 2-1 and 2-2 are operating CCW pump 2-3 is out of service If one of the running CCW pumps is lost, the operator would throttle _________ HCV-670, RHR heat Exchanger Bypass valve, and throttle __________ HCV-637/638, RHR HX Outlet valve in order to maintain the same cooldown rate while maintaining the flowrate below the limit in OP B-2:V, RHR Place in Service.
A. open: open B. open; shut C. shut; open D. shut; shut Proposed Answer: C. shut; open Explanation:
A. Incorrect - Throttling open HCV-670 will bypass the heat exchanger and limit cooldown rate in addition opening HCV-670 will exceed the 5000 GPM maximum flowrate to the cold legs B. Incorrect - Throttling open HCV-670 will bypass the heat exchanger and shutting HCV-637/638 will reduce the flow through the heat exchanger resulting in a smaller cooldown rate.
C. Correct - Throttling shut HCV-670 while opening HCV-637/638 will maintain the flowrate to the loops below the 5000 GPM limit in addition to increasing flow through the RHR HX to maintain the same RCS cooldown rate due to the reduced CCW flow to the RHR HX.
D. Incorrect - Throttling HCV-670 and HCV 637/638 shut will limit the flow through the RHR heat exchanger and reduce total flow to the RCS cold legs resulting in a reduced cooldown rate.
Technical
References:
OP B-2:V, RHR - Place in Service, Revision 34 LB2, Residual Heat Removal System OP L-5, Plant Cooldown from Minimum Load to Cold Shutdown, Revision 89 References to be provided to applicants during exam: None Learning Objective: 35319 - Describe the operation of the RHR system Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 27 Examination Outline Cross-Reference Level RO Ability to monitor automatic operation of the MRSS, including:
Isolation of the MRSS (Main and Reheat Steam)
Tier #
2 Group #
1 K/A #
039 A3.02 Rating 3.1 Question 24 Unit 2 is at full power.
The Main Generator output breakers open and turbine speed increases to 1854 rpm.
Which of the following valves will be closed by the Overspeed Protection Circuit (OPC)?
A. Turbine Governor valves only B. Turbine Governor and Intercept valves C. Reheat Stop valves only D. Reheat Stop and Intercept valves Proposed Answer: B. Turbine Governor valves and Intercept valves Explanation:
A. Incorrect. The primary feature of the OPC is the 103% overspeed protection. At 103 percent, the governor and intercept valves are closed until the speed decays to approximately 101 percent.
B. Correct. Both sets of valve close as outlined above.
C. Incorrect. The reheat stop valves do not close.
D. Incorrect. Intercept valves close, however, the Reheat stop valves do not close.
Technical
References:
LC-3B References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Turbine Control Oil System. (37644)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.5 / 45.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 28 Examination Outline Cross-Reference Level RO Knowledge of MFW design feature(s) and/or interlock(s) which provide for automatic feedwater isolation of MFW Tier #
2 Group #
1 K/A #
059 K4.19 Rating 3.2 Question 25 GIVEN:
Unit 1 tripped from 30% power RCS Loop average temperature:
Loop 11: 552°F and lowering slowly Loop 12: 557°F and lowering slowly Loop 13: 553°F and lowering slowly Loop 14: 555°F and lowering slowly Pressurizer Pressure is 1840 psig on all channels and stable Steam Generator Narrow Range Level:
Loop 11: 72% and rising slowly Loop 12: 75% and rising slowly Loop 13: 74% and rising slowly Loop 14: 73% and rising slowly The operator reports the Main Feedwater Isolation valves have closed.
Which of the following signals could have caused the valves to close?
A. SI ONLY B. SI or P-4 coincident with Low RCS Tave C. SI or P-14 D. P-4 coincident with Low RCS Tave ONLY Proposed Answer: A. SI ONLY Explanation:
A. Correct - The SI actuated at primary pressure of 1850 and initiated feedwater isolation to close the isolation valves.
B. Incorrect - The P-4 interlock actuates following a Rx trip with 2/4 average coolant temperatures
<554°F. The SI actuated at primary pressure of 1850 and initiated feedwater isolation and close the main feedwater reg and bypass valves. Only SI closed the isolation valves.
C. Incorrect - The P-14 interlock closes the isolation valves but requires 1/4 Steam Generators level to be > 90% narrow range.
D. Incorrect - The P-4 interlock actuates following a Rx trip with 2/4 average coolant temperatures
<554°F does not close the isolation valves.
Technical
References:
AR PK09-11, AR PK09-12, OIM B-6-12, LC-8A, LC-6A References to be provided to applicants during exam: None Learning Objective: 37614 - Analyze automatic features and interlocks associated with the Main Feedwater System.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
DCPP L111 NRC Exam 21 November, 2012 P a g e l 29 Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 30 Examination Outline Cross-Reference Level RO Knowledge of the effect that a loss or malfunction of the AFW will have on the following: S/G Tier #
2 Group #
1 K/A #
061 K3.02 Rating 4.2 Question 26 GIVEN:
The reactor tripped and Safety Injection actuated due to Steam Generator 1-1 steam break inside Containment Containment pressure peaked at 4.5 psig Containment pressure is currently 2.0 psig Wide Range level in Steam Generator 1-1 is zero Steam Generator 1-1 has been isolated Narrow Range levels in Steam Generators 1-2, 1-3 and 1-4 are all 18% and rising slowly The crew has established 250 gpm of AFW flow to each intact steam generator The TDAFW pump has tripped Motor Driven AFW (MDAFW) pump 1-3 trips.
What is the effect of the trip of the MDAFW 1-3 pump on the steam generators ability to remove decay heat at this time?
A. Heat sink requirements are NOT met; there is less AFW flow than required for a secondary heat sink; the crew should reestablish AFW flow to Steam Generator 1-1.
B. Heat sink requirements are NOT met; there is less AFW flow than required for a secondary heat sink; the crew should go to FR-H.1, Response to a Loss of Secondary Heat Sink.
C. Heat sink requirements are met based on sufficient AFW flow to all the intact steam generators from MDAFW pump 1-2.
D. Heat sink requirements are met based on sufficient level in the intact steam generators.
Proposed Answer: D. Heat sink requirements are met based on sufficient level in the intact steam generators.
Explanation:
A incorrect. There is less than 435 gpm AFW flow, however, there is adequate steam generator level in the intact steam generators and AFW flow is not reestablished to a faulted steam generator unless needed for cooldown.
B incorrect. There is less than 435 gpm, however there is adequate level in the intact steam generators.
C incorrect. The 1-2 MDAFW pump is only feeding the 1-2 steam generator which has 250 gpm, less than the normally required 435 gpm. If the adverse containment value of 25% is thought to be in effect, this would be correct.
D correct. MDAFW 1-2 pump feeds steam generators 1-1 and 1-2. MDAFW pump feeds 1-3 and 1-4.
There will only be the 250 gpm of AFW flow to the 1-2 steam generator, however, the non-adverse containment narrow range level requirement for heat sink is one steam generator greater than 15%,
therefore there is adequate level and a secondary heat sink exists.
Technical
References:
F-0.3 References to be provided to applicants during exam: None
DCPP L111 NRC Exam 21 November, 2012 P a g e l 31 Learning Objective: State the minimum AFW flow required to maintain S/Gs as a heat sink. (6982)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.7 / 45.6
DCPP L111 NRC Exam 21 November, 2012 P a g e l 32 Examination Outline Cross-Reference Level RO Knowledge of the effect of a loss or malfunction of the fuel oil storage tanks will have on the EDG system Tier #
2 Group #
1 K/A #
064 K6.08 Rating 3.2 Question 27 If there is less than the required level in the diesel fuel oil storage tanks, the emergency diesel generators may not operate for the required Engineered Safeguards MINIMUM assumed time of:
A. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. 7 days C. 14 days D. 30 days Proposed Answer: B. 7 days.
Explanation:
A. Incorrect. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a frequently used tech spec action time.
B. Correct. Calculated fuel oil consumption necessary to support the operation of the DGs to power the minimum engineered safety feature (ESF) systems required to mitigate a design basis accident (LOCA) in one unit and those minimum required systems for a concurrent non-LOCA safe shutdown in the remaining unit (both units initially in MODE 1 operation). The fuel oil consumption is calculated for a period of 7 days operation of minimum ESF systems.
C. Incorrect. 14 days is a frequent tech spec action time.
D. Incorrect. 30 days is a frequent tech spec action time.
Technical
References:
LJ-6B References to be provided to applicants during exam: None Learning Objective: 41342 - Explain significant Diesel Generator System design features and the importance to nuclear safety.
Question Source:
Bank #
X (note changes; attach parent)
Modified Bank #
New Question History:
Last NRC Exam Yes Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.8 Components, capacity, and functions of emergency systems
DCPP L111 NRC Exam 21 November, 2012 P a g e l 33 Examination Outline Cross-Reference Level RO Knowledge of the physical connections and/or cause effect relationships between the containment system and the following systems: CCS Tier #
2 Group #
1 K/A #
103 K1.01 Rating 3.6 Question 28 Initial Conditions:
Unit 1 is at 100% power Containment Fan Cooling Units (CFCU) 1-1, 1-2, and 1-3 are in service in HIGH speed Component Cooling Water Heat Exchanger outlet temperatures are normal PK 01-16, Containment Environment PPC alarm, comes in Containment temperature is 110ºF and slowly rising In this situation, the operators should:
A. Place an additional CFCU in service.
B. Initiate a plant shutdown.
C. Increase CCW flow to maximum available.
D. Secure one or more CRDM fan coolers and lower reactor power level.
Proposed Answer: A. Place an additional CFCU in service.
Explanation:
A. Correct. Place an additional CFCU in service, this will help cool Containment.
B. Incorrect. The technical Specification limit of 120ºF has not been reached; at which point, operators have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore temperature to within limits.
C. Incorrect. The CCW HX outlet temperatures are normal, therefore increasing CCW flow to the CFCUs is not necessary. The problem is that there are not enough CFCUs on service; not the temperature of the cooling water.
D. Incorrect. If one or more CRDM fan coolers were secured, then the correct action would be to start one or more, not to secure them.
Technical
References:
Containment Environment PPC alarm, PK 01-16; Containment Fan Cooling Units, LH-2.
References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the CFCUs. (40812)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.2 to 41.9 / 45.7 to 45.8
DCPP L111 NRC Exam 21 November, 2012 P a g e l 34 Examination Outline Cross-Reference Level RO Ability to (a) predict the impacts of a loss of pressurizer level (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of the loss of pressurizer level Tier #
2 Group #
2 K/A #
011 A2.03 Rating 3.8 Question 29 Unit 1 is at 100% power with CCP 1-1 operating and normal letdown is in service.
The operator observes pressurizer level is 45% and lowering slowly and charging flow rate has risen to 120 GPM. The SFM directs you to take actions in accordance with AP-1, Excessive Reactor Coolant System Leakage.
The sequence of actions would be to (1) and if pressurizer level continues to decrease then (2).
A. (1) Start an additional charging pump; (2) isolate letdown B. (1) Start an additional charging pump; (2) initiate Safety Injection C. (1) Isolate letdown; (2) initiate Safety Injection D. (1) Isolate letdown; (2) trip the reactor Proposed Answer: A. (1) Start an additional charging pump; (2) isolate letdown Explanation:
A. Correct - Order of actions is correct in accordance with AP-1.
B. Incorrect - Isolating letdown is done first to see if the leak is isolated.
C. Incorrect - a charging pump is started first to increase charging.
D. Incorrect - Charging pump is started first, SI is initiated not a reactor trip.
Technical
References:
AP-1, Excessive Reactor Coolant System Leakage, Revision 20 PK05-22, Pressurizer Level Hi/Lo Control, Revision 6 References to be provided to applicants during exam: None Learning Objective: 3538 - List the leakage criteria which would require manual initiation of a safety injection Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 35 Examination Outline Cross-Reference Level RO Knowledge of the physical connections and/or cause effect relationships between the NIS and the following systems: RPS Tier #
2 Group #
2 K/A #
015 K1.01 Rating 4.6 Question 30 The plant is performing a reactor startup. Intermediate Range channels are at 10-9 amps.
A loss of PY-11 occurs.
Which of the following describes the expected plant response?
A. The plant trips on Source Range high flux only.
B. The plant trips on Intermediate Range high flux only.
C. The plant trips on either Source Range or Intermediate Range high flux.
D. The plant remains critical at 10-9 amps.
Proposed Answer: B. The plant trips on Intermediate Range high flux only.
Explanation:
A. Incorrect. At 10-9 amps, the source ranges are de-energized (around 8x10-11 amps on P-6).
B. Correct. Intermediate range channel N-35 de-energizes and the plant trips on 1/2 high flux.
C. Incorrect. Only the intermediate range high flux trip occurs..
D. Incorrect. Plant trips on high flux.
Technical
References:
OIM drawing B-4-2 B-6-2, B-6-4a and b References to be provided to applicants during exam: None Learning Objective: Describe system interrelationships between the Excore Nuclear Instrumentation System and other plant systems. (36973)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.2 to 41.9 / 45.7 to 45.8
DCPP L111 NRC Exam 21 November, 2012 P a g e l 36 Examination Outline Cross-Reference Level RO Ability to manually operate and/or monitor the Non-nuclear instrumentation channel select controls in the control room Tier #
2 Group #
2 K/A #
016 A4.01 Rating 2.9 Question 31 Unit 1 is at full power.
RCS pressure channels indicate as follows:
PT - 455 - 2242 psig PT - 456 - 2235 psig PT - 457 - 2238 psig PT - 474 - 2244 psig Which pressure signal will be used by the PCS for pressurizer pressure control?
A. PT-455 B. PT-456 C. PT-457 D. PT-474 Proposed Answer: A. PT-455 Explanation: Note: Unit Difference A. Correct - Unlike Unit 2, which is the selected channel, Unit 1 Pressurizer pressure control is maintained by the second highest pressure signal.
B. Incorrect - Pressurizer pressure control is maintained by the second highest pressure signal.
C. Incorrect - Pressurizer pressure control is maintained by the second highest pressure signal.
D. Incorrect - Pressurizer pressure control is maintained by the second highest pressure signal.
Technical
References:
OP A-4A:I, Pressurizer - Make Available, Revision 29 References to be provided to applicants during exam: None Learning Objective: 4560 - Describe the operation of the Pzr, Pzr Pressure and Level Control System Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 37 Examination Outline Cross-Reference Level RO Ability to interpret and execute procedure steps: In-core Temperature Monitor.
Tier #
2 Group #
2 K/A #
017 2.1.20 Rating 4.6 Question 32 GIVEN:
The Unit 2 crew is performing the actions in FR-C.1, Response to Inadequate Core Cooling CETC are 900°F The crew is about to start an ECCS charging pump.
Following the start of the ECCS charging pump, INITIALLY the Core Exit Thermocouples (CET) indications will _________.
A. Decrease due to saturated steam forming a frothy two phase mixture.
B. Decrease due to superheated steam forming a frothy two phase mixture.
C. Increase due to saturated steam being forced out of the core.
D. Increase due to superheated steam being forced out of the core.
Proposed Answer: D. Increase due to superheated steam being forced out of the core.
Explanation:
A. Incorrect. Inititally CET indication will Increase, not decrease. As the core begins to fill, heat transfer from the fuel will cause fluid entering the core to form a frothy two phase mixture.
B. Incorrect. Same as A.
C. Incorrect. Steam exiting the core will be superheated.
D. Correct. Increase due to superheated steam exiting the core.
Technical
References:
F0.2; FR.C-1, Functional Recovery Core Cooling Background pg LPE-C, page 10 References to be provided to applicants during exam: Steam Tables Learning Objective: Describe the emergency operating procedure strategies for: (41700) SI Termination and SI Reinitiation Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.10 / 43.5 / 45.12
DCPP L111 NRC Exam 21 November, 2012 P a g e l 38 Examination Outline Cross-Reference Level RO Knowledge of bus power supplies to the containment iodine removal fans Tier #
2 Group #
2 K/A #
027 K2.01 Rating 3.1 Question 33 Iodine Removal Fan 1-E15 receives electrical power from:
A. PY-16 B. 4 kV Bus H C. MCC 12I D. SD-11 Proposed Answer: C. MCC12I Explanation:
A. Incorrect - MCC 12I powers Iodine Removal Fan 1-E15.
B. Incorrect - May believe its vital equipment.
C. Correct - MCC 12I powers Iodine Removal Fan 1-E15.
D. Incorrect - MCC 12I powers Iodine Removal Fan 1-E15.
Technical
References:
OIM J-1-1, Electrical Distribution Overview, Revision 28 OIM J-1-4, Non-vital Electrical Power Distribution Overview, Revision 28 LH3, Iodine Removal System, Revision 11 References to be provided to applicants during exam: None Learning Objective: State the power supplies to Iodine Removal System components. (40819)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 39 Examination Outline Cross-Reference Level RO Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment Purge System controls including: Radiation levels Tier #
2 Group #
2 K/A #
029 A1.02 Rating 3.4 Question 34 Containment Purge is in progress.
The following sequence of events occur:
AR PK02-06, CONTMT VENT ISOLATION, alarms It is determined that the alarm was due to Containment Purge Radiation Monitor, RE-44A, failing high The operator attempts to reset the CVI signal with RE-44A still in alarm.
When the operator presses the RESET pushbuttons on VB1 the signal resets:
A. but immediately occurs again once the operator releases the reset pushbuttons.
B. and is available to actuate again if RE-44B detects high radiation.
C. however, a subsequent high radiation condition will not cause isolation to occur.
D. however, the containment purge cannot be re-established until the signal is cleared.
Proposed Answer: C. however, a subsequent high radiation condition will not cause isolation to occur.
Explanation:
A. Incorrect. Resetting the Containment Ventilation Isolation signal without first clearing the condition(s) that brought in the alarm will inhibit automatic containment ventilation isolation from another high radiation signal.
B. Incorrect. Auto CVI is blocked.
C. Correct. Auto CVI from RE-44B is blocked.
D. Incorrect. With the signal reset, purge can be re-established.
Technical
References:
AR PK02-06, Contmt Vent Isolation; LH-4, Containment Purge System.
References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Containment Purge System. (5119)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.5 / 45.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 40 Examination Outline Cross-Reference Level RO Knowledge of design features which provide spent fuel pool cooling:
anti-siphon devices Tier #
2 Group #
2 K/A #
033 K4.03 Rating 3.1 Question 35 Which of the following design features of the Spent Fuel Pool (SFP) cooling system mitigates inadvertent draining of the SFP?
A. SFP pump low level trip B. SFP auto makeup C. Anti-siphon hole in the SFP cooling return piping D. SFP pump high discharge flow trip Proposed Answer: C. Anti-siphon hole in the SFP cooling return piping Explanation:
A. Incorrect - There is no low level trip for the pump (unlike other systems, such as the RHR low RWST level trip).
B. Incorrect - there is no auto makeup.
C. Correct - Per the lesson plan the anti-siphon device is on the cooling water return line and is 8 feet above the top of the spent fuel.
D. Incorrect - There is no high flow (essentially runout protection) trip for the pump.
Technical
References:
OIM 106713, Spent Fuel Cooling System, Sheet 3 LB-7, Spent Fuel Pool Cooling System References to be provided to applicants during exam: None Learning Objective: Explain significant Spent Fuel Pool Cooling System design features and the importance to nuclear safety. (40506)
Question Source:
Bank # L061 (6/2008) NRC exam #62 X
(note changes; attach parent)
Modified Bank #
New Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 41 Examination Outline Cross-Reference Level RO Fuel Handling Equipment. - Knowledge of physical connections and/or cause-effect relationships between Fuel handling system and the following system: RHR Tier #
2 Group #
2 K/A #
034 K1.02 Rating 2.5 Question 36 Unit 1 is in MODE 6.
Preparations are underway to fill the refueling cavity using a Containment Spray pump in accordance with OP B-2:II, RHR-Filling the Refueling Cavity.
During the fill, one train of RHR must be discharging to the reactor _______ legs to _________.
A. Hot; aid in maintaining refueling cavity clarity B. Cold; aid in maintaining refueling cavity clarity C. Hot; ensure circulation through the core D. Cold; ensure circulation through the core Proposed Answer: D. Cold; ensure circulation through the core Explanation:
A. Incorrect. In accordance OP B-2, hot leg flow may be used to maintain cavity clarity.
B. Incorrect. Clarity is not enhanced with flow through the cold legs.
C. Incorrect. Flow through the Hot Legs bypasses the lower reactor vessel and the core.
D. Correct. RHR should be injecting into the cold legs to ensure core cooling.
Technical
References:
OP B-2 RHR Filling the Refueling Cavity References to be provided to applicants during exam: None Learning Objective: Discuss significant precautions and limitations associated with the Fuel Handling System. (36965)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.2 to 41.9 / 45.7 to 45.8
DCPP L111 NRC Exam 21 November, 2012 P a g e l 42 Examination Outline Cross-Reference Level RO Knowledge of the effect that a loss or malfunction of the Steam Dump/Turbine Bypass Control will have on the Reactor Coolant System Tier #
2 Group #
2 K/A #
041 K3.02 Rating 3.8 Question 37 GIVEN:
Unit 2 has tripped by opening 23D and E Reactor Trip Breakers are closed Steam Dump Mode Select Switch (43/SDI) is selected to Tavg With no operator action, the steam dump valves should stabilize average reactor coolant temperature at approximately _______.
A. 543°F B. 547°F C. 550°F D. 555°F Proposed Answer: C. 550°F Explanation:
A. Incorrect - P-12 will cause the valves to close at 543°F but the Load Reject Controller will maintain reactor coolant temperature modulating at 550°F with no operator action.
B. Incorrect - Reactor coolant temperature tracks to 547°F when the steam dumps are controlled in Tavg mode by the Reactor Trip controller.
C. Correct - Reactor coolant temperature tracks to 550°F when the steam dumps are controlled in Tavg mode by the Load Reject controller in Unit 2.
D. Incorrect - Reactor coolant temperature tracks to 550°F when the steam dumps are controlled in Tavg mode by the Load Reject controller in Unit 2.
Technical
References:
OIM C-2-6, Steam Dump System Composite, Revision 29 LC-2B, Steam Dump System, Revision 13 References to be provided to applicants during exam:
None Learning Objective: 37812 - Describe Steam Dump System components.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 43 Examination Outline Cross-Reference Level RO Area Radiation Monitoring. Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including: Radiation levels Tier #
2 Group #
2 K/A #
072 A1.01 Rating 3.4 Question 38 A fuel assembly becomes damaged while being moved in the Spent Fuel Pool and RE-58, Spent Fuel Pool Area Radiation Monitor, alarms.
Which of the following automatic actions occur?
A. Fuel Handling Building (FHB) evacuation alarm sounds and FHB ventilation exhaust fan E-5 OR E-6 starts.
B. FHB and Containment evacuation alarms sound and FHB ventilation exhaust fans E-5 OR E-6 start.
C. FHB evacuation alarm sounds and FHB ventilation exhaust fans E-5 AND E-6 start.
D. FHB and Containment evacuation alarms sound and FHB ventilation exhaust fan E-5 AND E-6 starts.
Proposed Answer: A. Fuel Handling Building (FHB) evacuation alarm sounds and FHB ventilation exhaust fan E-5 OR E-6 starts.
Explanation:
A correct. The FHB building evacuation alarm sounds and the selected fan (E-5 or E-6) will start.
B incorrect. The containment evacuation alarm will not alarm.
C incorrect. Only one of the fans will start.
D incorrect. Only one fan starts and the containment evacuation alarm does not alarm.
Technical
References:
LH-7, page 15 References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the Radiation Monitoring System. (37875)
Question Source:
Bank # S-32157 X
(note changes; attach parent)
Modified Bank #
New Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.5 / 45.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 44 Examination Outline Cross-Reference Level RO Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: Change in leak rate with change in pressure Tier #
1 Group #
1 K/A #
AK1.02 Rating 3.1 Question 39 GIVEN:
A Pressurizer PORV fails open and cannot be isolated The plant trips and SI actuates 30 minutes after the reactor trip the crew enters E-1.2, Post LOCA Cooldown and Depressurization Which of the following describes the expected plant conditions as the crew enters E-1.2?
A. Break flow is unchanged from its original value; Pressurizer level off-scale high.
B. Break flow has decreased from its original value; Pressurizer level on scale and decreasing.
C. Break flow is unchanged from its original value; Pressurizer level on scale and decreasing.
D. Break flow has decreased from its original value; Pressurizer level off-scale high.
Proposed Answer:
D. Break flow has decreased from its original value; Pressurizer level off-scale high Explanation:
A. Incorrect - RCS pressure will be less than NOP as a result break flow will be reduced.
B. Incorrect - Pressurizer level will be off scale high (even if still on scale, it would be increasing due to increased SI flow and the open PORV).
C. Incorrect - Break flow will be reduced.
D. Correct - As RCS pressure lowers, the break flow rate will lower and due to the failed open PORV, pressurizer level will be off scale high.
Technical
References:
LMCD-FRC page 17 and page 37 References to be provided to applicants during exam: None Learning Objective: 41697 - Describe the plant response to a loss of reactor coolant including: Vapor Space LOCAs Question Source:
Bank # DC 2010-01 Question #39 X
(note changes; attach parent)
Modified Bank #
New Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 45 Examination Outline Cross-Reference Level RO Knowledge of the interrelations between the small break LOCA and the Steam Generators Tier #
1 Group #
1 K/A #
EPE 009 EK2.03 Rating 3.0 Question 40 GIVEN:
A Small Break LOCA has occurred in Containment RCS pressure is 1500 psig and slowly lowering Only Train A of ECCS has actuated All MSIVs are closed All RCPs are running All steam generators are available A secondary heat sink ______ required to maintain adequate core cooling because _____.
A. is not; core cooling is provided by the operation of the RCPs B.
is not; core cooling is provided by ECCS flow C.
is; core cooling is provided by the secondary heat sink to allow ECCS flow to equal break flow D. is; secondary heat sinks are required for all LOCA events Proposed Answer:
C. is; core cooling is provided by the secondary heat sink to allow ECCS flow to equal break flow Explanation:
For breaks in this category, the establishment of an equilibrium pressure where pumped SI equals break flow constitutes a safe and stable condition for the long term, provided that the steam generator heat sink is maintained until such time that the break flow and SI sensible heat can remove all the decay heat. Once equilibrium pressure was established, the core was covered and adequate flow existed to remove decay heat through the steam generator with a small amount of voiding. The only change in the primary system conditions through the transient for these cases is a gradual decrease in fluid temperatures which is beneficial, since it indicates that adequate core cooling is being maintained.
A. Incorrect - RCPs only transport heat from the core to the steam generators B. Incorrect - a secondary heat sink is required to remove sufficient heat to lower primary pressure to allow for ECCS flow to equal leak rate.
C. Correct core cooling is provided by the secondary heat sink to allow ECCS flow to equal break flow D. Incorrect - secondary heat sink is not required for a LBLOCA due to the primary completely depressurizes during the event Technical
References:
WOG Background Information, E-1 Loss of Reactor or Secondary Coolant LPE-1A - Loss of Coolant Response, Revision 13 References to be provided to applicants during exam: None Learning Objective: 7920 - Explain basis of emergency procedure step Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA
DCPP L111 NRC Exam 21 November, 2012 P a g e l 46 Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 47 Examination Outline Cross-Reference Level RO Knowledge of the reasons for the following responses as the apply to the Large Break LOCA:
Hot-leg injection/recirculation Tier #
1 Group #
1 K/A #
EK3.13 Rating 3.8 Question 41 GIVEN:
A LOCA occurred on Unit 1 about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> ago RCS pressure is 25 psig RWST level is 22%
Procedure E-1.3, Transfer to Cold Leg Recirculation, is in effect Based on these conditions, the next operator action and the reason for it is to:
A. Maintain Cold Leg Recirculation to mitigate the effects of boric acid precipitation in the core.
B. Maintain Cold Leg Recirculation until the Refueling Water Storage Tank level is less than 4% to maximize the water inventory in the Containment sump.
C. Transfer to Hot Leg Recirculation to mitigate the effects of boric acid precipitation in the core.
D. Transfer to Hot Leg Recirculation to prevent a reverse flow through the core as the flow rate out of the break exceeds the injection flow rate.
Proposed Answer:C. Transfer to Hot Leg Recirculation to mitigate the effects of boric acid precipitation in the core.
Explanation:
A. Incorrect. During Cold leg recirculation, core boil-off will concentrate boric acid in the core region.
Hot leg recirculation will remix boron to a more even distribution in the cooling water flow. This mitigates the effects of boric acid precipitation in the core.
B. Incorrect. Swapover to Hot Leg Recirc begins at 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the LOCA, per E-1, step 20, Loss of Reactor or Secondary Coolant, transition to Hot Leg Recirculation (and is aligned after 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />).
C. Correct. Hot Leg Recirculation prevents boric acid precipitation and quenches the steam bubble in the Reactor vessel head.
D. Incorrect. Hot Leg Recirculation establishes a reverse flushing flow through the core.
Technical
References:
LPE-1C pages 2,4,5,18; LMCDFRC pages 4,16; E-1 pages 1,19; 1.4 page 1.
References to be provided to applicants during exam:
None Learning Objective: 8914-Explain the reason for transferring to hot leg recirculation after a LOCA.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR 41.5 / 41.10 / 45.6 / 45.13 Added RWST level is 22% to add credence to B.
DCPP L111 NRC Exam 21 November, 2012 P a g e l 48
DCPP L111 NRC Exam 21 November, 2012 P a g e l 49 Examination Outline Cross-Reference Level RO Ability to determine and interpret when to secure RCPs on high stator temperature as it applies to Reactor Coolant Pump Malfunctions (Loss of RC Flow).
Tier #
1 Group #
1 K/A #
APE 015/17 AA2.09 Rating 3.4 Question 42 The plant is operating at 100% power.
Annunciator Response Procedure AR PK05-01, RCP NO. 11, directs the operator to manually trip the reactor and RCP 11, if not already tripped, after getting Shift Foreman concurrence if:
A. Motor bearing temperature is 175°F.
B. Seal injection flow is lost.
C. Motor stator temperature is 310°F.
D. CCW flow to the thermal barrier heat exchanger is lost.
Proposed Answer: C. Motor stator temperature is 310°F.
Explanation:
A. Incorrect - A trip is required if motor bearing temperature is greater than 200 degrees.
B. Incorrect - Not required if CCW cooling to thermal barrier exists.
C. Correct - In accordance with AR PK05-01, Section 2.7, RCP 1-1 High Temperature PPC, requires that if the temperature is at or above 300°F then trip the RCP following manual trip of the reactor.
D. Incorrect - Not required provided there is seal injection.
Technical
References:
AR PK05-01, RCP No. 11, Revision 33 LAR-1, RCP Failures, Revision 11 References to be provided to applicants during exam: None Learning Objective: 7927 - Given initial conditions and assumptions, determine if a reactor trip or safety injection is required.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 50 Examination Outline Cross-Reference Level RO Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. 000022 Loss of Rx Coolant Makeup/2 Tier #
1 Group #
1 K/A #
022 2.4.21 Rating 4.0 Question 43 Which of the following would have the greatest impact on the mitigation of a small break LOCA caused by a failed open Pressurizer PORV and possibly challenge the Core Cooling Critical Safety Function?
A. Loss of both RHR pumps B. Loss of both ECCS charging pumps C. Accumulator isolation valves are closed D. RCS pressure remains above steam generator pressure Proposed Answer: B. Loss of both ECCS charging pumps Explanation:
A. Incorrect. this would have an impact on long term core cooling for a large LOCA B. Correct. With no injection, inventory would continue to be lost, possibly leading to a challenge to core cooling. This is the most probable cause and a failed PORV is the most probable scenario (or initiator).
C. Incorrect. Accumulators do not inject and are isolated as part of the small break LOCA mitigation in E-1.2 D. Incorrect. RCS pressure will remain above steam generator pressure for a small break to ensure heat removal from the RCS by the steam generators.
Technical
References:
LMCDFRC-Mitigating Core Damage-Core Cooling, page 26 References to be provided to applicants during exam: None Learning Objective: Differentiate between:
- a.
Adequate Core Cooling
- b.
Degraded Core Cooling
- c.
Inadequate Core Cooling Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.7 / 43.5 / 45.12
DCPP L111 NRC Exam 21 November, 2012 P a g e l 51 Examination Outline Cross-Reference Level RO Ability to operate and/or monitor the LPI pump control switch, indicators, ammeter, running light and flow meter as they apply to the Loss of Residual Heat Removal System.
Tier #
1 Group #
1 K/A #
APE 025 AA1.09 Rating 3.2 Question 44 GIVEN:
Five days ago Unit 1 entered MODE 6 RHR pump 1-1 is aligned for shutdown cooling FI-970A, RHR flow from RHR heat exchanger 1-1, decreases to 0 gpm RHR pump current is 15 amps and oscillating The Red light for FCV-641A, RHR pump 11 Recirc valve, is lit The crew has entered AP SD-5, Loss of Residual Heat Removal What action, if any, should the operator take for RHR pump 1-1?
A. Secure the pump, there is no flow through the pump if there is no flow indication on FI-970.
B. Secure the pump, low amps indicate insufficient flow through the pump.
C. No action, pumps amps and FCV-641A indication, are consistent with adequate recirc flow through the pump.
D. No action, as long as FCV-641A is open, there will be sufficient flow through the pump.
Proposed Answer: B. Secure the pump, low amps indicate insufficient flow through the pump.
Explanation:
A. Incorrect - FI-970A is downstream of the heat exchanger and where the recirc valve connection is located. If there is no flow going to the RCS, but there is recirc flow, FI-970A will read 0 gpm.
B. Correct. Recirc flow inidication is amps in the normal band (greater than 28 amps) AND the recirc valve open (red light). Low amps indicates very little flow and the pump should be secured.
C. Incorrect. there is not adequate flow and it should secured. If there was adequate flow, current would be greater than 28 amps.
D. Incorrect - there must also be indication of normal pump amps.
Technical
References:
LB-2, Residual Heat Removal System, Revision 16 AP SD-5, Loss of Residual Heat Removal, Revision 9A References to be provided to applicants during exam: None Learning Objective: 7050 - Discuss abnormal conditions associated with the RHR system Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 52 Examination Outline Cross-Reference Level RO Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: RCP injection flow Tier #
1 Group #
1 K/A #
AA2.14 Rating 2.8 Question 45 GIVEN:
Unit 2 is at 100% power Charging is in MANUAL The controlling Pressurizer pressure channel fails high.
Based on these conditions, initially charging flow to the Regen heat exchanger will ______ and seal injection flows to each of the RCPs will _______:
A. decrease: increase B. decrease; not change C. increase; increase D. increase; not change Proposed Answer: C. increase; increase Explanation:
A. Incorrect. Seal Injection is supplied by the CVCS System at 8 to 13 gpm per RCP. The flow rate is normally adjusted by throttling HCV-142 to divert charging flow to the seals. The purpose of HCV-142 is to create sufficient backpressure in the charging line to ensure that adequate flow is maintained through the RCP seal water injection line upstream of valve HCV-142. PT-455 failing high opens the spray valves (and two PORVs), which come off of cold legs 1 and 2. When spray is initiated charging RCS decreases and pressure on the charging pump side of HCV-142 lowers; thereby changing the backpressure of HCV-142. Charging will increase to the seals and the heat exchanger.
B. Incorrect - RCP seal injection will increase, because differential pressure across HCV-142 will increase, increasing total charging flow.
C.
Correct - The lower RCS pressure will INCREASE total charging, causing increased flow to the seals and the heat exchanger.
D. Incorrect - Seal injection will increase due to increased DP across HCV-142 will increase.
Technical
References:
LB-1A page 48, LA-6 page 13, OIM A-4-4B Rev 24, OIM B-1-1 Rev 27 References to be provided to applicants during exam: None Learning Objective: 40449 - Discuss abnormal conditions associated with the CVCS Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 53 Examination Outline Cross-Reference Level RO Knowledge of the operation implications of the following concepts as they apply to ATWS: Effects of boron on reactivity Tier #
1 Group #
1 K/A #
EPE 029 EK1.03 Rating 3.6 Question 46 In accordance with FR-S.1, Response to Nuclear Power Generation, ATWS, the reason boron is added to the reactor is to counter the _________________ during an ATWS.
A. decrease in core temperature B. decrease in reactor coolant temperature C. increase in RCS pressure D. rods not being fully inserted Proposed Answer:
D rods not being fully inserted Explanation:
A. Incorrect - Decrease in core temperatures result in a positive reactivity addition but the concern is the rods and the SDM B. Incorrect - Decrease in the reactor coolant temperatures will result in a positive reactivity addition but the concern is the rods and the SDM.
C. Incorrect - Increase in RCS pressure result in a negative reactivity addition verses a positive reactivity addition D. Correct - Boron addition provides negative reactivity to ensure shutdown margin is established following a ATWS Technical
References:
E-0, Reactor Trip or Safety Injection, Rev. 31 FR-S.1, Response to Nuclear Power Generation / ATWS, Rev. 13A LCMDFRS, Mitigating Core Damage - Subcriticality, Revision 6 WOGBD, FR-S.1, Response to Nuclear Power Generation / ATWS References to be provided to applicants during exam: None Learning Objective: 7920L - Explain basis of emergency procedure steps (FR-S series)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.8
DCPP L111 NRC Exam 21 November, 2012 P a g e l 54 Examination Outline Cross-Reference Level RO Knowledge of the interrelations between the Steam Line Rupture and the following: Valves Tier #
1 Group #
1 K/A #
AK2.01 Rating 2.6 Question 47 Unit 2 was in MODE 3. RCS pressure was 1900 psig.
A main steam line rupture inside Containment occurs.
Current plant conditions:
Containment pressure is 18.5 psig and rising Steam Generator (SG) pressure is 850 psig and lowering at a rate of 125 psi/minute Pressurizer pressure is 1800 psig and lowering SG-21 level swelled to 92% and is now 88% and lowering The operator observes the Main Steam Isolation Valves (MSIVs) are stroking closed.
The reason the MSIVs are closing is:
A. high steam generator water level.
B. low steam generator pressure.
C. high containment pressure.
D. rapidly lowering SG pressure.
Proposed Answer: D. rapidly lowering SG pressure Explanation:
A. Incorrect. High water level causes a feedwater isolation (P-14) but not a SG isolation.
B. Incorrect. A low steam generator pressure will cause safety injection and isolation if less than 600 psig ABOVE P-11.
C. Incorrect. High containment pressure will cause the MSIVs to close if greater than 22 psig.
D. Correct. MSIVs are closing because the high rate of SG pressure change exceeds 2 psi/sec over 50 seconds, or 100 psi/minute.
Technical
References:
LB-6A, B-6A, page 2.2-31, OIM B-6-2 References to be provided to applicants during exam: None Learning Objective: 7340 - Analyze automatic features and interlocks associated with the main steam system Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis 10CFR Part 55 Content:
CFR 41.7 / 45.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 55 Examination Outline Cross-Reference Level RO Ability to operate and monitor the following as they apply to a Station Blackout: Battery approaching fully discharged Tier #
1 Group #
1 K/A #
EPE 055 EA1.05 Rating 3.3 Question 48 The crew is performing the actions of ECA-0.0, Loss of All Vital AC.
The operator reports that voltages on the vital DC buses are approaching 110 VDC.
What is the concern of allowing 125 VDC battery voltage to decrease below 110 VDC?
A. Bus voltage may decrease to a voltage that supplied loads may not function.
B. Low bus voltage causes the battery output breaker to open on undervoltage.
C. The battery reverses polarity and damages remaining DC loads.
D. Battery current decreases rapidly and damages remaining DC loads.
Proposed Answer: A. Bus voltage may decrease to a voltage that supplied loads may not function.
Explanation:
A. Correct. Per Caution in ECA-0.0, Batteries may be irreversibly damaged and supplied loads may not function as battery voltages approach their end voltage (105 VDC for 125 VDC batteries; 210 VDC for 250 VDC batteries).
B. Incorrect. There is not an undervoltage trip on the output breaker.
C. Incorrect - The battery could be damaged.
D. Incorrect - current is controlled by the loads Technical
References:
ECA-0.0 References to be provided to applicants during exam: None Learning Objective: 7920G - Explain basis of emergency procedure steps (ECA-0 series)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 56 Examination Outline Cross-Reference Level RO Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation:
000056 Loss of Off-site Power / 6 Tier #
1 Group #
1 K/A #
2.1.7 Rating 4.4 Question 49 Unit 1 is at 100% power when a loss of all off-site power occurs. Only emergency diesel generator 1-3 starts and supplies its vital 4 KV bus.
Which of the following equipment would have power?
A. Condensate Booster Pump 1-1 and Component Cooling Water Pump 1-3 B. Component Cooling Water Pump 1-3 and Auxiliary Feedwater Pump 1-3 C. Residual Heat Removal Pump 1-1 and ECCS Centrifugal Charging Pump 1-2 D. Component Cooling Water Pump 1-1 and Auxiliary Saltwater Pump 1-1 Proposed Answer: D. Component Cooling Water Pump 1-1 and Auxiliary Saltwater Pump 1-1 Explanation:
A. Incorrect. Condensate Booster Pump 1-1 is powered from nonvital bus D. CCW Pump 1-3 is powered from Bus H, which is powered from DG 1-1.
B. Incorrect. CCW Pump 1-3 is powered from Bus H, which is powered from DG 1-1.
C. Incorrect. RHR Pump 1-1 and CCP 1-2 are powered from Bus G, which is energized from DG 1-2.
D. Correct. CCW Pump 1-1 and ASW Pump 1-1 are powered from Bus F, which is energized by DG 1-
- 3.
Technical
References:
LJ-6A 4KV System, AOP-26 Loss of Offsite Power, LPA-26 Loss of Offsite Power, OIM J-1-1 References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the 4 kV System. (41081)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.5 / 43.5 / 45.12 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 57 Examination Outline Cross-Reference Level RO Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: Actions contained in for loss of vital ac electrical instrument bus Tier #
1 Group #
1 K/A #
APE 057 AK3.01 Rating 4.1 Question 50 Unit 2 is at 30% power.
Power is lost to PY-21, Vital AC Instrument Bus. As a result, Power Range channel, N-41, and Turbine Impulse Pressure channel, PT-505 have failed low.
Which of the following describes why the operators must place Rod Control in MANUAL?
A. PT-505 failure is causing rods to insert.
B. NI-41 failure is causing rods to insert.
C. PT-505 failure is causing rods to withdraw.
D. NI-41 failure is causing rods to withdraw.
Proposed Answer: A. PT-505 failure is causing rods to insert.
Explanation:
A. PT-505 low causes a Tave/Tref mismatch. Rods begin to insert in an attempt to match Tave with the failed low Tref.
B. NI-41 fails low, but does not cause rods to move (auctioneered high).
C. PT-505 fails low, the mismatch is the opposite.
D. NI-41 failing low, could think that rods will withdraw to match turbine with reactor power. In fact, the failing low of the NI will not affect rod control.
Technical
References:
LPA-4 References to be provided to applicants during exam: None Learning Objective: 4274 - Explain the consequences of loss of vital instrument bus Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.5, 55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 58 Examination Outline Cross-Reference Level RO Knowledge of the reasons for the following responses as they apply to the Loss of DC Power:
Actions contained in for loss of dc power Tier #
1 Group #
1 K/A #
AK3.02 Rating 4.0 Question 51 The reason the operator must ensure the main feedwater pumps are immediately tripped following a loss of DC bus 12 is:
A. in anticipation of AFW being in service following a reactor trip.
B. to prevent damage to the feed pumps from over speeding.
C. to prevent over pressurization of the feedwater system.
D. to prevent over filling the steam generators.
Proposed Answer: C. to prevent over pressurization of the feedwater system.
Explanation:
A. Incorrect. A Caution in AP-23 states, MFW pump recirc valves fail closed and MFP trip solenoids lose power, immediate action is required to runback and locally trip running MFPs to prevent over-pressurization.
B. Incorrect. A Caution in AP-23 states, MFW pump recirc valves fail closed and MFP trip solenoids lose power, immediate action is required to runback and locally trip running MFPs to prevent over-pressurization.
C. Correct. A Caution Note in AP-23 gives the following information, MFW pump recirc valves fail closed and MFP trip solenoids lose power, immediate action is required to runback and locally trip running MFPs to prevent over-pressurization.
D. Incorrect. A Caution in AP-23 states, MFW pump recirc valves fail closed and MFP trip solenoids lose power, immediate action is required to runback and locally trip running MFPs to prevent over-pressurization.
Technical
References:
AP-23 Loss of Vital DC Bus, E-0 Reactor Trip or Safety Injection, Lesson LPA-23 Loss of Vital DC Bus, ARP PK20-18 125 VDC Bus 11, 12, or 13 Undervoltage References to be provided to applicants during exam: None Learning Objective: Given an abnormal condition, summarize the major actions of the abnormal operating procedure to mitigate an event in progress. (3477)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR 41.5,41.10 / 45.6 / 45.1
DCPP L111 NRC Exam 21 November, 2012 P a g e l 59 Examination Outline Cross-Reference Level RO Ability to identify and interpret diverse indications to validate the response of another indication: Loss of Nuclear Service Water (ASW)
Tier #
1 Group #
1 K/A #
APE 062 2.1.45 Rating 4.3 Question 52 The running ASW Pump trips due to over current. The standby ASW Pump fails to start.
Which of the following annunciators would the operator expect to alarm?
A. PK01-01, ASW Sys HX Delta P/HDR Press, and PK01-03, Aux Salt Water Pumps B. PK01-01, ASW Sys HX Delta P/HDR Press, and PK01-02, Aux Salt Water PPS Room C. PK01-02, Aux Salt Water PPS Room, and PK13-01, Bar Racks/Screens D. PK01-03, Aux Salt Water Pumps, and PK13-01, Bar Racks/Screens Proposed Answer:
A. PK01-01, ASW Sys HX Delta P/HDR Press, and PK01-03, Aux Salt Water Pumps Explanation:
A. Correct - PK01-01 actuates on ASW to CCW HX low pressure and PK01-03 actuates on a pump failure B. Incorrect - PK01-02 does not actuate on a loss of ASW Pump. Actuates on Aux Salt Water Pump Room adverse conditions.
C. Incorrect - PK01-02 does not actuate on a loss of ASW Pump and PK13-01 does not actuate on low ASW flow.
D. Incorrect - PK13-01 does not actuate on low ASW flow.
Technical
References:
AP PK01-01, ASW sys HX Delta P/HDR Press, Revision 21 AP PK01-02, Aux Salt Water PPS Room, Revision 14 AP PK01-03, Aux Salt Water Pumps, Revision 15 LPA-10, Loss of Auxiliary Salt Water, Revision 11 LE-5, Auxiliary Salt Water System, Revision 11 References to be provided to applicants during exam: None Learning Objective: 5330 - Describe controls, indications, and alarms associated with the ASW System Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.7, 55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 60 Examination Outline Cross-Reference Level RO Knowledge of annunciator alarms, indications, or response procedures:
000065 Loss of Instrument Air / 8 Tier #
1 Group #
1 K/A #
2.4.31 Rating 4.2 Question 53 The plant is at 100% power and the Plant Instrument Air alarm is illuminated. Air Header Pressure indicates 92 psig and is slowly lowering.
Based on these conditions, AR PK13-16 directs the crew to:
A. Trip the reactor.
B. Verify Reciprocating Air Compressors 0-1 through 0-4 standby start has occurred and that the standby start light (Blue) is lit.
C. Press the Instrument Air Compressor standby start relay RESET button.
D. When air header pressure reaches 90 psig, check if letdown has isolated and place excess letdown in service if necessary.
Proposed Answer: B. Verify Reciprocating Air Compressors 0-1 through 0-4 standby start has occurred and that the standby start light (Blue) is lit.
Explanation:
A. Incorrect. In accordance with ARP PK13-16 if at any time instrument air head drops below 90 psig or is 93 psig and not rising, then go to AP-9 Loss of Instrument Air. The reactor is tripped if there is a loss of control, for instance, the main Feedwater reg valves fail closed.
B. Correct. In accordance with ARP PK13-16 section 2.1.2.a Verify Air Compressors 0-1 through 0-4 standby start has occurred and that the standby start light (Blue) is lit.
C. Incorrect. In accordance with ARP PK13-16, pressing the reset button should be performed per step 2.1.2c After instrument air header pressure has recovered to normal (100 psig or greater).
D. Incorrect. In accordance with ARP PK 13-16, various actions are required for Air Header Pressure
<90 psig, at 93 psig, from 94-100 psig, and over 100 psig. Letdown will isolate when instrument air to containment is isolated (about 85 psig).
Technical
References:
ARP PK 13-16 References to be provided to applicants during exam:
None Learning Objective: 7226 - Describe controls, indications, and alarms associated with the Compressed Air System Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.10 / 45.3
DCPP L111 NRC Exam 21 November, 2012 P a g e l 61 Examination Outline Cross-Reference Level RO Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments Tier #
1 Group #
1 K/A #
EA2.2 Rating 3.4 Question 54 The crew is performing step 6, of ECA-1.1, Loss of Emergency Recirculation, "Determine Containment Spray Requirements.
In order to determine the number of Containment Spray pumps required for containment heat removal, the crew will check containment pressure,:
A. Containment Recirculation Sump level and number of containment fan coolers running.
B. Refueling Water Storage Tank level and number of containment fan coolers running.
C. Refueling Water Storage Tank level and hydrogen concentration.
D. Containment Recirculation Sump level and hydrogen concentration.
Proposed Answer:
B. Refueling Water Storage Tank level and number of containment fan coolers running Explanation:
A. Incorrect - Containment Spray pumps do not take a suction off of the Containment Recirculation Sump B. Correct - ECA-1.1, Step 6 requires RWST level, containment pressure and number of containment coolers running to determine the required number of containment spray pumps.
C. Incorrect - RWST does supply the containment spray pump but, number of required containment spray pumps is affected by the total number of containment fan coolers running not just available.
D. Incorrect - Containment Spray pumps do not take a suction off of the Containment Recirculation Sump, number of required containment spray pumps is affected by the total number of containment fan coolers running not just available.
Technical
References:
ECA-1.1, Loss of Emergency Recirculation, Revision 24 Background Documents WOG Emergency Response Guidelines, ECA-1.1, Loss of Emergency Coolant Recirculation, Revision 2 References to be provided to applicants during exam: None Learning Objective: 20400 - Determine containment spray requirements during response to LOCA with loss of ECR capabilities Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 62 Examination Outline Cross-Reference Level RO Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following:
EK2.2 Facilitys 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.
Tier #
1 Group #
1 K/A #
EK2.2 Rating 3.9 Question 55 GIVEN:
Unit 1 operators are establishing RCS bleed and feed in accordance with FR-H.1, Loss of Secondary Heat Sink While verifying RCS bleed path per step 18, the Reactor Operator observes that ONLY one of the Pressurizer PORVs was able to be opened Which describes the appropriate action in accordance with FR-H.1, Loss of Secondary Heat Sink, and the reason for this response?
A. Open the Reactor Vessel Head Vent valves because the RCS may not depressurize sufficiently to permit adequate SI flow to remove core decay heat.
B. Close the open PORV and continue efforts to restore AFW flow because one PORV will not depressurize the RCS sufficiently to allow SI to maintain RCS inventory.
C. No action is required because the RCS will still depressurize sufficiently with one PORV open to permit adequate SI flow to remove core decay heat.
D. Close the open PORV, then open the Reactor Vessel Head Vent valves to restrict the mass loss sufficiently to ensure RCS inventory can be maintained with SI.
Proposed Answer: A. Open the Reactor Vessel Head Vent valves because the RCS may not depressurize sufficiently to permit adequate SI flow to remove core decay heat.
Explanation:
A. Correct. Open the Reactor Vessel Head Vent valves to allow the RCS to depressurize sufficiently to permit adequate feed of subcooled SI flow to remove core decay heat.
B. Incorrect. The open PORV is to remain open and the RV Head Vent Valves are to be opened to provide additional RCS depressurization.
C. Incorrect. One PORV may not depressurize the RCS sufficiently to allow enough SI flow tomaintain inventory and remove decay heat.
D. Incorrect. The procedure does not direct the open PORV to be closed. It must remain open to provide RCS depressurization along with the RV head vent valves.
Technical
References:
FR-H.1, Loss of Secondary Heat Sink, Background Information for Westinghouse Owners Group Emergency Response Guideline FR-H.1, Operations Lesson: LMCDFRH References to be provided to applicants during exam: None Learning Objective: Explain the plant response to bleed and feed operations. (3817)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
DCPP L111 NRC Exam 21 November, 2012 P a g e l 63 Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.7 / 45.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 64 Examination Outline Cross-Reference Level RO Ability to operate and//or monitor the following as they apply to Generator Voltage and Electrical Grid Disturbances: Grid frequency and voltage Tier #
1 Group #
1 K/A #
APE 077 AA1.01 Rating 3.6 Question 56 GIVEN:
Unit 1 and 2 are operating at 100%
The series capacitors are NOT in service The main generator voltage regulation is in automatic A grid disturbance results in both Diablo units remaining as one of the few remaining generation sources.
Load on the grid is unchanged.
Based on these conditions, initially, grid voltage will ___________ and grid frequency will A. Decrease; increase B. Decrease; decrease C. Increase; increase D. Increase; decrease Proposed Answer: B. Decrease; decrease Explanation:
A. Incorrect - Increased line current will result in voltage drop due to E=IR, lower voltage, however, frequency will also.
B. Correct - Both will lower.
C. Incorrect - Grid Voltage and frequency will drop.
D. Incorrect - Grid Voltage will drop.
Technical
References:
LJ2VIII LPA-26, Loss of Offsite Power, Revision 7 J4A, Main Generator, Revision 17 References to be provided to applicants during exam: None Learning Objective: Recognize the circumstances which could lead to grid instability. (35626)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.5, 55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 65 Examination Outline Cross-Reference Level RO Ability to determine and interpret the following as they apply to the Emergency Boration:
AA2.04 Availability of BWST Tier #
1 Group #
2 K/A #
APE 024 AA2.04 Rating 3.4 Question 57 GIVEN:
Unit 1 has tripped from 100% power RCS Tave is 547°F 4 Control Rods are stuck out Boric Acid Storage Tank #1 is in service and level is 93%
Boric Acid Storage Tank #2 level is 92%
Following completion of the minimum required per AP-6, Emergency Boration, Boric Acid Storage Tank #1 level will be between _____:
A. 37% and 38%, which meets the minimum required level for plant conditions.
B. 38% and 39%, which does NOT meet the minimum required level for plant conditions.
C. 38% and 39%, which meets the minimum required level for plant conditions.
D. 37% and 38%, which does NOT meet the minimum required level for plant conditions.
Proposed Answer: D. 37% and 38%, which does NOT meet the minimum required level for plant conditions.
Explanation:
A. Incorrect. This level does not meet the Technical Specification requirement of 14042 gallons.
B. Incorrect. Level will be 3882 gallons, which is between 37% and 38%.
C. Incorrect. Level will be between 37% and 38% and is less than the requirement of 14, 042 gallons.
D. Correct. 4 rods require 900 gallons of boration each. Current level is 7482 gallons - 3600 gallons = 3882 gallons, which is between 37% and 38%. Combine this with Tank #2s 7352 gallons is 11234 gallons, which is less than the minimum required by Technical Specifications of 14042 gallons.
Technical
References:
STP C-20 BAST Usable Volume, attachment 2, ECG 8.9, AP-06 Emergency Boration References to be provided to applicants during exam: STP C-20 BAST Usable Volume, attachment 2, Boric Acid Storage Tanks Volume Data, ECG 8.9 Learning Objective: Explain the Emergency Boration process (4149) 66041 - Apply the requirements of System 8 ECGs Question Source:
Bank #
(note changes; attach parent)
New X
Modified Bank #
Question History:
Last NRC Exam N/A No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 66 Examination Outline Cross-Reference Level RO Ability to explain and apply system limits and precautions: Related to Loss of Condenser Vacuum Tier #
1 Group #
2 K/A #
APE 051 2.1.32 Rating 3.8 Question 58 GIVEN:
Unit 1 is currently at 47% power and reducing power to 45%
AR PK10-11, Condenser Pressure/Level is alarming Indicated condenser pressure on VB3 is 7.8 Hg and stabilizing No automatic actions have occurred With these conditions, AP-7, Degraded Condenser, directs the crew to trip the:
A. Turbine, Reactor, AND MFW Pump(s)
B. Turbine AND MFW Pump(s) ONLY C. Turbine AND Reactor ONLY D. Turbine ONLY Proposed Answer:
D. Turbine ONLY Explanation:
A. Incorrect - Reactor trip not required because power is less than the P-9 setting (50%) and the MFW pump trip is at 10 Hg B. Incorrect - MFW pump trip is at 10 Hg C. Incorrect - Reactor trip not required because power is less than the P-9 setting (50%)
D. Correct. The turbine is tripped. The reactor is not tripped per the procedure.
Technical
References:
AP-7, Degraded Condenser, Revision 39 AR-PK10-11, Condenser Pressure/Level, Revision 18 LPA-7, Degraded Condenser, Revision 11 References to be provided to applicants during exam: AP-7, Attachment 6.2 Learning Objective: 7968 - Given initial conditions and assumptions, determine if a turbine trip is required Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 67 Examination Outline Cross-Reference Level RO Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the following: Auxiliary building ventilation system Tier #
1 Group #
2 K/A #
APE 060 AK2.02 Rating 2.7 Question 59 PK11-25, PLANT VENT RADIATION and PK11-21, HIGH RADIATION alarm. The source of the radiation is unknown.
The SFM directs operators to enter AP-14, Tank Ruptures.
In order to minimize the release, the crew should ______________.
A. secure the Auxiliary Building Ventilation System.
B. place the Auxiliary Building Ventilation System in the BUILDING AND SAFEGUARDS mode.
C. place the Auxiliary Building Ventilation System in the SAFEGUARDS ONLY mode with an S signal with charcoal heaters off.
D. place the Auxiliary Building Ventilation System in the SAFEGUARDS ONLY mode with an S signal with charcoal heaters energized.
Proposed Answer: D. place the Auxiliary Building Ventilation System in the SAFEGUARDS ONLY mode with an S signal with charcoal heaters energized.
Explanation:
A. Incorrect. The Auxiliary Building Ventilation System is placed into the SAFE GUARDS only mode.
B. Incorrect. While the lineup of the Auxiliary Building Ventilation System will momentarily pass through the BUILDING-and-SAFE GUARDS mode, this is not the desired condition and AP-14 directs operators to continue shifting the lineup to SAFE GUARDS only to minimize the release or radiation.
C. Incorrect. The charcoal filters will be energized to reduce relative humidity to less than 70% to improve iodine absorption in the charcoal filters.
D. Correct. Places the Aux Bldg Ventilation in the SAFEGUARDS ONLY Mode, discharging through the Charcoal filters.
References to be provided to applicants during exam: None Technical
References:
LPA-14, Tank Ruptures Learning Objective: Given an abnormal condition, summarize the major actions of the abnormal operating procedure to mitigate an event in progress. (3477)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR 41.7 / 45.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 68 Examination Outline Cross-Reference Level RO Knowledge of the reason for the following responses as they apply to High Containment Pressure: Manipulation of controls required to obtain desired operating results during abnormal and emergency situations.
Tier #
1 Group #
2 K/A #
W/E 14 EK3.3 Rating 3.5 Question 60 GIVEN:
The Shift Foreman has entered FR-Z.1, Response to High Containment Pressure All steam generators are faulted The minimum feed flow specified in FR-Z.1 for this condition is __________ and the reason for this minimum feed flow is to _________.
A. 435 gpm; provide sufficient flow to meet heat sink requirements B. 435 gpm; prevent dryout of the steam generator in order to minimize thermal shock to the steam generator C. 25 gpm per steam generator; provide sufficient flow to meet heat sink requirements D. 25 gpm per steam generator; prevent dryout of the steam generator in order to minimize thermal shock to the steam generators Proposed Answer: D. 25 gpm per steam generator; prevent dryout of the steam generator in order to minimize thermal shock to the steam generators Explanation:
A. Incorrect - E-1, Loss of Reactor or Secondary Coolant requires a minimum feedflow of 435 GPM to an intact steam generator (stem states all steam generators are faulted). WOG BD, E-I, Loss of Reactor or Secondary Coolant, states that the minimum feed flow requirement satisfies the feed flow requirement of the Heat Sink Status Tree until level in at least one SG is restored into the narrow range.
B.
Incorrect - E-1, Loss of Reactor or Secondary Coolant requires a minimum feedflow of 435 GPM to intact steam generator (stem states all steam generators are faulted). WOG BD, FR-Z.1, Response to High Containment Pressure, states that 25 GPM maintains the steam generator in a wet condition, thereby minimizing any thermal shock effects if feed flow is increased.
C.
Incorrect - FR-Z.1, Caution 2, states, if all steam generators are faulted, at least 25 GPM should be maintained to each steam generator. E-1, Loss of Reactor or Secondary Coolant, WOG BD, FR-Z.1, Response to High Containment Pressure, states that 25 GPM maintains the steam generator in a wet condition, thereby minimizing any thermal shock effects if feed flow is increased.
D. Correct - FR-Z.1, Caution 2, states, if all steam generators are faulted, at least 25 GPM should be maintained to each steam generator. WOG BD, FR-Z.1, Response to High Containment Pressure, states that 25 GPM maintains the steam generator in a wet condition, thereby minimizing any thermal shock effects if feed flow is increased.
Technical
References:
E-1, Loss of Reactor or Secondary Coolant, Revision 25A FR-Z.1, Response to High Containment Pressure, Revision 10 WOG BD, E-1, Loss of Reactor or Secondary Coolant, Revision 2 WOG BD, FR-Z.1, Response to High Containment Pressure, Revision 2 References to be provided to applicants during exam: None Learning Objective:
7920Q - Explain basis of emergency procedure steps (FR-Zs)
Question Source:
Bank #
DCPP L111 NRC Exam 21 November, 2012 P a g e l 69 (note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.5, 55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 70 Examination Outline Cross-Reference Level RO Ability to operate and / or monitor the following as they apply to the High Reactor Coolant Activity: Failed fuel-monitoring equipment Tier #
1 Group #
2 K/A #
APE 076 AA1.04 Rating 3.2 Question 61 GIVEN:
Unit 1 tripped from 100% power 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> ago Core Exit Thermocouples temperatures have been indicating higher than expected The Post Accident Sampling System is inoperable You are directed by the SFM to place normal letdown in service If the fuel has suffered cladding damage, high reactor coolant activity from the failed fuel will:
A. be cleaned up by the ion exchanger and remain unnoticed until the PASS is restored and reactor coolant is sampled.
B. cause area radiation monitors in the Auxiliary Building to indicate higher in the areas of CVCS and seal injection.
C. cause radiation monitor readings in Containment to rise when letdown flow is established.
D. only affect radiation levels in the Auxiliary Building if there is a letdown line failure.
Proposed Answer: B. cause area radiation monitors in the Auxiliary Building to indicate higher in the areas of CVCS and seal injection.
Explanation:
A. Incorrect. The Ion Exchanger will not remove all of the activity from the Reactor Coolant.
B. Correct. Radiation levels will increase at the Letdown Heat Exchanger and the RCP seal injection.
C. Incorrect. radiation levels in Containment are or will already be elevated and not affected by placing letdown in service.
D. Incorrect. While a Letdown Line Failure will cause radiation levels in the Auxiliary Building to increase, that is not the ONLY reason for increased radiation levels.
Technical
References:
LMCDCDA Core Damage Assessment; LB-1A, CVCS References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the CVCS. (40449)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR 41.7 / 45.5 / 45.6
DCPP L111 NRC Exam 21 November, 2012 P a g e l 71 Examination Outline Cross-Reference Level RO Knowledge of the specific basis for: Re-diagnosis and SI termination Tier #
1 Group #
2 K/A #
W/E01 2.4.18 Rating 3.3 Question 62 Unit 1 operators are performing E-1.1, SI Termination.
According to procedure E-1.1, the preferred RCP to run is:
A. RCP 1-2 because it provides normal pressurizer spray capabilities.
B. RCP 1-2 because it reduces the pressurizer level and pressure transients.
C. RCP 1-1 because it provides normal pressurizer spray capabilities.
D. RCP 1-1 because it reduces the pressurizer level and pressure transients.
Proposed Answer:
A. RCP 1-2 because it provides normal pressurizer spray capabilities Explanation:
A. Correct - Step 24 of E-1.1 states to verify RCP 2 running for spray capabilities.
B. Incorrect - reduced pressurizer level and pressure transients is due maintaining a saturated system in the pressurizer.
C. Incorrect - Step 24 of E-1.1 states that RCP 2 is checked operating to obtain normal pressurizer spray capabilities and if RCP 2 is not available then more than one running RCP may be necessary to provide normal pressurizer spray capabilities.
D. Incorrect - reduced pressurizer level and pressure transients is due maintaining a saturated system in the pressurizer.
Technical
References:
LPE-1B, E-1.1, SI Termination, Revision 12 E-1.1, SI Termination, Revision 26 WOGBD ES-1.1, SI Termination, Revision 2 References to be provided to applicants during exam: None Learning Objective: 7920S - Explain basis of emergency procedure steps (E-1.1, E-1.2)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 72 Examination Outline Cross-Reference Level RO Knowledge of the interrelations between the (Containment Flooding) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Tier #
1 Group #
2 K/A #
WE15 EK2.2 Rating 2.7 Question 63 Unit 1 has experienced a large break Loss of Coolant Accident.
The operator reports that containment recirc sump level is 99 feet. The STA reports that the Critical Safety Function Status tree for Containment Integrity is MAGENTA due to the sump level.
The crew will enter FR-Z.2, Response to Containment Flooding:
A. to determine what additional water source is filling containment above design level.
B. but immediately exit it once it is verified that this is approximately the expected level when the RWST reaches the transfer to cold leg recirc setpoint.
C. to check for indications of a faulted steam generator.
D. but immediately exit it once it is verified that this is approximately the expected level when the RWST reaches the level that trips the containment spray pumps.
Proposed Answer:
A. to determine what additional source of water is filling containment above design level.
Explanation:
A. Correct. FRZ-2 step 1 directs operators to exit Flooding in Containment for recirc sump levels less than 95.75 feet. A level of 99 feet is much higher than what is expected for a LCOA. The Response to Containment Flooding Background Document, FRZ-2, states, The maximum level of water in the containment following a major accident generally is based upon the entire water contents of the reactor coolant system, refueling water storage tank, condensate storage tank, and SI accumulators.
B. Incorrect. The status tree may just be magenta during a LOCA but will not reach 99 feet. This is the water volume from the RCS, RWST, CST, and 4 SI Accumulators. This approximates the maximum water volume introduced following a LOCA plus a steamline or feedline break inside of containment.
C. Incorrect. The correct level per step 1 of FRZ-2 is 95.75 feet. The maximum level of water in the containment following a major accident generally is based upon the entire water contents of the reactor coolant system, refueling water storage tank, condensate storage tank, and SI accumulators. This is the volume of LOCA plus a steamline or feedline break.
D. Incorrect. The status tree may just be magenta during a LOCA but will not reach 99 feet. The correct level per step 1of FRZ-2 is 95.75 feet.
Technical
References:
F-0, Critical Safety Function Status Trees, FR-Z.2, Response to Containment Flooding, Operations Lesson Mitigating Core Damage - Containment, Systems Lesson Containment Structure. Response to Containment Flooding Background Document, FRZ-2, References to be provided to applicants during exam:
None Learning Objective: Discuss abnormal conditions associated with the Containment Structure. (37590)
DCPP L111 NRC Exam 21 November, 2012 P a g e l 73 Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.7 / 45.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 74 Examination Outline Cross-Reference Level RO Ability to operate and / or monitor the following as they apply to the Natural Circulation Cooldown: Operating behavior characteristics of the facility Tier #
1 Group #
2 K/A #
EA1.2 Rating 3.6 Question 64 Unit 1 is performing a Natural Circulation Cooldown in accordance with E-0.2, Natural Circulation Cooldown.
Given these conditions, the maximum cooldown rate that the crew can establish is _____ °F/hr.
A. 25 B. 50 C. 90 D. 100 Proposed Answer: A. 25 Explanation: (Unit difference)
A. Correct -25°F/hr cooldown is the maximum rate for Unit 1 in E-0.2.
B. Incorrect - Unit 2 E-0.2, Step 10 states a cooldown rate of 50°F/hr C. Incorrect - A 90°F/hr cooldown rate is the administrative limit for a normal cooldown.
D. Incorrect - A 100°F/hr cooldown rate is the tech spec limit for a normal cooldown.
Technical
References:
E0.2, U1, Natural Circulation Cooldown, Revision 23 E-0.2, U2, Natural Circulation Cooldown, Revision 18 Operational Phase Training LPE-0.2, Natural Circulation Cooldown, Revision 14 References to be provided to applicants during exam: None Learning Objective: 5812 - State the operating limits during natural circulation cooldown Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.7
DCPP L111 NRC Exam 21 November, 2012 P a g e l 75 Examination Outline Cross-Reference Level RO Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
Tier #
1 Group #
2 K/A #
WE08 EA2.1 Rating 3.4 Question 65 The Shift Foreman enters FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, due to a RED path.
The first major action of FR-P.1 is to:
A. depressurize the RCS B. terminate Safety Injection if the criteria is satisfied C. stop the RCS cooldown D. perform RCS soak Proposed Answer: C. stop the RCS cooldown Explanation:
The major actions in FR P.1 are:
Stop the RCS cooldown, Terminate SI if the criteria are satisfied, Depressurize the RCS to minimize pressure stress, Establish normal operating conditions and stable RCS conditions, Soak if necessary prior to further restricted cooldown.
A. Incorrect. This is the third major action.
B. Incorrect. This is the second major action.
C. Correct. This is the first major action.
D. Incorrect. This is the final major action.
Technical
References:
FR-P.1, Response to Imminent Pressurized Thermal Shock Condition References to be provided to applicants during exam: None Learning Objective: Explain the general purpose/function (including MACs) of EOPs (FR-Ps). (3551Q)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.10 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 76 Examination Outline Cross-Reference Level RO Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Tier #
3 Group #
K/A #
2.1.7 Rating 4.4 Question 66 Unit 1 is at 30% power.
An inadvertent dilution will result in automatic rod _________________ due to ________________ RCS temperature.
A. withdrawal; lowering B. insertion; rising C. withdrawal; rising D. insertion, lowering Proposed Answer: B. insertion; rising Explanation:
A.
Incorrect - dilution will result in an RCS temperature increase vice a lowering temperature. The automatic rod withdrawal is consistent with the lowering RCS temperature.
B.
Correct - dilution will result in an RCS temperature increase and automatic rod insertion.
C.
Incorrect - dilution will result in an RCS temperature increase. The automatic rod withdrawal is consistent with a lowering not rising RCS temperature.
D.
Incorrect - dilution will result in an RCS temperature increase vice a lowering temperature. The automatic rod insertion is consistent with a rising not lowering RCS temperature.
Technical
References:
LTAA5, Reactivity Addition Accidents, Revision 10 DCPP U1 and U2 UFSAR, Revision 19 References to be provided to applicants during exam: None Learning Objective: 40581 - Discuss abnormal conditions associated with the reactor makeup control system Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.41.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 77 Examination Outline Cross-Reference Level RO Ability to direct personnel activities inside the control room.
Tier #
3 Group #
K/A #
2.1.9 Rating 2.9 Question 67 The Unit 1 Control Operator sees an unexpected increase in steam flow and reactor power is rapidly approaching the trip setpoint. The Shift Foreman is attending the Shift Brief.
Whose concurrence, if any, is required to trip the plant?
A. The Shift Manager ONLY B. Another Licensed Operator ONLY C. Shift Foreman ONLY D. No concurrence is required Proposed Answer: D. No concurrence is required Explanation:
A. Incorrect. May direct the plant to be tripped, but permission need not come from this position.
B. Incorrect. The CO may also direct the plant to be tripped, but permission does not necessarily have to come from him.
C. Incorrect. The SF may direct the plant to be tripped, but permission need not come from this position.
D. Correct. In accordance with OP1.DC10, Conduct of Operations, Licensed Operators are expected to manually initiate Engineered Safety Feature (ESF) actions, (Reactor Trips and Safety Injections), under the following circumstances: when plant parameters are approaching an automatic set point, such that the automatic action is judged to be unavoidable.
Technical
References:
OP1.DC10, Conduct of Operations; Operations Lesson Conduct of Operations, LADM-1.
References to be provided to applicants during exam: None Learning Objective: Describe the general duties and responsibilities of Control Room Operators.
(41662)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.10 / 45.5 / 45.12 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 78 Examination Outline Cross-Reference Level RO Knowledge of the process for controlling equipment configuration or status.
Tier #
3 Group #
K/A #
2.1.14 Rating 3.9 Question 68 Excess Letdown has been placed in service per the operating procedure OP B-1A:IV, Excess Letdown -
Place In Service and Remove From Service.
The control room operator placed Pink Tags on the Excess Letdown Isolation valves.
The use of the Pink Tag is:
A. Acceptable, providing a LBIE screen is performed prior to the end of the shift.
B. Unacceptable, unusual system lineups are tracked by using of the abnormal status board only.
C. Unacceptable, it can only be used in conjunction with an approved clearance.
D. Acceptable, it can be used to draw attention to components that are not normally in service, for short durations.
Proposed Answer: D. Acceptable, it can be used to draw attention to components that are not normally in service, for short durations.
Explanation:
Answer A is incorrect. LBIE screen is not required for Pink Tags; they do not control status of equipment.
Answer B is incorrect. While the system lineup is required to be placed on the abnormal status board, nothing precludes placing a pink tank on the component as well.
Answer C is incorrect. Pink Tags are not used for clearances.
Answer D is correct. The Pink Tag process may be used to designate an unusual component status and to draw attention to off normal components.
Technical
References:
OP1.DC10, Conduct of Operations, Revision 30 References to be provided to applicants during exam: None Learning Objective: 3313 - State the different methods of documenting abnormal system/component lineups/position.
Question Source:
Bank #
X (note changes; attach parent)
Modified Bank #
New Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 79 Examination Outline Cross-Reference Level RO Knowledge of conditions and limitations in the facility license.
Tier #
3 Group #
K/A #
2.2.38 Rating 3.6 Question 69 Per the Unit 1 Operating License, Pacific Gas and Electric Company, the maximum allowed reactor thermal power is:
A. 3421 MWT B. 3411 MWT C. 3401 MWT D. 3312 MWT Proposed Answer: B. 3411 MWT Explanation:
A. Incorrect. 3411 MwT is the maximum allowed power level.
B. Correct. 3411 MwT is the maximum allowed power level.
C. Incorrect. 3411 MwT is the maximum allowed power level.
D. Incorrect. 3411 MwT is the maximum allowed power level.
Technical
References:
Diablo Canyon Unit 1 Operating License References to be provided to applicants during exam: None Learning Objective: 9666 - Identify the operating license contents Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.7 / 41.10 / 43.1 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 80 Examination Outline Cross-Reference Level RO Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
Tier #
3 Group #
K/A #
2.2.44 Rating 4.2 Question 70 Based on the current control room indications and status of the plant systems, the Unit 2 crew has determined they need to transition to a different Emergency Operating procedure.
During the transition brief;
- 1. Specific foldout page assignments should be made to appropriate Control Room operators by assigning the foldout item number and the operator repeating back the high level action.
- 2. Specific foldout page parameters and values are required to be repeated back.
- 3. A copy of the foldout page should be given to any operator with an assignment.
A. 1 and 2 ONLY B. 2 and 3 ONLY C. 1 and 3 ONLY D. 1, 2, and 3 Proposed Answer: C. 1 and 3 ONLY Explanation:
A. Incorrect - Statement 2 is incorrect; specific values and parameters are not required to be repeated back.
B. Incorrect - Statement 2 is incorrect; specific values and parameters are not required to be repeated back.
C. Correct - According to the conduct of Operations, statements 1 and 3 are correct, but statement 2 is incorrect.
D. Incorrect - Statement 2 is incorrect; specific values and parameters are not required to be repeated back.
Technical
References:
OP1.DC10, Conduct of Operations References to be provided to applicants during exam: None Learning Objective: 41678 - Describe the expectations and standards for abnormal procedure use and adherence Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.5 Examination Outline Cross-Reference Level RO Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel Tier #
3 Group #
DCPP L111 NRC Exam 21 November, 2012 P a g e l 81 monitoring equipment, etc.
K/A #
2.3.5 Rating 2.9 Question 71 An operator is performing a frisk of their hands using a hand held frisker. Background counts are 75 cpm.
Which of the following describes how fast the operator should frisk their hands and what is the lowest count rate that would require the operator to contact RP?
A. 1 to 2 inches per second; 175 cpm B. 4 to 5 inches per second; 175 cpm C. 1 to 2 inches per second; 370 cpm D. 4 to 5 inches per second; 370 cpm Proposed Answer: A 1 to 2 inches per second; 175 cpm Explanation:
A correct. Proper speed is 1 to 2 inches per second. If the count rate (i.e., meter reading) increases to 100 cpm over background (i.e., 100 cpm more than the amount you started with), or the alarm goes off, then contact RP.
B incorrect. Proper speed is 1 to 2 inches per second.
C incorrect. Count rates 100 cpm above background. If background is 300 cpm or greater then the frisk should be done in another location.
D incorrect. Speed and counts are incorrect.
Technical
References:
GET lesson GRPA400 page 75 References to be provided to applicants during exam: None Learning Objective: F6.State the methods used to monitor personnel for contamination, the actions necessary upon discovering contamination, and list decontamination techniques for personnel.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.11 / 41.12 / 43.4 / 45.9
DCPP L111 NRC Exam 21 November, 2012 P a g e l 82 Examination Outline Cross-Reference Level RO Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Tier #
3 Group #
K/A #
2.3.13 Rating 3.4 Question 72 The DCPP Emergency Exposure Guidelines state the dose limit for saving plant equipment is ____
TEDE.
A. 5 REM B. 10 REM C. 25 REM D. 75 REM Proposed Answer: B. 10 REM Technical
References:
Operational Phase Lesson Plan LEP-3, EP RB Procedures, Revision 8 EP RB-2, DCPP Emergency Exposure Guidelines, Revision 8 References to be provided to applicants during exam: None Learning Objective: 7954 - State the emergency dose limits Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.12
DCPP L111 NRC Exam 21 November, 2012 P a g e l 83 Examination Outline Cross-Reference Level RO Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Tier #
3 Group #
3 K/A #
2.3.15 Rating 2.9 Question 73 A Personal Electronic Dosimeter, (PED), for a watchstander in the Auxiliary Building will measure what kinds(s) of radiation?
A. Gamma ONLY B. Beta ONLY C. Gamma and Beta ONLY D. Neutron, Beta and Gamma Proposed Answer: C. Gamma and Beta ONLY Explanation:
A incorrect. PED measures beta and gamma.
B incorrect. PED measures beta and gamma C Correct. PED mesures both.
D Incorrect. TLD can measure all 3.
Technical
References:
Rad Worker - Dosimetry page 57 References to be provided to applicants during exam:
None Learning Objective: E3.List the types of radiation detected by DCPP dosimetry Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.12 / 43.4 / 45.9 Reworded question. A PED will detect neutron (and a PED that corrects for neutron exposure is worn when entering Containment). A PED for a watchstander, not entering Containment, measures Gamma and Beta.
DCPP L111 NRC Exam 21 November, 2012 P a g e l 84 Examination Outline Cross-Reference Level RO Knowledge of operational implications of warnings, cautions, and notes.
Tier #
3 Group #
K/A #
2.4.20 Rating 3.8 Question 74 The function of a _______________ is to provide ____________________________.
A. Caution; information which supports operator action B. Note; conditional statements that provide for prompt operator response C. Note; information about potential hazards to personnel or equipment D. Caution; information about potential hazards to personnel or equipment Proposed Answer: D. Caution; information about potential hazards to personnel or equipment Explanation:
A. Incorrect - Cautions are information about potential hazards to personnel or equipment. The description contained in this answer describes a Note.
B. Incorrect - This statement describes a foldout page, not a Note.
C. Incorrect - Notes are information which supports operator action. The description contained in this answer describes a Caution.
D. Correct - Cautions are information about potential hazards to personnel or equipment. The description contained in this answer describes a Note.
Technical
References:
LPE-Rule, Rules of Usage, Revision 11 References to be provided to applicants during exam: None Learning Objective: 5428 - Explain definition of terms used in the EOPs Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.41.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 85 Examination Outline Cross-Reference Level RO Knowledge of entry conditions and immediate action steps.
Tier #
3 Group #
K/A #
2.4.1 Rating 4.6 Question 75 Which of the following procedures contain immediate action steps that can be performed by the operator at the controls from memory?
A. - E-0, Reactor Trip/Safety Injection
- ECA-0.0, Loss of All Vital AC
- E-0.1, Reactor Trip Response B. - E-0, Reactor Trip/Safety Injection
- FR-S.1, Response to Nuclear Power Generation/ATWS
- AP-17, Loss of Charging C. - E-0, Reactor Trip/Safety Injection
- E-1, Loss of Reactor or Secondary Coolant
- AP-12A, Continuous Insertion or Withdrawal of a Control Rod Bank D. - ECA-0.0, Loss of All Vital AC
- AP-15, Loss of Feedwater
- FR-S.1, Response to Nuclear Power Generation/ATWS Proposed Answer: D. - ECA-0.0, Loss of All Vital AC
- AP-15, Loss of Feedwater
- FR-S.1, Response to Nuclear Power Generation/ATWS Explanation:
A. Incorrect. -0.1 does not have immediate action steps.
B. Incorrect. AP-17 does not have immediate action steps.
C. Incorrect. E-1 and AP-12A do not have immediate action steps.
D. Correct. All of these have immediate action steps.
Technical
References:
Operations Lesson LPE Rule - Rules of Usage; E-0, Reactor Trip/Safety Injection; ECA-0.0, Loss of All Vital AC; -0.1, Reactor Trip Response; FR-S.1, Response to Nuclear Power Generation/ATWS; AP-17, Loss of Charging; E-1, Loss of Reactor or Secondary Coolant; AP-12A, Continuous Insertion or Withdrawal of a Control Rod Bank; AP-15, Loss of Feedwater.
References to be provided to applicants during exam: None Learning Objective: Describe the expectations and standards for abnormal procedure use and adherence, including: Performance of immediate actions (41678)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam NA Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.10 / 43.5/45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 86 Examination Outline Cross-Reference Level SRO Knowledge of abnormal condition procedures related to small break LOCA Tier #
1 Group #
1 K/A #
000009 2.4.11 Rating 4.2 Question 76 GIVEN:
- Unit 1 is in MODE 4
- Pressurizer level is 20% and lowering
- RCS subcooling is 15°F
- The PZR PORVs and Safeties are all closed
- Train A RHR is aligned for shutdown cooling Which of the following procedures should be entered?
A. E-1, Loss of Reactor or Secondary Coolant B. AP-24, Shutdown LOCA C. AP SD-0, Loss of, or Inadequate Decay Heat Removal D. AP SD-2, Loss of RCS Inventory Proposed Answer: B. AP-24, Shutdown LOCA Explanation:
E. Incorrect - E-1 used for actions following events while at power.
F. Correct - Applicable to Mode 4 for loss of RCS inventory.
G. Incorrect - Applicable to Mode 5 or 6.
H. Incorrect-Applicable to Mode 5 or 6.
Technical
References:
E-1, Loss of Reactor or Secondary Coolant, Revision 30 AP-24, Shutdown LOCA, Revision 10 AP SD-2, Loss of RCS Inventory, Revision 18 AP SD-0, Loss of, or Inadequate Decay Heat Removal, Revision 12 References to be provided to applicants during exam: None Learning Objective: 3478 - Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 87
DCPP L111 NRC Exam 21 November, 2012 P a g e l 88 Examination Outline Cross-Reference Level SRO Ability to determine or interpret the following as they apply to a Large Break LOCA: That equipment necessary for functioning of critical pump water seals is operable Tier #
1 Group #
1 K/A #
011 EA2.07 Rating 3.2 Question 77 According the bases of Technical Specification 3.7.7, Component Cooling Water, what is the MINIMUM ASW and CCW equipment necessary to support a Design Bases Accident?
A. One ASW pump, one CCW heat exchanger and one CCW pump B. One ASW pump, one CCW heat exchanger and two CCW pumps C. One ASW pump, two CCW heat exchangers and one CCW pump D. Two ASW pumps, two CCW heat exchangers and two CCW pumps Proposed Answer: B. One ASW pump, one CCW heat exchanger and two CCW pumps Explanation:
A is incorrect. Per Tech Spec 3.7.7, The CCW system is designed to provide sufficient heat removal for normal and post accident ESF heat loads without overheating. The CCW system and ASW system are essentially considered a single heat removal system for the purpose of assessing the ability to sustain either a single active or passive failure and still perform design basis heat removal. Only one ASW pump and one CCW heat exchanger is required, as assumed in the safety analysis, to provide sufficient heat removal from containment to mitigate a DBA.
However, to ensure maximum heat removal capability, operators are instructed to place the second CCW heat exchanger in service early in the emergency operating procedures. (However, for CCW, 2 pumps are required for a vital loop.)
B correct. The Tech Spec states: In the event of a DBA, one vital CCW loop is required to provide the minimum heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water. A vital CCW loop is considered OPERABLE when:
- b. The associated piping, valves, and instrumentation and controls required to perform the safety related function are OPERABLE.
C incorrect. While two heat exchangers are placed in service, it is more than what is required.
D incorrect. All equipment is started during an accident, but it is more than required.
Technical
References:
TS Bases 3.7.7 References to be provided to applicants during exam: None Learning Objective: 9694G - Apply TS 3.7 Technical Specification bases Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
DCPP L111 NRC Exam 21 November, 2012 P a g e l 89 Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR 43.5 / 45.13
DCPP L111 NRC Exam 21 November, 2012 P a g e l 90 Examination Outline Cross-Reference Level SRO Ability to determine and interpret the operable control channel as it applies to the Pressurizer Pressure Control Malfunctions.
Tier #
1 Group #
1 K/A #
027 AA2.18 Rating 3.4 Question 78 Unit 2 is at 100% power.
A pressurizer pressure channel failure has resulted in:
- PT-456 indication failing low
- PK04-04, OTT C-3, activating
- PK04-06, Protection Channel, activating
- PK02-04, Safeguard Channel, activating Based on these conditions, the Shift Foreman should direct the operator to select:
A. PT-455/474 on the Pressure Channel Selector Switch in accordance with AP-5, Malfunction of Eagle 21 Protection or Control Channel.
B. PT-474/457 on the Pressure Channel Selector Switch in accordance with AP-5, Malfunction of Eagle 21 Protection or Control Channel.
C. PT-455/474 on the Pressure Channel Selector Switch in accordance with AP-13, Malfunction of Reactor Pressure Control System.
D. PT-474/457 on the Pressure Channel Selector Switch in accordance with AP-13, Malfunction of Reactor Pressure Control System.
Proposed Answer: A. PT-455/474 on the Pressure Channel Selector Switch in accordance with AP-5, Malfunction of Eagle 21 Protection or Control Channel.
Explanation:
A. Correct - AP-5 is entered from all three ARPKs due to failure of the control system due to the channel failure. With PT-456 failed, PT 474 is the only used as a backup channel for PT-455.
B. Incorrect - AP-5 is entered from all three ARPKs due to failure of the control system due to the channel failure. PT 457 as the controlling channel can only have PT-456 as the backup channel and PT-456 is the failed channel.
C. Incorrect - AP-13 would not be entered due to this failure would not result in a plant transient.
D. Incorrect - AP-13 would not be entered due to this failure would not result in a plant transient.
Technical
References:
OIM 4-4a, Pressurizer Pressure Channel Functions OIM 4-4b, Pressurizer Pressure Channel Failures AP-5, malfunction of Eagle 21 Protection or Control Channel, Revision 28B
DCPP L111 NRC Exam 21 November, 2012 P a g e l 91 AP-13 malfunction of Reactor Pressure Control System, Revision 5A LA-4A, Pressurizer, Pressure and Level Control References to be provided to applicants during exam: None Learning Objective: 3478 - Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 92 Examination Outline Cross-Reference Level SRO Ability to determine operability and/or availability of safety related equipment. Steam Gen. Tube Rupture / 3 Tier #
1 Group #
1 K/A #
038 2.2.37 Rating 4.6 Question 79 GIVEN:
- A steam generator tube rupture has occurred in Steam Generator 2-2
- RCS pressure is 1950 psig and lowering slowly
- A loss of offsite power has occurred
- Intact steam generator pressures are 1000 psig and stable The Shift Foreman should direct the operator to:
A. depressurize to less than 1915 psig using normal spray; after blocking low steamline pressure SI, initiate a cooldown using the condenser steam dumps on the intact steam generators.
B. depressurize to less than 1915 psig using normal spray; after blocking low steamline pressure SI, initiate a cooldown using the 10% steam dumps on the intact steam generators.
C. depressurize to less than 1915 psig using a Pressurizer PORV and initiate a cooldown using the condenser steam dumps in parallel with the depressurization and blocking of low steamline pressure SI.
D. initiate a RCS cooldown using the 10% steam dumps on the intact steam generators.
Proposed Answer: D. initiate a RCS cooldown using the 10% steam dumps on the intact steam generators.
Explanation:
Answer: D. initiate a RCS cooldown using the 10% steam dumps on the intact steam generators.
Justification:
A incorrect. normal spray and condenser steam dumps not available due to loss of offsite power.
B incorrect. the depressurization is not required because the reason for the depressurization is to block the low pressure SI and prevent an Main Steam Isolation. For the given plant conditions the isolation valves are already closed. Additionally, the steps are done in parallel.
C incorrect. Condenser steam dumps not available.
D correct. The isolation valves are already closed, there is no need to depressurize the RCS to less than P-11.
Technical
References:
E-3, Steam Generator Tube Rupture steps 8 thru 10 References to be provided to applicants during exam: None Learning Objective: (Lic Trng) SIM0063 Implement E-3 Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
DCPP L111 NRC Exam 21 November, 2012 P a g e l 93 New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 94 Examination Outline Cross-Reference Level SRO Ability to determine and interpret DC loads that are lost and the impact on the ability to operate and monitor plant systems.
Tier #
1 Group #
1 K/A #
058 AA2.03 Rating 3.9 Question 80 Unit 2 is at 100% power.
The operators report:
- The reactor has tripped
- Feed flow indicates zero on all steam generators
- Main Feed pump #1 is tripped and Main Feed pump #2 is running
- 4KV Bus F is de-energized
- PY-21 is de-energized
- None of the 40% steam dumps are open
- Loss of indication for 4 kV bus loads
- None of the pressurizer heaters are energized What procedure should the Shift Foreman use to address the current plant conditions:
A. OP AP-4, Loss of Vital or Nonvital Instrument AC B. OP AP-15, Loss of Feedwater Flow C. OP AP-23, Loss of Vital DC D. OP AP-27, Loss of 4KV and/or 480V Vital Bus Proposed Answer: C. OP AP-23, Loss of Vital DC Explanation:
A. Incorrect - Breaker control is DC not AC, but, this will result in multiple alarms. PY-21 is lost due to the loss of bus F.
B. Incorrect - Loss of feedwater is due to closure of the feed regulating valves due to loss of DC power to the train A solenoids.
C. Correct - Loss of vital DC power will result in the failure of the feed regulating valve, the failure of the 4KV bus to transfer due to loss of DC power to the breakers.
D. Incorrect - loss of the 4KV bus resulted from the loss of vital DC. AP-27 is used to recover from a loss of AC power.
Technical
References:
OP AP-23 References to be provided to applicants during exam: None Learning Objective: 3478 - Give initial conditions, assumptions, and symptoms, determine the
DCPP L111 NRC Exam 21 November, 2012 P a g e l 95 correct abnormal operating procedure to be used to mitigate an operational event.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 96 Examination Outline Cross-Reference Level SRO Knowledge of how abnormal operating procedures are used in conjunction with s. Reactor Trip Tier #
1 Group #
1 K/A #
EPE 007 2.4.8 Rating 4.5 Question 81 GIVEN:
- Unit 1 has tripped
- The actions of E-0, Reactor Trip or Safety Injection steps 1 through 4 have just been completed
- Safety Injection is not required
- RCS Temperature is 535ºF and slowly lowering
- All Steam Dumps, Steam Generator (SG) Blowdown Isolation valves are closed; MSRs are reset; and all the MSIVs and MSIV Bypass valves are closed
- All SG Narrow Range Water Levels are 32% and slowly rising
- All SG pressures are 1000 psig and steady Based on these conditions, the Shift Foreman should enter:
A. E-1.1, SI Termination, then direct operators to Go To E-1.2, Post LOCA Cooldown and Depressurization. In accordance with OP1.DC10, Conduct of Operations, the words Go To mean to exit E-1.1 and enter E-1.2.
B. E-1.1, SI Termination, then direct operators to Go To E-1.2, Post LOCA Cooldown and Depressurization. In accordance with OP1.DC10, Conduct of Operations, the words Go To mean to perform E-1.1 in parallel with E-1.2.
C. E-0.1, Reactor Trip Response, then direct operators to implement AP-6, Emergency Boration, and continue to step 2 of E-0.1. In accordance with OP1.DC10, Conduct of Operations, the word implement means to perform AP-6 after completing the immediate actions of E-0.1.
D. E-0.1, Reactor Trip Response, then direct operators to implement AP-6, Emergency Boration, and continue to step 2 of E-0.1. In accordance with OP1.DC10, Conduct of Operations, the word implement means to perform AP-6 in parallel with E-0.1.
Proposed Answer: D E-0.1, Reactor Trip Response, then direct operators to implement AP-6, Emergency Boration, and continue to step 2 of E-0.1. In accordance with OP1.DC10, Conduct of Operations, the word implement means to perform AP-6 in parallel with E-0.1.
Explanation:
A. Incorrect. For the given plant conditions enter E 0.1 (not E-1.1) and direct operators to implement AP-6 Emergency Boration and continue to step 2. In accordance with OP1.DC10, Conduct of Operations, the words Go To do mean to exit the procedure in effect and enter the directed procedure. Step 4 of E-0.0, the Response Not Obtained column directs operators to enter E-0.1 for the given plant conditions.
DCPP L111 NRC Exam 21 November, 2012 P a g e l 97 B. Incorrect. E0.1 should be entered, not E-1.1. Also, AP-6 (not E-1.2) should be entered in parallel with the procedure in effect. This is the definition of implement not Go To.
C. Incorrect. E-0.1 is correct. This is also the correct action in accordance with AP-6 but E0.1 should still be executed in parallel. Answer C has an incorrect definition of implement.
D. Correct. For the given plant conditions enter E 0.1 and direct operators to implement AP-6 Emergency Boration and continue to step 2. In accordance with OP1.DC10, Conduct of Operations, the word implement means to perform the second procedure in parallel with the procedure in effect.
Technical
References:
E-0, Reactor Trip or Safety Injection; OP1.DC10, Conduct of Operations; E-0.1, Reactor Trip Response.
References to be provided to applicants during exam: None Learning Objective: Describe what procedure or procedure set would be used in an emergency event, based on plant mode/conditions. (6764)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 98 Examination Outline Cross-Reference Level SRO Knowledge of the Emergency Boration system set points, interlocks and automatic actions associated with entry conditions Tier #
1 Group #
2 K/A #
000024 2.4.2 Rating 4.6 Question 82 Unit 1 is experiencing an uncontrolled cooldown following a reactor trip with no ESF actuation.
In _______, the Shift Foreman should ensure the boric acid flow rate is at least ____ gpm per AP-6, Emergency Boration.
A. E-0, Reactor Trip or Safety Injection; 30 B. E-0, Reactor Trip or Safety Injection; 50 C. E-0.1, Reactor Trip Response; 30 D. E-0.1, Reactor Trip Response; 50 Proposed Answer: C E-0.1, Reactor Trip Response; 30 Explanation:
A. Incorrect - E-0 does not direct entry into AP-6 to emergency borate the plant. AP-6 Step 1 verifies at least 30 GPM of Boric Acid Flow.
B. Incorrect - E-0 does not direct entry into AP-6 to emergency borate the plant. AP-6 Step 1 verifies at least 30 GPM of Boric Acid Flow instead of the 50 GPM specified in the answer. The 50 GPM emergency boration flowrate is the flow in which the Emergency Boration Flowmeter FI-113 pegs high specified in the Note on page 3 of AP-6.
C. Correct - E-0.1 requires entry into AP-6 due to uncontrolled cooldown per the Step 1 ROIs. AP-6 Step 1 verifies at least 30 GPM of Boric Acid Flow.
D. Incorrect - E-0.1 requires entry into AP-6 due to uncontrolled cooldown per the Step 1 ROIs. AP-6 Step 1 verifies at least 30 GPM of Boric Acid Flow instead of the 50 GPM specified in the answer. The 50 GPM emergency boration flowrate is the flow in which the Emergency Boration Flowmeter FI-113 pegs high specified in the Note on page 3 of AP-6.
Technical
References:
E-0, Reactor Trip and Safety Injection, Revision 40 E-0.1, Reactor Trip Response, Revision 35 AP-6, Emergency Boration, Revision 19 References to be provided to applicants during exam: None Learning Objective: 3477F - Given an abnormal condition, summarize the major actions of OP AP-6 to mitigate an event in progress.
Question Source:
Bank #
DCPP L111 NRC Exam 21 November, 2012 P a g e l 99 (note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.45.7-8
DCPP L111 NRC Exam 21 November, 2012 P a g e l 100 Examination Outline Cross-Reference Level SRO Ability to determine and interpret the following as they apply to the Fuel Handling Incidents: AA2.01 ARM system indications Tier #
1 Group #
2 K/A #
APE 036 AA2.01 Rating 3.9 Question 83 Unit 1 is in a refueling outage with the conditions:
- Alarm PK 11-10 FHB HIGH RADIATION RE-58/59 comes in
- Alarm PK 02-06 CONTMT VENT ISOLATION comes in
- The Spent Fuel Pool level is 25 feet above spent fuel assemblies and steady The Shift Foreman should enter:
A. AP-21 Irradiated Fuel Damage; the Fuel Handling Building Evacuation Horn is sounding, and the Fuel Handling Building Ventilation is operating in the iodine removal mode.
B. AP-22 Spent Fuel Pool Abnormalities; the Fuel Handling Building Evacuation Horn is sounding, and Containment Ventilation has isolated.
C. AD8.DC54 Containment Closure; the Containment Evacuation Alarm is sounding, and the Fuel Handling Building Ventilation is operating in the iodine removal mode.
D. AP-21 Irradiated Fuel Damage; the Containment Evacuation Alarm is sounding, and the Fuel Handling Building Ventilation is operating in the iodine removal mode.
Proposed Answer: A AP-21 Irradiated Fuel Damage; the Fuel Handling Building Evacuation Horn is sounding, and the Fuel Handling Building Ventilation is operating in the iodine removal mode Explanation:
A. Correct. AP-21 is correct. RE-58/59 automatically sound the FHB Evacuation Horn, and shift the FHB Ventilation to Iodine Removal mode.
B. Incorrect. AP-22 is not correct since SFP level meets the minimum level and is not changing. Also, the Containment Evacuation Alarm is not sounded until Containment Radiation alarms come in and are verified by the Refueling SRO or SFM. Containment only isolates if the accident is inside containment; plant conditions are for an accident in the FHB.
C. Incorrect. Plant conditions are not met for entering AD8.DC54, which is for an accident inside of Containment. Also, the Containment Evacuation Alarm is not sounded until Containment Radiation alarms come in and are verified by the Refueling SRO or SFM.
D. Incorrect. The Containment Evacuation Alarm is not sounded until Containment Radiation alarms come in and are verified by the Refueling SRO or SFM. The FHB Evacuation Horn should be sounding due to RE-58.59 alarming.
DCPP L111 NRC Exam 21 November, 2012 P a g e l 101 Technical
References:
Operations Lesson Radiation Monitoring; AP-21 Irradiated Fuel Damage; AP-22 Spent Fuel Pool Anomaly; AD8.DC54 Containment Closure.
References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the Radiation Monitoring System. (37875)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 102 Examination Outline Cross-Reference Level SRO Knowledge of the emergency action level thresholds and classifications associated with Accidental Gaseous Radwaste Release.
Tier #
1 Group #
2 K/A #
000060 2.4.41 Rating 4.6 Question 84 An Unusual Event would have to be declared per the Emergency Plan if an unplanned release of gaseous radioactivity occurs that is greater than 2 times the limit contained in _________ for greater than 60 minutes.
A. OP G-2:V, Gaseous Radwaste-Gas Decay Tank Discharge B. The Radiological Effluent Technical Specifications (RETS)
C. The Final Safety Analysis Report D. 10 CFR 20; Standards for Protection Against Radiation Proposed Answer: B. The Radiological Effluent Technical Specifications Explanation:
A. Incorrect - The procedure do not contain the gaseous radioactivity release limits.
B. Correct - EAL R1 states that the limit is based on the values found in the Radiological Effluent Technical Specifications.
C. Incorrect - The FSAR does not contain gaseous radioactivity release limits.
D. Incorrect 10 CFR 20 does not contain gaseous radioactivity release limits Technical
References:
EP G-1 Emergency Action Level Matrix, Revision 40 References to be provided to applicants during exam: None Learning Objective: 42285 - Given indications of an event, use EP G-1 to classify the event with 100% accuracy within 15 minutes Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.43.5 Added (RETS)
DCPP L111 NRC Exam 21 November, 2012 P a g e l 103 Examination Outline Cross-Reference Level SRO Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: Whether an alarm channel is functioning properly Tier #
1 Group #
2 K/A #
061 AA2.04 Rating 3.5 Question 85 GIVEN:
- Unit 1 is in MODE 4
- The operator performing a channel check of RM-11, Containment Air Particulate, and RM-12, Containment Rad Gas, reports the following:
o RM-11 is hard on its low level peg and motionless o
RM-12 is reading is 1.5 times what is was 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago on the last channel check After investigation, no change reported in calculated RCS leakrate or Containment atmosphere and parameters.
Based on this, Technical Specification LCO 3.4.15, RCS Leakage Detection Instrumentation:
A. Is still met and requires no action.
B. Requires grab samples for the containment atmosphere be analyzed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C. Requires grab samples for the containment atmosphere be analyzed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D. Requires Unit 1 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 />.
Proposed Answer: B Requires grab samples for the containment atmosphere be analyzed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Explanation:
OPS policy B-5 states: if the meter needle is hard against its low limit peg and motionless, the channel may have failed low and should be considered inoperable.
A. Incorrect. Because RM-11 is reading downscale and the needle is motionless, it is inoperable.
B. Correct. This action is required if the instrument are inoperable. ACTION B C. Incorrect. This action is required if the containment sump monitor, with no CFCU drain collection in service, is inoperable.
D. Incorrect. This action is required if a required action in this technical specification is not met.
Technical
References:
Technical Specification 3.4.15. OPS policy B-5 References to be provided to applicants: TS 3.4.15 and Page 18 of STP I-1A (Unit1) 2.1.
Learning Objective: Discuss significant Technical Specifications and Equipment Control Guidelines associated with the Radiation Monitoring System. (9694, 9697, 9633)
Question Source:
Bank #
DCPP L111 NRC Exam 21 November, 2012 P a g e l 104 (note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 105 Examination Outline Cross-Reference Level SRO Knowledge of entry conditions and immediate actions steps associated with the Chemical and Volume Control system.
Tier #
2 Group #
1 K/A #
000004 2.4.1 Rating 4.8 Question 86 The reactor trips due to a spurious Safety Injection actuation.
Procedurally, when will the crew terminate charging injection and establish normal charging?
A. E-0, Reactor Trip or Safety Injection to minimize the possibility of Pressurizer overfill.
B. E-0, Reactor Trip or Safety Injection to align the plant to the conditions necessary to then transition to E-0.1, Post-Trip Response.
C. E-1, Loss of Reactor or Secondary Coolant to sequentially secure SI pumps while checking for indications of a RCS break.
D. E-1.1, Safety Injection Termination to allow for a check of the Critical Safety Functions prior to stopping ECCS pumps.
Proposed Answer: A. E-0, Reactor Trip or Safety Injection to minimize the possibility of Pressurizer overfill Explanation:
A. Correct - Charging is realigned prior to transitioning out of E-0 to minimize pressurizer overfill. The stopping of the 1-3 CCP and realigning the charging is a TCOA.
B. Incorrect - While normal charging will be inservice, E-0.1 is only entered if SI did not actuate.
C. Incorrect - A transition to E-1 is not appropriate.
D. Incorrect - The status trees have not been checked until the transition, however this is a reason to wait until E-1.1 to realign charging.
Technical
References:
E-0, Reactor Trip or Safety Injection, Revision 40 E-1.1, SI Termination LPE0 References to be provided to applicants during exam: None Learning Objective: 5433 - Identify exit conditions for the EOPs Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.43.5 Modified setup (spurious to first sentence, rearranged question wording).
DCPP L111 NRC Exam 21 November, 2012 P a g e l 106 Examination Outline Cross-Reference Level SRO Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: erratic power supply operation Tier #
1 Group #
2 K/A #
012 A2.04 Rating 3.2 Question 87 Unit 1 is shutting down and:
- Reactor Power is 10E-8 amps
- Intermediate Range Nuclear Instrument Channel N35 loses compensating voltage due to a power supply malfunction resulting in erratic voltage.
Technical Specification 3.3.1, Reactor Trip System Instrumentation, requires _______. The bases for having the Intermediate Range Nuclear Instruments OPERABLE is to mitigate the consequences of ________.
A. Reactor power be less than P-6 or greater than P-10 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; a dilution accident.
B. Placing the channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; an RCCA rod bank withdrawal accident.
C. Placing the channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; a dilution accident.
D. Reactor power be less than P-6 or greater than P-10 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; an RCCA rod bank withdrawal accident.
Proposed Answer: D Reactor power be less than P-6 or greater than P-10 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; an RCCA rod bank withdrawal accident.
Explanation:
A. Incorrect. The bases for this TS is to mitigate the consequences of a rod withdrawal accident.
B. Incorrect. The correct TS action is to have reactor power less than P-6 or greater than P-10 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C. Incorrect. The bases for this TS is to mitigate the consequences of a rod withdrawal accident. Second, the action is incorect.
D. Correct. Per TS 3.3.1, reactor power must be less than P-6 or greater than P-10 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; The TS bases document states this is to mitigate the consequences of an RCCA rod bank withdrawal accident.
Technical
References:
TS 3.3 and Bases TS 3.3; Systems Lesson Reactor Protection System LB-6A; ARPK 03-06 References to be provided to applicants during exam: TS 3.3.1 with page 1 of table 3.3.1-1 ONLY
DCPP L111 NRC Exam 21 November, 2012 P a g e l 107 Learning Objective: Discuss significant Technical Specifications and Equipment Control Guidelines associated with the Reactor Protection System. (9694, 9697, 9633)
Apply TS 3.3 Technical Specification LCOs. (9697C)
Apply TS 3.3 Technical Specification bases (SROs only). (9694C)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 108 Examination Outline Cross-Reference Level SRO Ability to (a) predict the impacts of a stuck open PORV or code safety on the Pressurizer Relief Tank; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of a stuck open PORV or code safety.
Tier #
2 Group #
1 K/A #
007 A2.01 Rating 4.2 Question 88 GIVEN:
- Unit 1 is at 100% power
- PK05-20, PZR Relief/Safety Valve Open - ON
- PK05-23, PZR Safety or Relief Line Temp - ON
- Pressurizer pressure dropped rapidly
- SI has not actuated The Shift Foreman should enter ________ and at 10 psig, __________ will close if open.
A. AP-13, Malfunction of Reactor Pressure Control; PCV-472, PRT Vent Header Pressure Control valve B. AP-1, Excessive Reactor Coolant System Leakage; RCS-8034A/B, Gas Analyzer valves C. E-1, Loss of Reactor Coolant or Secondary Coolant; PCV-472, PRT Vent Header Pressure Control valve D. AP-13, Malfunction of Reactor Pressure Control; RCS-8034A/B, Gas Analyzer valves Proposed Answer: A AP-13, Malfunction of Reactor Pressure Control; PCV-472, PRT Vent Header Pressure Control valve Explanation:
A. Correct - PK05-23 directs going to AR PK05-20 if PK05-20 and PK05-23 are in alarm.
AR PK05-20 directs entry into AP-13 due to pressurizer pressure has dropped rapidly. At 10 psig in the PRT PCV-472 will close if the valve is in the open position to isolate the PRT from the vent header.
B. Incorrect - AP-1 has potential symptoms as given above but is not directed to be entered from the indication given. RCS-8034A/B will close on a Phase A containment isolation not on PRT pressure.
C. Incorrect - E-1 entry is post SI Actuation therefore E-1 would not be entered given this information. At 10 psig in the PRT PCV-472 will close if the valve is in the open position to isolate the PRT from the vent header.
D. Incorrect - PK05-23 directs going to AR PK05-20 if PK05-20 and PK05-23 are in alarm.
AR PK05-20 directs entry into AP-13 due to pressurizer pressure has dropped rapidly.
RCS-8034A/B will close on a Phase A containment isolation not on PRT pressure.
Technical
References:
AR PK05-20, PZR Relief/Safety Valves Open, Revision
DCPP L111 NRC Exam 21 November, 2012 P a g e l 109 AR PK05-23, PZR Safety or Relief Line Temp, Revision 21 AP-1, Excessive Reactor Coolant System Leakage, Revision 20 AP-13, Malfunction of Reactor Pressure Control System, Revision 5A E-1, Loss of Reactor or Secondary Coolant, Revision 30 References to be provided to applicants during exam: None Learning Objective: 3478 - Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis 10CFR Part 55 Content:
55.43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 110 Examination Outline Cross-Reference Level SRO Ability to perform specific system and integrated plant procedures during all modes of plant operation. Service Water Tier #
1 Group #
2 K/A #
076 2.1.23 Rating 4.4 Question 89 GIVEN:
- Unit 1 is at 40% power
- Annunciator BAR/RACKS SCREENS, PK13-01, is lit
- A nuclear operator calls the control room and reports heavy kelp loading on the Circulating Water System traveling screens
- Unit 1 Screen differential pressure is 80 inches and rising
- Only Circulating Water Pump 1-1 is in service With these conditions, the Shift Foreman should:
A. Enter AP-25, Rapid Load Reduction, and reduce power to less than 25%.
B. Enter AP-7, Degraded Condenser, and reduce power to less than 25%.
C. Enter AP-25, Rapid Load Reduction, and trip the reactor D. Enter AP-7, Degraded Condenser, and trip the reactor.
Proposed Answer: D Enter AP-7, Degraded Condenser, and trip the reactor Explanation:
A. Incorrect. 1. Enter AP-7, Degraded Condenser should be entered, not AP-25. 2. While entering AP-25 is a judgment call, plant conditions call for AP-7 which directs operators to trip the reactor vice reducing power because only 1 CWP is operating and reactor power is above 25%.
B. Incorrect. 2. AP-7 directs operators to trip the reactor vice reducing power.
C. Incorrect. 1. Enter AP-7, Degraded Condenser should be entered, not AP-25. 3. OP F-1:VI should be entered to provide alternate cooling water to the SCW HX D. Correct. These are the correct actions for plant conditions.
Technical
References:
AP-25, Rapid Load Reduction; AP-7, Degraded Condenser; OP F-1:VI Service Cooling Water - Alternate Cooling; Operations Lesson Service Water System References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the Service Cooling Water System. (37109)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam
DCPP L111 NRC Exam 21 November, 2012 P a g e l 111 Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 41.10 / 43.5 / 45.2 / 45.6
DCPP L111 NRC Exam 21 November, 2012 P a g e l 112 Examination Outline Cross-Reference Level SRO Ability to (a) predict the impacts of a Containment Evacuation (including recognition of the alarm) on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of a Containment Evacuation.
Tier #
2 Group #
1 K/A #
000103 A2.04 Rating 3.6 Question 90 Unit 1 is off-loading the core.
While in process of withdrawing a fuel assembly the Red overload light on the Manipulator Crane Control Console illuminated and the refueling crew observed a release of gas bubbles from the fuel assembly.
Given the following indications:
- PK02-06, Containment Vent Isolation - off
- PK11-19, Containment Radiation - off
- PK11-21, High Radiation - on The Containment evacuation alarm has been sounded.
The alarm is a(n) __________ and the Shift Foreman should ensure the ________.
A. electronic warbler that sounds a series of falling tones; Containment Purge Exhaust System is in service per OP H-4:I, Containment Ventilation Make Available and Place in Service B. constant, high-pitched squeal; Containment Purge Exhaust System is in service per OP H-4:I, Containment Ventilation Make Available and Place in Service C. electronic warbler that sounds a series of falling tones; Containment Purge Exhaust System is isolated per AP-21, Irradiated Fuel Damage D. constant, high-pitched squeal; Containment Purge Exhaust System is isolated per AP-21, Irradiated Fuel Damage Proposed Answer: C electronic warbler that sounds a series of falling tones; Containment Purge Exhaust System is isolated per AP-21, Irradiated Fuel Damage Explanation:
A. Incorrect - Correct alarm but AP-21 requires that containment purge exhaust is verified closed vice in-service during a containment evacuation. OP H-4:I does give direction for placing the containment purge system in service but this condition requires the system to be isolated.
B. Incorrect - Incorrect alarm and AP-21 requires that containment purge exhaust is verified closed vice in-service during a containment evacuation. OP H-4:I does give direction for placing the containment purge system in service but this condition requires the system to be isolated.
C. Correct - Correct alarm and AP-21 requires that containment purge exhaust is verified
DCPP L111 NRC Exam 21 November, 2012 P a g e l 113 closed.
D. Incorrect - Incorrect alarm and AP-21 requires that containment purge exhaust is verified closed.
Technical
References:
AP-21, Irradiated Fuel Damage, Revision 10 References to be provided to applicants during exam: None Learning Objective: 3477 - Given an abnormal condition, summarize the major actions of OP AP-21 to mitigate an event in progress.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
55.43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 114 Examination Outline Cross-Reference Level SRO Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of forced circulation of Reactor Coolant Tier #
2 Group #
2 K/A #
002 A2.03 Rating 4.1 Question 91 GIVEN:
- Unit 1 is at 100% power
- The Plant Process Computer shows that RCP 13 Top Radial Bearing Temp is 201ºF and steady The Shift Foreman should _________.
A. direct the operator to trip the reactor and enter E-0 Reactor Trip or Safety Injection; then trip RCP 13, and close the associated PZR Spray Valve.
B. direct the operator to trip the reactor and enter E-0 Reactor Trip or Safety Injection; then trip RCP 13.
C. check if No.1 and No. 2 seal status allow continued RCP operation.
D. verify CCW flow to the RCP thermal barrier.
Proposed Answer: B direct the operator to trip the reactor and enter E-0 Reactor Trip or Safety Injection; then trip RCP 13 Explanation:
A. Incorrect. 1) IAW AOP-28, RCP Malfunction Section A, these actions should be performed - as indicated by both PK 5-01 and 5-05 having alarmed. 2) Probable causes include RCP Motor Oil Cooler CCW temperature too LOW but not Low RCS pressure or High Seal Injection Temperature. RCP 1-3 does not have an associated spray valve.
B. Correct. RCP trip is required.
C. Incorrect. 1) These would be the correct actions for a No. 1 seal failure; which would be indicated by abnormal differential pressure across RCP seal number 1. 2) Low RCS pressure would not cause PK 5-01.
D. Incorrect. 1) No.1 seal differential pressure is normal, therefore, this is the wrong action to take. 2). High Seal Injection Temperature would not cause PK 5-01.
Technical
References:
Operations Lesson LAR-1 RCP Malfunctions, AOP-28 RCP Malfunction
DCPP L111 NRC Exam 21 November, 2012 P a g e l 115 References to be provided to applicants during exam: None Learning Objective: Explain the causes of RCP abnormal conditions. (6104) Given an abnormal condition, summarize the major actions of the AP-28 to mitigate an event in progress.
(3477)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 116 Examination Outline Cross-Reference Level SRO Knowledge of the Steam Generator system parameters and logic used to assess the status of the safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
Tier #
2 Group #
2 K/A #
035 2.4.21 Rating 4.6 Question 92 GIVEN:
- Safety Injection has actuated
- RCPs are secured
- Containment pressure peaked at 4.2 psig and is now reading 2.5 psig
- Total AFW flow is 400 gpm Steam Generator Parameters:
Steam Generator Narrow Range Level Pressure 1-1 0%
0 PSIG 1-2 14%
1060 PSIG 1-3 19%
1040 PSIG 1-4 22%
1135 PSIG The WCSFM is evaluating the Heat Sink Critical Safety Function on SPDS.
What is the expected terminus color and what will be displayed on the SPDS panel for F-0.3, Heat Sink?
A. RED; GO TO FR-H.1 because of an inadequate heat sink B. YELLOW; GO TO FR-H.2 because of high steam generator pressure C. YELLOW; GO TO FR-H.5 because of low steam generator level D. GREEN; CSF SAT because the heat sink critical safety function is satisfied Proposed Answer: B YELLOW; GO TO FR-H.2 because of high steam generator pressure.
Explanation:
A incorrect. Because containment pressure is below adverse setpoint, non-adverse steam generator levels are used. Minimum level for heat sink is 15% in ONE steam generator.
Currently two are above the required setpoint. If adverse numbers are used, all would be low and
DCPP L111 NRC Exam 21 November, 2012 P a g e l 117 H.1 would be correct.
B correct. Steam generator pressure of greater than 1115 psig is setpoint. Because it is the next parameter checked, it would be the terminus (H.5 is also not satisfied).
C incorrect. Not all steam generators are above 15% however, the high pressure is the higher priority.
D incorrect. High pressure and low level exist.
Technical
References:
FR-0, Critical Safety Function Status Trees, Revision 17 References to be provided to applicants during exam: None Learning Objective: 3856 - State the parameters monitored by CSFSTs Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis 10CFR Part 55 Content:
55.43.5
DCPP L111 NRC Exam 21 November, 2012 P a g e l 118 Examination Outline Cross-Reference Level SRO Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. Station Air Tier #
2 Group #
2 K/A #
079 2.4.4 Rating 4.7 Question 93 GIVEN:
- Unit 1 is at 100% power.
- Tavg is 570ºF
- PZR Level is 55%
- PZR pressure is slightly above normal and slowly rising
- PI-380, Instrument Air Header Pressure reads 75 psig and is slowly lowering
- Feed Water Regulating valves are closing
- Steam Generator Water Levels are lowering The Shift Foreman should enter:
A. AP-5, Malfunction of Protection or Control Channel.
B. AP-9, Loss of Instrument Air.
C. AP-13, Malfunction of Reactor Pressure Control System.
D. AP-15, Loss of Feedwater Flow.
Proposed Answer: B AP-9, Loss of Instrument Air Explanation:
A. Incorrect.
B. Correct. From the given indications the correct procedure to enter is a loss of instrument air.
C. Incorrect.
D. Incorrect.
Technical
References:
AP-9, Loss of Instrument Air; E-0, Reactor Trip or Safety Injection; AP-15, Loss of Feedwater Flow; Operations Lesson LPA-9, Loss of Instrument Air.
References to be provided to applicants during exam: None Learning Objective: Given initial conditions, assumptions, and symptoms, determine the correct abnormal operation procedure to be used to mitigate an operational event. (3478) List the effects that a loss of Instrument Air would have on the plant. (3541)
DCPP L111 NRC Exam 21 November, 2012 P a g e l 119 Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.2 / 45.6
DCPP L111 NRC Exam 21 November, 2012 P a g e l 120 Examination Outline Cross-Reference Level SRO Knowledge of procedures and limitations involved in core alterations Tier #
3 Group #
K/A #
2.1.36 Rating 4.1 Question 94 Unit 1 is currently performing core loading in accordance with OP B-8DS2, Core Loading.
Which of the following would result in the suspension of core loading?
A. An unexpected increase in count rate on ONE responding nuclear channel by a factor of 2.
B. An unexpected increase in count rate on TWO responding nuclear channels by a factor of 3.
C. A change in Reactor Coolant System temperature of 15°F.
D. A change in boron concentration from the nominal value at the start of core loading by +/-30 ppm.
Proposed Answer:
B An unexpected increase in count rate on TWO responding nuclear channels by a factor of 3 Explanation:
A. Incorrect - OP B-8DS2 specifies that an unexpected increase of one responding nuclear channel by a factor of 3 will result in a suspension of core loading.
B. Correct - OP B-8DS2 specifies that an unexpected increase of one responding nuclear channel by a factor of 3 will result in a suspension of core loading. Therefore, two responding channels with unexpected increase in count rate by a factor of three would result in a suspension of core loading.
C. Incorrect - OP B-8DS2 specifies a change in Reactor Coolant System temperature of greater than 20°F will result in a suspension of core loading.
D. Incorrect - OP B-8DS2 specifies a change in boron concentration from the nominal value at the start of core loading by +/-50 ppm will result in a suspension of core loading.
Technical
References:
OP B-8DS2, Core Loading, Revision 50 References to be provided to applicants during exam: None Learning Objective: 36965 - Discuss significant precautions and limitations associated with the Fuel Handling system Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.43.6
DCPP L111 NRC Exam 21 November, 2012 P a g e l 121 Examination Outline Cross-Reference Level SRO Ability to identify and interpret diverse indications to validate the response of another indication. Conduct of Operations Tier #
3 Group #
K/A #
2.1.45 Rating 4.3 Question 95 GIVEN:
- Unit 1 is at 100% power
- Tave is 572ºF
- LT-459 indicates 60%, LT-460 indicates 53%, LT-461 indicates 56%.
- PT-455 indicates 2235 psig, PT-456 indicates 2100 psig, PT-457 indicates 2215 psig.
- RM-3, Oily Water Separator Detector, needle is pegged low.
- RM-11, Containment Air Particulate Detector, FILTER NOT IN MOTION light is OUT.
Based on these indications; A. LT-460 and PT-456 are inoperable B. LT-460 and RM-11 are inoperable C. PT-456 and RM-3 are inoperable D. LT-460, PT-456, RM-3, and RM-11 are all inoperable Proposed Answer: C PT-456 and RM-3 are inoperable Explanation:
A. Incorrect. PT-456 is operable because Pressurizer Level channels may be 8.5% different.
B. Incorrect. RM-11 is operable; this is the expected condition for channel checks of Westinghouse radiation monitors.
C. Correct. PT-456 is greater than 5% different from the other channels and RM-3 should have slight needle oscillations.
D. Incorrect. See above.
Technical
References:
Channel Check Criteria B-5 Rev. 5 References to be provided to applicants during exam: None Learning Objective: 56218 - Discuss operator behaviors and practices related to the operator fundamental of closely monitoring plant indications and conditions Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5 / 45.4
DCPP L111 NRC Exam 21 November, 2012 P a g e l 122 Examination Outline Cross-Reference Level SRO Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
Tier #
3 Group #
K/A #
2.2.36 Rating 4.2 Question 96 Unit 1 is operating at 100% power and:
- TDAFW Pump 1-1 is inoperable due to maintenance
- Twelve hours later the Emergency Diesel Generator 1-3 is declared inoperable due failure of the Monthly Surveillance test.
In order to be in compliance with the Technical Specifications, the maximum time allowed to be in MODE 3 is ____________ hours.
A. 6 B. 10 C. 66 D. 70 Proposed Answer: B. 10 Explanation:
A. Incorrect - This is a direct entry into TS 3.7.5 C.1 due to Two AFW Trains inoperable in Modes 1, 2 or 3. The second train does not become inoperable until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the EDG is declared inoperable according to TS 3.8.1 B.2.
B. Correct - This includes the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from TS 3.8.1 B.2 and the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from TS 3.7.5 C.1 C. Incorrect - This includes the 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> left on TS 3.7.5 B.1 plus the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from TS 3.7.5 C.1. TS 3.7.5 C.1 is entered after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> have expired after declaring the EDG inoperable (TS 3.8.1 B.2).
D. Incorrect - This includes the 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> left on TS 3.7.5 B.1 plus the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from TS 3.7.5 C.1 and the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from TS 3.8.1 B.2.
Technical
References:
Technical Specifications 3.7.5 and 3.8.1 References to be provided to applicants during exam: Technical Specifications 3.7.5 and 3.8.1 Learning Objective: 9697G - Apply TS 3.7 Technical Specification LCOs Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental
DCPP L111 NRC Exam 21 November, 2012 P a g e l 123 Comprehensive/Analysis X
10CFR Part 55 Content:
55.43.2
DCPP L111 NRC Exam 21 November, 2012 P a g e l 124 Examination Outline Cross-Reference Level SRO Ability to comply with radiation work permit requirements during normal or abnormal conditions. Radiation Control Tier #
3 Group #
K/A #
2.3.7 Rating 3.6 Question 97 GIVEN:
- The Shift Manager has just declared an Alert
- The Emergency Operations Facility and the Technical Support Center have NOT yet been activated; no turnovers have occurred For the purpose of Radiological Assessment Sampling, workers will be exposed to radiation levels in excess of the 10 CFR Part 20 exposure limits.
In the current situation, the ____________ has the authority to authorize the dose on an EP RB-2, Attachment 9.7, Emergency Exposure Permit.
A. Shift Manager ONLY B. Shift Manager OR the Site Emergency Coordinator C. Site Emergency Coordinator OR the Radiation Protection Manager D. Radiation Protection Manager OR the Plant Manager Proposed Answer: A Shift Manager ONLY Explanation: A (minor formatting)
A. Correct. Only the Emergency Director or in this case, the Shift Manager (acting ED) has the has the unilateral authority and non delegable responsibility for authorizing an individual emergency worker to exceed normal 10 CFR 20 exposure limits.
B. Incorrect. The TSC is not yet manned, the SEC is not yet stationed.
C. Correct. The Shift Manger has Command and Control of the emergency response for the given plant conditions and until a turnover happens with the SEC or ED.
D. Incorrect. As explained above.
Technical
References:
RCP-D-201, RWP Writing Guide; AP-31, Rapid Containment Entry; Operations Lesson LEP 3 EP RB Procedures.
References to be provided to applicants during exam: None Learning Objective: State who may authorize emergency doses. (9848)
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental X
DCPP L111 NRC Exam 21 November, 2012 P a g e l 125 Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 41.12 / 45.10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 126 Examination Outline Cross-Reference Level SRO Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.
Tier #
3 Group #
K/A #
2.3.12 Rating 3.7 Question 98 Unit 1 is 100%
A non-emergency containment entry is being performed in accordance with RCP D-230, Radiological Control for Containment Entry.
The Moveable Incore Detector keys are maintained in the possession of the _______ during all containment entries.
A. Operations Manager B. Shift Manager C. Radiological Protection Manager D. Radiological Protection Foreman Proposed Answer: D Radiological Protection Foreman Explanation:
A. Incorrect - RCP D-230 states that the RP Foreman or designee maintains the MIDS keys during containment entries.
B. Incorrect - RCP D-230 states that the RP Foreman or designee maintains the MIDS keys during containment entries.
C. Incorrect - RCP D-230 states that the RP Foreman or designee maintains the MIDS keys during containment entries.
D. Correct - RCP D-230 states that the RP Foreman or designee maintains the MIDS keys during containment entries.
Technical
References:
RCP D-230, Radiological Control for Containment Entry, Revision 21 References to be provided to applicants during exam: None Learning Objective: 7913 - State the admin requirements for acquiring plant keys Question Source:
Bank #
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Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis 10CFR Part 55 Content:
55.45.9-10
DCPP L111 NRC Exam 21 November, 2012 P a g e l 127
DCPP L111 NRC Exam 21 November, 2012 P a g e l 128 Examination Outline Cross-Reference Level SRO Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. Emergency Procedures/Plan Tier #
3 Group #
K/A #
2.4.22 Rating 4.4 Question 99 Given the following conditions following a reactor accident:
- RCS has cooled down from 572ºF to 481ºF in the last hour
- RCS pressure is 460 psig
- RCPs are secured
- All Core Exit Thermocouples indicate approximately 650ºF
- RVLIS indicates 25%
- Containment Pressure is 24 psig and no Containment Spray Pumps are running
- All TC instruments indicate 225ºF In accordance with the bases for F-0, Critical Safety Function Status Trees, which Safety Function has the highest priority and why?
A. Core Cooling is magenta. When multiple Safety Functions are under severe challenge (magenta), the highest priority safety function is restored first; Core Cooling is the highest priority safety function under severe challenge.
B. RCS Integrity is magenta. The 6 safety functions have an order of priority; RCS Integrity is the highest priority under a severe challenge and is therefore addressed first.
C. RCS Integrity is magenta. The safety function with the highest severity level is addressed first. The other safety functions are yellow.
D. Containment is magenta. The safety function with the highest severity level is addressed first. The other safety functions are yellow.
Proposed Answer: A Core Cooling is magenta. When multiple Safety Functions are under severe challenge (magenta), the highest priority safety function is restored first; Core Cooling is the highest priority safety function under severe challenge.
Explanation:
A. Correct. IF no extreme challenges exist, THEN the operator shall stop procedure in effect and initiate functional restoration to restore the highest priority critical safety function under severe challenge. Core Cooling, RCS Integrity, and Containment are under severe challenge (magenta) and Subcriticality is Not Satisfied (yellow). Therefore, Core Cooling is the highest priority safety function at this time.
B. Incorrect. True - RCS Integrity is under severe challenge (magenta). The bases further prioritize the safety functions during accidents based on the level of the challenge to each function. RCS Integrity is not a higher priority than Core Cooling.
DCPP L111 NRC Exam 21 November, 2012 P a g e l 129 C. Incorrect. RCS Integrity is also under severe challenge (magenta) but is a lower priority than another safety function under severe challenge. Core Cooling is also under severe challenge and is higher priority. Containment is also under severe challenge.
D. Incorrect. Core Cooling is under severe challenge (magenta). The second part of the answer is a true statement but it does not have any effect on the priority of the safety functions.
Technical
References:
F-0, Critical Safety Function Status Trees; Westinghouse Background Document Emergency Response Guidelines F-0 Critical Safety Function Status Trees References to be provided to applicants during exam: Critical Safety Function Status Trees (C, I, Z)
Learning Objective: Apply the Rules of Usage in s for the CSFSTs and FRGs, including:
the six status trees the priority of use of the status trees the priority of use of the color of each CSF when to monitor and/or implement the CSFSTs and FRGs (38107)
Question Source:
Bank #
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Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
10CFR Part 55 Content:
CFR: 43.5 / 45.12
DCPP L111 NRC Exam 21 November, 2012 P a g e l 130 Examination Outline Cross-Reference Level SRO Knowledge of the lines of authority during implementation of the emergency plan.
Tier #
3 Group #
K/A #
2.4.37 Rating 4.1 Question 100 The ______________has the authority to downgrade an Emergency Classification Level at the A. Shift Manager; Notice of Unusual Event level, ONLY B. Site Emergency Coordinator; Notice of Unusual Event level, ONLY C. Shift Manager; Alert or Unusual Event level D. Site Emergency Coordinator; Alert or Unusual Event level Proposed Answer: A Shift Manager; Notice of Unusual Event level, ONLY (minor formatting)
Explanation:
A. Correct - EP G-1 states that the Shift Manager may downgrade an Unusual Event to no ECL.
B. Incorrect - EP G-1 states that the Shift Manager may downgrade an Unusual Event to no ECL. The Site Emergency Coordinator may upgrade an event to a higher ECL until the Emergency Director assumes responsibility in the EOF.
C. Incorrect - EP G-1 states that the Shift Manager and Site Emergency Coordinator shall not downgrade an event classified at the Alert or Higher level at any time.
D. Incorrect - EP G-1 states that the Shift Manager and Site Emergency Coordinator shall not downgrade an event classified at the Alert or Higher level at any time.
Technical
References:
EP G-1, Emergency Plan and Emergency Plan Activation, Revision 40 LEP-2, Emergency Plan Procedures, Revision 12 References to be provided to applicants during exam: None Learning Objective: 5270 - As described in EP G-1, state the responsibilities of Shift Manager.
Question Source:
Bank #
(note changes; attach parent)
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content:
55.45.13