ML16336A573
| ML16336A573 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/20/2016 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | |
| References | |
| Download: ML16336A573 (151) | |
Text
U.S. Nuclear Regulatory Commission Diablo Canyon SRO Written Examination Applicant InformationName: KEY Date: 14 October, 2016 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 fi nal grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items. Examination papers will be collected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the examination begins.
Applicant Certification All work done on this examination is my ow
- n. I have neither gi ven nor received aid.
______________________________________
Applicant's Signature ResultsRO/SRO-Only/Total Examination Values
/ / Points Applicant's Scores / / Points Applicant's Grade / / Percent
DCPP NRC Exam 14 October 2016 i Multiple Choice (Circle your choice) NAME:___ KEY_______
If you change your answer, write your se lection in the bl ank and initial.
- 1. 26. 2. 27.3. 28.4. 29. 5. 30.6. 31. 7. 32. 8. 33.9. 34. 10. 35. 11. 36.12. 37. 13. 38. 14. 39.15. 40. 16. 41. 17. 42.18. 43. 19. 44. 20. 45.21. 46. 22. 47. 23. 48.24. 49. 25. 50. A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D DCPP NRC Exam 14 October 2016 ii Multiple Choice (Circle your choice) NAME:__KEY_____
If you change your answer, write your se lection in the bl ank and initial.
- 51. 76. 52. 77.53. 78. 54. 79. 55. 80.56. 81. 57. 82. 58. 83.59. 84. 60. 85. 61. 86.62. 87. 63. 88. 64. 89.65. 90. 66. 91. 67. 92.68. 93. 69. 94. 70. 95.71. 96. 72. 97. 73. 98.74. 99. 75. 100. A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D A B C D DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 003 A1.08
- Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including Seal water temperature.
Tier # 2 Group # 1 K/A # 003 A1.08 Rating 2.5 Question 01 GIVEN: RCS Pressure is 800 psig All RCPs are running RCP 1-1 indications on VB2:
o Seal Injection flow is 10 gpm to each RCP o #1 Seal leakoff flow is 0.4 gpm on FR-157 o Seal water leakoff temperature is 1 75°F, rising slowly on TI-148 o Radial bearing temperature is 1 7 5°F, rising slowly on TI-1 6 5 The operator is directed to open CVCS-1-8142 #1 Seal Bypass.
When the crew opens CVCS-1-8142, ________________ temperature should lower to prevent exceeding the OP AP-28, Reactor Coolant Pump Malfunctions, RCP trip setpoint of
__________.
A. RCP radial bearing outlet; 200°F B. #1 Seal Outlet; 200°F C. RCP radial bearing outlet; 2 2 5°F D. #1 Seal Outlet; 2 2 5°F Proposed Answer:
C. RCP radial bearing outle t; 2 2 5°F Explanation:
A. Incorrect. Opening 8142 will lower seal outlet temperature and radial bearing temperature, however, 200°F is the trip setpoint for any motor bearing temperature in AP
-28. B. Incorrect. Opening 8142 will lower seal outlet temperature and radial bearing temperature, however, 200°F is the trip setpoint for any motor bearing temperature in AP
-28. C. Correct. Opening 8142 should lower seal outlet temperature and radial bearing temperature, 2 25°F is the trip setpoint for radial bearing outlet temperature in AP
-28 and (seal outlet is 2 3 5°F). On sim the noticeable effect is radial bearing temperature.
D. Incorrect. Opening 8142 will lower seal outlet temperature and radial bearing temperature, however, 2 25°F is the trip setpoint for radial bearing outlet temperature in AP
-28. Seal temperature limit is 235°F.
Technical References
- LA6 Reactor Coolant Pump
. AR PK05-01 RCP No. 11 References to be provided to applicants during exam: None Learning Objective
- Describe RCP components: seal return. (35743) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
DCPP L 1 61 Exam Rev 0 10CFR Part 55 Content:
55.4 1.5 / 45.5 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 003 G2.2.36
- RCP: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
Tier # 2 Group # 1 K/A # 003 G2.2.36 Rating 3.1 Question 02 GIVEN: Unit 1 is in M ODE 3 Control Rods are incapable of being withdrawn RCPs 1-1, 1-2, are OPERABLE and running RCP 1-3 is inoperable for breaker replacement RCP 1-4 is stopped, OPERABLE and capable of being started
12 kV bus 1E breaker trip s on overcurrent
.
Currently there is/are _______________ and LCO 3.4.5, RCS LOOPS - MODE 3 is _________.
A. two RCS Loops are OPERABLE with one loop in operation; NOT met
B. two RCS Loops are OPERABLE with one loop in operation; met C. only one RCS Loop is OPERABLE and in operation; NOT met
D. o nly on e RCS Loop is OPERABLE and in operation; met Proposed Answer:
B. two RCS Loops are OPERABLE with one loop in operation; met Explanation:
A. Incorrect because the LCO is met. Plausible because two loops are required to be in operation when rods are capable of being withdrawn.
B. Correct. 12 kV bus 1E trips, de
-energizing RCPs 1
-1 and 1-3. RCP 1-2 is OPERABLE and in operation, RCP 1
-4 is OPERABLE and stopped. In mode 3 with rods incapable of being withdrawn (reactor trip breakers open) only one loop is required to be in operation.
C. Incorrect because two loops are OPERABLE. Plausible if RCPs 1
-2 and 1-4 are thought to be on bus 1E and the others on 1D.
D. Incorrect because two loops are OPERABLE. Plausible because only one loop is required to be in operation when rods are incapable of being withdrawn.
Technical References
- LCO 3.4.5 RCS Loops Mode 3; OIM J 1 References to be provided to applicants during exam: None Learning Objective
- Discuss significant Technical Specifications and Equipment Control Guidelines associated with the RCP.
Apply TS 3.3, 3.4, and 3.6 Technical Specification LCOs.
Apply TS 3.3, 3.4, and 3.6 Technical Specification bases (SROs only).
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
DCPP L 1 61 Exam Rev 0 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 004 A1.09
- Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including RCS pressure and temperature.
Tier # 2 Group # 1 K/A # 00 4 A1.09 Rating 3.6 Question 03 Given: One train of RHR is in service All RCPs are stopped The RCS and Pressurizer are solid PCV-135, Letdown Pressure Control is in MANUAL Tcold for all loops is 135 °F Steam Generator 1
-2 temperature is 190 °F The crew is preparing to start RCP 1
-2. Based on these conditions, what action should the operators be prepared to take to maintain pressure within limits when the RCP is started?
PCV-135 controller output (demand) should be ______________
to prevent
_____________. A. raise d; violating minimum pressure for RCP operation B. raise d; lifting the LTOP relief C. lower ed; violating minimum pressure for RCP operation D. lower ed; lifting the LTOP relief Proposed Answer:
B. raise d; lifting the LTOP relief Explanation
- A. Incorrect because pressure will rise , and minimum pressure for RCP operations will not be an issue. Plausible because when starting a n RCP to perform pump sw aps to fill SG tubes, RCS pressure is expected to lower, but in this case the RCS and Pressurizer are solid. Also plausible because PCV
-135 is fail open, air to close, and it may be thought that higher controller output is needed to close the valve to raise pressure.
B. Correct.SG temperature is higher than Tc so when the RCP is started and flow initiates through the SG, RCS temperature will rise, and there will be a corresponding rise in Pressurizer Pressure. To check this rise in pressure and prevent the potential for lifting the LTOP relief, PCV
-135 must be opened more, requiring the operator to raise controller output. C. Incorrect because pressure will rise, therefore minimum pressure for RCP operation is not an issue. Plausible because lowering controller output is the correct response for lowering pressure. D. Incorrect because lowering controller output is the wrong response to pressure rising. Plausible because pressure will rise due to the dT between the RCS and SG.
Technical References
- OP L-1 Plant Heatup From Hot Shutdown to Hot Standby, P&L 6.1.2.a (pressure) and 6.1.2.k (temperature); LB
-1A Chemical and Volume Control System, page 19.
References to be provided to applicants during exam: None DCPP L 1 61 Exam Rev 0 Learning Objective
- State the purpose of CVCS components. (35749)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.5 Difficulty: 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 005 K5.05
- Knowledge of the operational implications of the plant response during "solid plant": pressure change due to the relative incompressibility of water as they apply the RHRS.
Tier # 2 Group # 1 K/A # 005 K5.05 Rating 2.7 Question 04 Given: Unit 1 is in MODE 5 The RCS is solid at 350 psig One train of RHR is in service One charging pump is running PCV-135, Letdown Pressure Control Valve, is in AUTOMATIC The running RHR pump trips.
Initially , RCS pressure will _______, and letdown pressure will _________.
A. rise; rise B. lower; rise C. rise; lower D. lower; lower Proposed Answer:
C rise; lower Explanation:
A. Incorrect because PCV
-135 will throttle closed. When the RHR pump tripped, RHR discharge pressure, and the pressure at the letdown line, will lower. PCV
-135 will throttle closed to attempt to maintain pressure.
B. Incorrect because RCS pressure rises due to charging adding inventory , while no RHR pump is running to remov e inventory.
C. Correct. Flow into (charging) and out of (RHR/letdown) the RCS was initially balance
- d. When the RHR pump tripped, two things resulted: 1) the RCS pressure rose due to the flow imbalance, and 2) letdown line pressure lowered due to the loss of discharge pressure when
-135 will respond to the lowering letdown line pressure by closing to attempt to maintain pressure at setpoint.
D. Incorrect because RCS pressure rises. Plausible because if RCS pressure lowers, PCV
-135 would normally be expected to throttle closed.
Technical References
- OIM B-1-1 References to be provided to applicants during exam: None Learning Objective
- Discuss solid plant operation (23186)
Question Source:
Bank # X (note changes; attach parent)
Modified Bank #
New NRC Exam Question DCPP 04/2007 #3 X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.41.5 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 006 K4.14
- Knowledge of ECCS design feature(s) and/or interlock(s) which provide for Cross
-Connection of HPI/LPI/SIP.
Tier # 2 Group # 1 K/A # 006 K4.14 Rating 3.9 Question 05 GIVEN: A LBLOCA has occurred on Unit 1 Containment spray has actuated The Control Room is aligning the ECCS for recirculation The crew is preparing to CLOSE 8974A and B, SI Pump recirc valve
- s. Which of the following lists the action(s) that must be taken to close 8974A
? 1. Place Series Contactor in "CUT
-IN" 2. Close 8804A OR 8804B
- 3. Close 8804A AND 8804B A. 1 only B. Both 1 and 2 C. 2 only D. 3 only Proposed Answer:
A. 1 only Explanation:
A. Correct To close the recirc valve, the only action required is to cut
-in the series contactor.
B. Incorrect. 8804A and B are interlocked with 8974A and B. To open 8974 A (B), both must be closed. (Note:To open 8804A, 8974A or B must be closed
.) C. Incorrect This is a common setup, such as with 8804A/B, only requires one of the valves to be closed. D. Incorrect. To OPEN the valve, both valves must be closed. Technical References
- L B-2; LB-3 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 Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 007 K5.02
- Knowledge of the operational implications of the method of forming a steam bubble in the PZR as they apply to PRTS. Tier # 2 Group # 1 K/A # 007 K5.02 Rating 3.1 Question 0 6 The crew is preparing to draw a bubble in the pressurizer using OP A
-2:IX, "Reactor Vessel
-Vacuum Refill of the RCS
" by raising PRT level to 85%.
- 1. PRT level raised to 85% to:
A. 1. prevent air-in leakage past the PORVs and Safety Valves
- 2. 10 psig B. 1. prevent air-in leakage past the PORVs and Safety Valves
- 2. 14.7 psig C. 1. purge out air via the vent header
- 2. 10 psig D. 1. purge out air via the vent header
- 2. 14.7 psig Proposed Answer:
A. 1. prevent air-in leakage past the PORVs and Safety Valves 2. 10 psig Explanation:
A. Correct. OP A-2:IX note on page 13 states that level is raised to 85% to prevent air in leakage past the PORVs and Safety Valves. This is accomplished by raising level enough to submerge all openings in the sparger line so it is not open to the PRT gas space.
During the fill, OP A
-4B:I states that pressure shall not exceed 10 psig B. Incorrect. Pressure shall not exceed 10 psig. If its thought that because the procedure for drawing a bubble is "vacuum refill", then atmospheric pressure is a plausible limit
. C. Incorrect. Its plausible that submerging the sparger will purge air out of the line (which then could aid in the vacuum refill).
D. Incorrect. As stated above, both plausible to aid in the vacuum refill Technical References
- OP A-2:IX, OP A-4B:I. References to be provided to applicants during exam: None Learning Objective
- Draw a bubble in the Pressurizer (28370)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.10 Difficulty: 3.
0 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 008 G2.1.28
- CCW: Knowledge of the purpose and function of major system components and controls.
Tier # 2 Group # 1 K/A # 008 G2.1.28 Rating 4.1 Question 07 A CCW pump Standby Select switch must be in AUTO in order for the pump to automatically start in response to:
- 1) Low System Pressure
- 2) Safety Injection
- 3) Transfer to Diesel A. 1 only B. 1 and 2 only C. 1 and 3 only D. 1, 2, and 3 Proposed Answer:
A. 1 only Explanation:
All the listed possibilities are trips of the CCW pump and therefore plausible if its not known which require the MANUAL/AUTO switch to be in AUTO to be an active trip.
A. Correct. The Standby Select switch does not affect the pump's response to a Safety Injection or loss of power resulting in the bus transferring to Diesel. In those cases, the pumps will start regardless of whether or not it is in Auto.
If in Auto, the pump will start in response to low system pressure.
B. Incorrect because Auto is not required for the pump to start in response to a SI
. C. Incorrect because Auto is not required for the pump to start in response to a Transfer to Diesel. D. Incorrect because Auto is not required for the pump to start in response to a SI or Transfer to Diesel. Technical References
- LF-2, Component Cooling Water page 9.
References to be provided to applicants during exam: None Learning Objective
- Analyze automatic features and interlocks associated with the CCW System. (35487)
Question Source:
Bank # (note changes; attach parent)
Modified Ba nk # New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.7 Difficulty: 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 008 k3.03Knowledge of the effect that a loss or malfunction of the CCWS will have on the RCP.
Tier # 2 Group # 1 K/A # 008 K3.03 Rating 4.1 Question 08 GIVEN: Unit 1 is at 100% power Valve FCV-750, RCP Thermal Barriers RTN ISOL VLV I.C., closes PK01-08, CCW HEADER C, alarms (input 428, RCP Thermal Barrier CCW Flo Lo)
RCP seal injection flows are approximately 9 gpm on all RCPs Which of the following describes how the above conditions affect RCP 1
-1 pu mp and motor radial bearing cooling? A. Cooling is lost to both the pump and motor radial bearing
- s. B. Cooling is maintained to both the pump and motor radial bearing
- s. C. Pump radial bearing cooling is lost. Motor radial bearing cooling is maintained
. D. Pump radial bearing cooling is maintained. Motor radial bearing cooling is lost.
Proposed Answer:
B. Cooling is maintained to both the pump and motor radial bearing
- s. Explanation:
A. Incorrect. Plausible because it may be thought that the isolated line supplies not only the thermal barrier, but motor cooling also.
B. Correct. The thermal barrier cooling being isolated will not affect pump radial bearing cooling as long as seal injection is maintained. This cooling line does not supply motor cooling, so those bearings will not be affected.
C. Incorrect because pump radial bearing is unaffected as long as seal injection is still in service. Plausible because motor cooling is unaffected
, and it may be thought that thermal barrier cooling provides some cooling to water going through the pump radial bearing.
D. Incorrect because the isolated line does not provide motor cooling. Plausible because it may be thought that it does provide motor cooling, and it may be remembered that pump bearing cooling is not affected because seal injection is in service.
Technical References
- LA-6 Reactor Coolant Pumps References to be provided to applicants during exam: None Learning Objective
- Explain the causes of RCP abnormal conditions (28536)
Question Source:
Bank # P-36011 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 010 A4.03
- Pressurizer Pressure Control System: Ability to manually operate and/or monitor PORV and block valves in the control room.
Tier # 2 Group # 1 K/A # 010 A4.03 Rating 4.0 Question 0 9 The crew is performing FR
-S.1, Response to Nuclear Power Generation/ATWS
. At step 4, when Pressurizer pressure is checked, RCS pressure is 234 5 psig. PORVs indicate as shown below:
- 1. What action should be taken by the operator?
- 2. What is the basis for checking RCS pressure?
A. 1. Open the PORVs
- 2. To prevent passing two phase flow through the safety valves C. 1. Continue to monitor the PORVs and ensure they open if pressure rises to setpoint
- 2. To allow sufficient borated injection flow into the RCS D. 1. Continue to monitor the PORVs and ensure they open if pressure rises to setpoint
- 2. To prevent passing two phase flow through the safety valves Proposed Answer:
A. 1. Open the PORVs
- 2. To allow sufficient borated injection flow into the RCS Explanation:
A. (allow boric acid flow) Correct. From FR
-S.1
Background:
The check on RCS pressure is intended to alert the operator to a condition which would reduce charging or SI pump injection into the RCS and, therefore, boration. The PRZR PORV pressure setpoint is chosen as that pressure at which flow into the RCS is insufficient. The contingent action is a rapid depressurization to a pressure which would allow increased injection flow. When primary pressure drops 200 psi below the PORV pressure setpoint, the PORVs should be closed. The operator must verify successful closure of the PORVs, closing the isolation valves, if necessary.
DCPP L161 Exam Rev 0 B. (2 phase flow) Incorrect. 2 phase flow through safeties is a concern for accidents such as steam generator safeties and overf ill, but not the bases for this check of pressure in FR-S.1. However, ATWS is an accident that challenges RCS pressure and design basis pressure spike is greater than the safety setpoint.
C. Incorrect. The PORVs should be open. Plausibl e if setpoint not known, reason is correct.
D. Incorrect. The PORVs should be open. Plausible if setpoint mistaken and reason thought to be related to lifting safeties.
Technical References
- FR-S.1 background References to be provided to applicants during exam:
None Learning Objective: Explain basis of emergency procedure step (7920) Question Source: Bank # (note changes; attach parent) Modified Bank # L091 Question 5 X New Question History: Last Two NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.41.5 Difficulty: 3.0 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 012 K2.01
- Knowledge of bus power supplies to the RPS channels, components, and interconnections.
Tier # 2 Group # 1 K/A # 012 K2.01 Rating 3.3 Question 10 A loss of PY
-12 occurs.
As a result of the loss of PY
-12, monitor lights for Channel II SSPS bistables are ______ and monitor lights for SSPS bistable on Channel ___________.
A. lit; III (Blue); will not be lit B. not lit; III (Blue) remain lit C. lit; I (Red) will not be lit D. not lit; I (Red) remain lit Proposed Answer:
A. lit; III (Blue) will not be lit Explanation:
A. Correct. Channel II relays are powered from PY
-12. Channel III bistable status lights (not the relays) are powered from PY
-12. B. Incorrect. Relays trip for channel II and lights go out for channel III. Plausible as the indications are reversed, or its thought that de
-energizing the bistables also turns out the light. C. Incorrect. Channel II bistables will be lit, Channel I status lights are powered from PY
-14 not PY-12. Plausible because the status lights for a channel is powered from the previous channel (II powers III) not the reverse (III power s the lights for II)
. D. Incorrect. relays will be lit and channel I lights are powered from PY
-14. Plausible, if its thought the loss of power causes the relays to lose power and the lights go out.
Technical References
- OP AP-4, LB-6B References to be provided to applicants during exam: None Learning Objective
- State the power supplies to Eagle 21 and Solid State Protection System components. (3291)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
41.7 Difficulty: 2.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 012 K6.03
- Knowledge of the effect of a loss or malfunction of the trip logic circuits will have on the RPS.
Tier # 2 Group # 1 K/A # 012 K6.03 Rating 3.1 Question 11 Given: Unit 1 is operating at 100% power Pressurizer Level Channel LT
-459 is in BYPASS While the channel is in BYPASS , Pressurizer Level channel LT-460 fails HIGH.
Which of the following describes the expected plant response?
A. The reactor trips due to satisfying the 2 of 3 High Pressurizer Level trip coincidence
. B. The reactor trips due to satisfying the 2 of 4 High Pressurizer Level trip coincidence.
C. The reactor does not trip due to NOT satisfying the 2 of 3 High Pressurizer Level trip coincidence
. D. The reactor does not trip due to NOT satisfying the 2 of 4 High Pressurizer Level trip coincidence
. Proposed Answer:
C. The reactor does not trip due to NOT satisfying the 2 of 3 High Pressurizer Level trip coincidence.
Explanation: , For Pressurizer high level, the logic is 2 of 3 channels to trip, (unlike trips, such Pressurizer High pressure 2 of 4 channels) or if they include cold cal channel
. BYPASS does not trip a channel and the channel does not input to the trip logic.
A. Incorrect. The reactor will not trip.
While in Bypass, the channel will not trip and the matrix is both of the remaining channels (2) to trip. Logic is 2 of 3.
Plausible if its thought the channel is tripped.
B. In correct The reactor will not trip. While in Bypass, the channel not cause a trip and the matrix is both of the remaining channels (2) to trip. Plausible
- Logic is 2 of 3, not 2 of 4 as it is for trips such as high or low pressurizer pressure.
C. Correct. The logic is still 2 channels to trip and only channel has tripped (LT
-460) D. Incorrect The logic is still 2 channels to trip and only channel has tripped (LT
-460). Plausible
- The trip logic is 2 of 3 not 2 of 4, as it is for trips such as high or low pressurizer pressure. Technical References
- OIM B-6-4b References to be provided to applicants during exam: None Learning Objective
- 37051 - Discuss abnormal conditions associated with the RPS Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 013 K2.01
- Knowledge of bus power supplies to ESFAS/safeguards equipment control.
Tier # 2 Group # 1 K/A # 013 K2.01 Rating 3.6 Question 12 PY-14 de-energizes while the unit is at full power.
Subsequently, a loss of PY
-11 occurs. As a result:
A. Only one train of Safety Injection actuates due to a loss of power to the MASTER relays for one train. B. Only one train of Safety Injection actuates due to a loss of power to the SLAVE relays for one train. C. Safety Injection does not actuate due to a loss of power to the MASTER relays for both trains. D. Safety Injection does not actuate due to a loss of power to the SLAVE relays for both trains.
Proposed Answer:
D. Safety Injection does not actuate due to a loss of power to the SLAVE relays for both trains.
. Explanation:
LB6B, SSPS Effects of ...
Consequence Loss of Vital Instrument AC bus #1 (#4) to output relays. PK02-18 (23) SSPS General Warning Train A (B) alarm.
Loss of power supply to Train A (B) slave relays. Automatic actuation of Train A (B) ESF equipment blocked.
Manual operation of Train A (B) equipment possible.
LPA4, Loss of Vital or Non
-Vital Instrument AC A. Incorrect. The loss of power affects the slave relays, which are powered from PY
-11 and 14. While instruments do lose power, if the slave relays have power, ESF will actuate if required. Plausible because could be thought it's the masters that lose power and that there is a different configuration of PY's to power the relays (ie I and II)
B. Incorr ect. The slave, not the master relays lose power, but its both trains that lose power. Plausible because could be thought that there is a different configuration of PY's to power the relays (ie I and II)
C. Incorrect. PY11 and 14 power the slave relays. Plausible because could be thought it's the masters that lose power.
D. Correct. Slave relays for both trains are de
-energized and while there is coincidence for SI DCPP L 1 61 Exam Rev 0 due to the loss of both channels, neither train will actuate
. Technical References
- LB6B , LPA4 , OIM B-6-1b References to be provided to applicants during exam: None Learning Objective
- Discuss abnormal conditions associated with Eagle 21 (37123)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
A-0156 X New Question History: Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 013 K5.02Knowledge of the operational implications of safety system logic and reliability as they apply to the ESFAS.
Tier # 2 Group # 1 K/A # 013 K5.02 Rating 2.9 Question 13 Which of the following describes the design feature of "Fail Safe" as it relates to the Sol id State Protection System?
A. Channels are physically and electrically separated.
B. If power is lost , system components will get a "GO" signal to actuate
. C. Capable of being calibrated at power without significant loss of protection.
D. Use of multiple channels of plant parameters sent to two identical and independent trains.
Proposed Answer:
B. If power is lost , system components will get a "GO" signal to actuate
. Explanation:
A. Incorrect because "Fail Safe" pertains to how a component fails such that a portion of the protection logic is made up. Plausible because separate channels could be interpreted as "fail safe". (design feature
- Independence)
B. Correct. If power to a channel of protection is lost, that channel fails in a manner that will cause actuation of that portion of the protection logic (Reactor Trip, Safety Injection, etc.)
(design feature
- fail-safe) C. Incorrect because it does not answer the question. Plausible because it is true, and it may be thought that since protection is still afforded while a calibration of a channel is performed, that it meets the meaning of fail safe. (design feature
- testability)
D. Incorrect because it does not answer the question. "Fail Safe" pertains to how a component fails such that a portion of the protection logic is made up. Plausible because multiple channels feeding two independent trains is a valid basic description of protection logic (design feature
- redundancy).
Technical References
- LB-6B Eagle 21 and Solid State Protection System, PG 6 References to be provided to applicants during exam: None Learning Objective
- Explain significant Eagle 21 and Solid State Protection System design features and the importance to nuclear safety. (41312)
Question Source:
Bank # DCPP bank S-54034 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.7 Difficulty:
3.0 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 022 A3.01
- Ability to monitor automatic operation of the CCS, including initiation of safeguards mode of operation.
Tier # 2 Group # 1 K/A # 022 A3.01 Rating 4.1 Question 14 GIVEN: Unit 1 is operating at 100% power D/G 1-2 cleared for maintenance The CFCU Speed Selector switch positions are as follows:
o CFCU 1 HIGH o CFCU 1 LOW o CFCU 1 LOW o CFCU 1 LOW o CFCU 1 LOW Offsite power is lost and Safety Injection occurs.
When sequencing is complete, what will be the status of the CFCUs?
A. CFCU 1-1 running in HIGH
- CFCU 1-2, and 1-4 running in LOW; CFCU 1-3 and 1-5 not running. B. CFCU 1-1, 1-2, 1-3, and 1-5 running in LOW; CFCU 1-4 not running
. C. CFCU 1-1 running in HIGH
- CFCU 1-2, 1-3, and 1-5 running in LOW; CFCU 1-4 not running. D. CFCU 1-1, 1-2, and 1-4 running in LOW; CFCU 1-3 and 1-5 not running
. Proposed Answer:
D CFCU 1-1, 1-2, and 1-4 running in LOW, CFCU 1
-3 and 1-5 not running. Explanation:
A. Incorrect because CFCUs will only start in LOW from timing relays. Plausible because the control switch is selected to HIGH, and CFCU 1
-3 and 1-5 would not be running (D/G 1-2 bus G). B. Incorrect because CFCU 1
-4 would be running, and CFCU 1-3 and 1-5 would be stopped due to loss of power (DG 1
-2, bus G).
Plausible because D/G - bus alignment is a difference between units. This cooler alignment would result on Unit 2 if D/G 2
-2 was cleared (powers bus H
- CFCU 2-4). C. Incorrect because CFCUs will only start in LOW from timing relays.
Plausible because CFCU 1-1 switch selected to HIGH, and the D/G
- bus alignment is a difference between units. This cooler alignment would result on Unit 2 if D/G 2
-2 was cleared (powers bus H
- CFCU 2-4). D. Correct. D/G 1-2 powers Bus G, therefore, CFCUs 3 and 5 are not powere d (CFCU 1-1, 1-2, 1-3, 1-4, 1-5 = buses F, F, G, H, G). CFCUs 1-1, 1-2 and 1-4 would start in LOW.
Technical References
- OIM J-6-1 References to be provided to applicants during exam: None Learning Objective
- Analyze automatic features and interlocks associated with the CFCUs. (6143)
DCPP L 1 61 Exam Rev 0 Question Source:
Bank # (note changes; attach parent)
Modified Bank #
A-0843 X New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.4 1.7 / 45.5 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 026 K1.01
- Knowledge of the physical connections and/or cause effect relationships between the CSS and ECCS.
Tier # 2 Group # 1 K/A # 026 K1.01 Rating 4.2 Question 15 Which of the following would prevent opening the residual heat removal (RHR) to containment spray rings isolation valve (9003A)?
A. RHR to containment spray rings isolation valve (9003B)
- OPEN. B. Containment recirculation sump suction valve (8982A)
- CLOSED. C. RHR cold leg injection isolation valve (8809A)
- OPEN. D. Containment spray pump discharge valve (9001A)
- CLOSED. Proposed Answer:
B. Containment recirculation sump suction valve (8982A)
- CLOSED. Explanation:
A. Incorrect because there is no interlock between the 9003 valves. Plausible because one train is aligned by the operators to supply sprays, and it may be thought that an interlock would be necessary to prevent both train s from supplying the spray header while on recirculation mode.
B. Correct. The associated 8982A/B must be open in order to open 9003A or B (this ensures the ECCS system is in the recirculation mode and not drawing suction from the RWST)
C. Incorrect because there is no interlock between 9003 and 8809 valves. Plausible because it may be thought that an interlock would be required to prevent one RHR pump from supplying both sprays and injection.
D. Incorrect because there is no interlock between 9003 and 9001 valves. Plausible because it may be thought that an interlock would be necessary to prevent a CS pump and RHR pump from supplying the spray header simultaneously.
Technical References
- LI2 Containment Spray System, PG 22 References to be provided to applicants during exam: None Learning Objective
- State the purpose of Containment Spray System components. (37576)
Question Source:
Bank # S-34412 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.2 to 41.9 / 45.7 to 45
.8 Difficulty: 2.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 039 A3.02
- Ability to monitor automatic operation of the Main and Reheat Steam System, including isolation of the MRSS.
Tier # 2 Group # 1 K/A # 039 A3.02 Rating 3.1 Question 16 Unit 2 is at full powe
- r. 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 valve s ONLY D. Reheat Stop and Intercept valves Proposed Answer:
B. Turbine Governor 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
- L C-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 NRC Exam #24 DCPP L111- 11/2012 X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.5 / 45.5 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 039 K4.06
- Knowledge of MRSS design feature(s) and/or interlock(s) which prevent reverse steam flow on steam line break. Tier # 2 Group # 1 K/A # 039 K4.06 Rating 3.3 Question 17 How does the design of the MSIVs limit the effect of a steam line break
? The MSIV has ____________ and will close when air is vented off by the DC solenoid (s) ___________.
A. a single check valve; energizing B. a single check valve; de-energizing C. two check valves; energizing D. two check valves; de-energizing Proposed Answer:
C. two check valves; energizing Explanation:
The MSIV is a reverse acting swing check type of valve. The disc is normally held open, out of the flow stream, by two air operated cylinders. If the disc rotates more than 6° into the flow stream, steam flow will close the valve against the air pressure in the operating cylinders. The check valve is also a swing check type of valve and is welded directly to the downstream side of the MSIV A. Incorrect because there is a swing check on the downstream side of the MSIV
. B. Incorrect because there is a swing check on the downstream side of the MSIV. The solenoid energizes to close the valve.
C. Correct. The MSIV has a double check valve arrangement and the solenoids are energize t o vent air, causing the MSIV to close.
D. Incorrect. The solenoids are energize to close. Plausible that its thought the "safe" design is de-energize to close.
Technical References
- LC-2a Main Steam System, page 14, 15 References to be provided to applicants during exam: None Learning Objective
- Describe Main Steam System components. (37594)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundament al X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 059 K3.02
- Knowledge of the effect that a loss or malfunction of the MFW will have on the AFW system.
Tier # 2 Group # 1 K/A # 059 K3.02 Rating 3.6 Question 18 GIVEN: Unit 1 is operating at 3 5% power The, Main Feed Water Regulating valve for S/G 1
-4, fail s closed Steam Generator 1
-4 narrow range level is 1 5% and lowering at 1% per minute What is the expected response of the AFW system and the reactor
? A. After a time delay, 1-2 and 1-3 MDAFW pumps start and the reactor trips. The TDAFW will not be started
. B. 1-2 and 1-3 MDAFW pumps start immediately and the reactor trips after a time delay. The TDAFW will not be started.
C. After a time delay, the reactor trips and all AFW pumps start.
D. All AFW pumps start immediately and the reactor trips after a time delay.
Proposed Answer:
A. After a time delay, 1
-2 and 1-3 MDAFW pumps start and the reactor trips. The TDAFW will not be started.
Explanation:
A. Correct. There is a time delay for BOTH the AFW pump start and reactor trip
. At this power level, shrink will be insufficient to start the TDAFW pump.
B. Incorrect The time delay is for both the reactor trip and the AFW pump start
. Plausible if its thought there is an immediate need for the AFW pumps to start.
C. Incorrect Only the motor driven pumps are started for one SG. Plausible, there is a time delay for both the trip and pump start and it could be thought the TDAFW pump will start or would be started due to the shrink from the trip (will not occur at this power level).
D. Incorrect. As stated previously, there is a time delay for both and the TDAFW pump will not be started.
Technical References
- OIM B-6-4b, D-1-2 References to be provided to applicants during exam: None Learning Objective
- Analyze automatic features and interlocks associated with the Auxiliary Feed Water System. (37637)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.3
DCPP L161 Exam Rev 0 Examination Outline Cross-Reference Level RO 061 K6.02 - Knowledge of the effect of a loss or malfunction of Pumps will have on the AFW components.
Tier # 2 Group # 1 K/A # 061 K6.02 Rating 2.6 Question 19 GIVEN: Unit 2 is operating at 100% power when a Reactor Trip occurs. One minute after AFW pump 2-3 st arts, it trips on overcurrent. The crew is preparing to cross-tie to feed the 2-3 and 2-4 Steam Generators.
What is the position of LCV-115 and LCV-113 and what action should be taken by the operator to regain control of the valves?
- 1. LCV-115 and LCV-113, AFW supply valves to Steam Generators 2-3 and 2-4 have failed
_________
- 2. To restore control of the valves, the MINIMUM action for the operator to take is to
________________.
A. 1. Open 2. operate the 2-3 AFW pump run interlock bypass switch B. 1. Closed 2. place the controller in MANUAL C. 1. Closed 2. operate the 2-3 AFW pump run interlock bypass switch D. 1. Open 2. place the controller in MANUAL Proposed Answer: A. 1. Open 2. operate the 2-3 AFW pump run in terlock bypass switch Explanation: A. Correct. Valves will fail open. The interlock bypass switches will allow closing the valves without the pump running.
B. Incorrect. The breaker opening will take away control power from the LCVs, causing them to fail open. Plausible because it may be thought that the LCVs can be controlled manually once the AFW train is actuated. However, manual control is not available from the controller until the interlock bypass sw itches are operated.
C. Incorrect. The valves fail open. Plausible because the interlock bypass switch will allow closing the valves without the pump running.
D. Incorrect. Manual control is not available from the controller until the interlock bypass switches are operated. Plausible because the valves will fail open.
Technical References
- LD-1 References to be provided to applicants during exam:
None Learning Objective
- Describe the operation of the AFW system (8402) Question Source: Bank # (note changes; attach parent) Modified Bank # L121 Question 19 X DCPP L 1 61 Exam Rev 0 New Question History:
Last Two NRC Exam s (8 August 2014)
Yes Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 / 45.7 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 062 A1.01
- Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including the significance of D/G load limits. Tier # 2 Group # 1 K/A # 062 A1.01 Rating 3.4 Question 20 GIVEN: Unit 1 has lost all offsite power DG 1-3 cannot be started DG 1-1 has been cross-tied to allow powering Unit 1 Vital 4kV bus F per ECA-0.3, Restore 4 KV Buses The DG is currently loaded to 19 8 5 kW Per ECA-0.3, which of the following is the maximum load that can be added to the two Vital buses connected to DG 1
-1 without exceeding its continuous (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) rating? Load KW Demand 1 Centrifugal Charging Pump 515 2 Safety Injection Pump 312 3 Fire Pump 147 4 Containment Fan Cooler 85 A. Centrifugal Charging Pump and Containment Fan Cooler B. Safety Injection Pump and Containment Fan Cooler C. Centrifugal Charging Pump and Safety Injection pump D. Safety Injection Pump and Fire Pump Proposed Answer:
A. Centrifugal Charging Pump and Containment Fan Cooler Explanation:
A. Correct. DG continuous load rating is 2600 kW. If the CCP and CFCU were added to the existing load, the total would be 19 85+515+85=25 85 kW. This is the maximum within the DG continuous load rating of the answer choices.
B. Incorrect. If the SI pump and CFCU were added to the existing load, the total load would be 19 85+312+85=23 82 kW. Plausible because it is within the DG's continuous rating. C. Incorrect. If the CCP and fire pump were added to the existing load, the total load would be 19 85+515+312 =2 812 kW which exceeds the continuous rating but within the DG max rating of 2 8 60 kW. Plausible because it is also within the DG within the 30 min rating of 3056 kW. D. Incorrect. If the SI pump and CFCU were added to the existing load, the total load would be 19 85+312+147=24 44 kW. Plausible because it is within the DG's continuous rating. Technical References
References to be provided to applicants during exam: None Learning Objective
- Describe the operation of the Diesel Generator System. (6437) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
DCPP L 1 61 Exam Rev 0 New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.5 / 45.5 Difficulty: 3.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 062 A4,03
- Ability to manually operate and/or monitor synchro scope, including an understanding of running and incoming voltages in the control room.
Tier # 2 Group # 1 K/A # 062 A4.03 Rating 2.8 Question 21 Diesel Generator 1-3 is being synchronized to bus F for testing. The synchroscope for the Diesel Generator breaker is on.
The following indications exist:
Bus voltage is ____________ than Diesel Generator voltage; turn Voltage Adjust to _______
__. A. higher; lower B. lower; lower C. higher; raise D. lower; raise Proposed Answer:
B. low er; lower Explanation:
NOTE: Depending on how the meters are interpreted, "Sync incoming" could be thought as bus voltage. In this case, diesel voltage is HIGHER than bus voltage. The Voltage Adjust raises/lowers diesel (not bus voltage) and should be lowered A. Incorrect. Bus voltage is lower than DG voltage (incorrect). The Voltage Adjust control should be moved to "lower" to lower DG voltage to bus voltage (correct) B. Correct Running voltage is the bus voltage. The DG is "incoming" and is higher than bus voltage ( or bus voltage less than incoming). The DG Voltage Adjust must be turned to "lower" to match DG voltage to bus voltage.
C. Incorrect. Bus voltage is lower than DG voltage (incorrect). The Voltage Adjust control should be moved to "lower" to lower DG voltage to bus voltage (not raise
- incorrect)
D. Incorrect Bus voltage is lower than DG voltage (correct). The Voltage Adjust control should be moved to "lower" to lower DG voltage to bus voltage (not raise
- incorrect)
Technical References
- STG J-6b Diesel Generator System page 2.7-21 References to be provided to applicants during exam: None Learning Objective
- List the three conditions required for paralleling AC sources. (65985), Describe controls, indications, and alarms associated with the Diesel Generator System. (37724) Question Source:
Bank #
DCPP L 1 61 Exam Rev 0 (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 2.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 063 G2.4.6
- D.C. Electrical Distribution: Knowledge of EOP mitigation strategies.
Tier # 2 Group # 1 K/A # 063 G2.4.6 Rating 3.7 Question 22 The crew is performing ECA-0.0, "Loss of All Vital AC Power," Appendix DC, "Shed Non-Essential DC loads." An operator has been sent to secure the Main Turbine DC Bearing oil pump and the MFP Emergency DC Oil pumps. These loads are secured in order to prevent losing:
A. DC breaker control power.
B. DC back up power supply to Vital UPS.
C. DC power to solenoids for air operated valves.
D. DC power for emergency lighting.
Proposed Answer:
D. DC power for emergency lighting.
Explanation:
A. Incorrect.
Plausible because breaker control power is important to plant control. However, DC lighting is the basis for load shed on non
-Vital DC. B. Incorrect. Plausible because it could be thought that as an emergency backup, a non
-Vital source could be provided to supply a Vital UPS. C. Incorrect. Plausible because solenoid valves are important to plant control in that they provide remote control for system valves.
D. Correct. If 250 VDC loads are not reduced on non
-Vital DC, emergency lighting could be lost in 90 minutes.
Technical References
-Essential DC Loads" page 1 of 1.
References to be provided to applicants during exam: None Learning Objective
Question Sou rce: Bank # P-45056 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.10 / 43.5 / 45.13 Difficulty: 3.
5 U.S. Nuclear Regulatory Commission Diablo Canyon RO Written Examination Applicant InformationName: KEY Date: 14 October, 2016 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 ow
- n. I have neither gi ven nor received aid.
______________________________________
Applicant's Signature Results Examination Value __________ Points Applicant's Score __________ Points
Applicant's Grade __________ Percent
U.S. Nuclear Regulatory Commission Diablo Canyon SRO Written Examination Applicant InformationName: KEY Date: 14 October, 2016 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 fi nal grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items. Examination papers will be collected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the examination begins.
Applicant Certification All work done on this examination is my ow
- n. I have neither gi ven nor received aid.
______________________________________
Applicant's Signature ResultsRO/SRO-Only/Total Examination Values
/ / Points Applicant's Scores / / Points Applicant's Grade / / Percent
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 064 K1.02
- Knowledge of the physical connections and/or cause effect relationships between the ED/G system and the D/G cooling water system.
Tier # 2 Group # 1 K/A # 064 K1.02 Rating 3.1 Question 23 Which of the following will ONLY trip an emergency diesel generator when it is in "LOCAL"?
A. Directional power relay
. B. Jacket water high temperature
. C. Low lube oil pressure
. D. Crankcase overpressure
. Proposed Answer:
B. Jacket Water High temperature
. Explanation: A. Incorrect. Directional power is normally not active but due to not being cut in
. B. Correct. The High JW Temp Trip is bypassed unless the D/G is in LOCAL control
. C. Incorrect Low lube oil pressure will trip the diesel in LOCAL or REMOTE
. D. Incorrect. Over cranking is a LOCAL only trip and the crankcase pressure does take action to relieve high pressure but it does not trip the diesel
. Technical References
- LJ-6B. References to be provided to applicants during exam: None Learning Objective
- Analyze automatic features and interlocks associated with the Diesel Generator System. (37725) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.2 to 41.9 / 45.7 to 45.8 Difficulty: 3.0
DCPP L161 Exam Rev 0 Examination Outline Cross-Reference Level RO 073 A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply.
Tier # 2 Group # 1 K/A # 073 A2.01 Rating 2.5 Question 24 A liquid rad waste discharge is in progress per OP G-1:II, Liquid Radwaste System - Discharge of Liquid Radwaste. RE-18, Liquid Radwaste, radiation monitor power supply fails.
Which of the following describes the auto action(s) that occur a nd action necessary to recommence the discharge?
Auto Action(s): 1. RCV-18, LRW Overboard Isol ation valve, closes, only 2. RCV-18, LRW Overboard Isolatio n valve closes and FCV-477, Discharge Recirc valve, opens Discharge Action: 3. A temporary radiation monitoring device is installed to take the action(s) of RE-18 if required 4.Two independent samples are analyzed a nd release rate calc ulations verified A. 1, 3 B. 2, 3 C. 1, 4 D. 2, 4 Proposed Answer: D. 2, 4 Explanation:
A. Incorrect. In addition to the discharge terminating with RCV-18 closing, FCV-477 opens. A comp measure is not necessa ry but a plausible fix to replace the failed monitor.
B. Incorrect. Actions are correct, however, per procedure, the discharge may be restarted if actions are taken - 2 independent samples taken, analyzed and release rate calculations checked. C. Incorrect. FCV-477 also opens.
D. Correct. Both actions occur and the discharge may be done if actions are taken to ensure the discharge will be below activity limits.
Technical References
- OP G-1:II References to be provided to applicants during exam:
None Learning Objective: Discuss abnormal conditions associated with the Radiation Monitoring System. (37878) Question Source: Bank # (note changes; attach parent) Modified Bank # 24 DCPP NRC Exam L141 (04/16) X New DCPP L 1 61 Exam Rev 0 Question History:
Last Two NRC Exam s #24 DCPP NRC L141, 04/2016 Yes Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.11 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 076 K2.01
- Knowledge of bus power supplies to Service Water.
(ASW) Tier # 2 Group # 1 K/A # 076 K2.0 1 Rating 2.7 Question 25 The power supply for ASW pump 1
-1 is bus _____, and for ASW pump 1
-2 is bu s _____. A. F; H B. G; H C. G; F D. F; G Proposed Answer:
D. F; G Explanation:
A. Incorrect. Pump 1-1 is train A powered from bus F. Pump 1
-2 is train B powered from bus G. Distractors have pump 1 powered from either F or G for plausibility, as no #1 pumps are powered from H.
B. Incorrect.Pump 1-1 is train A powered from bus F. Pump 1
-2 is train B powered from bus G. Distractors have pump 1 powered from either F or G for plausibility, as no #1 pumps are powered from H.
C. Incorrect.Pump 1-1 is train A powered from bus F. Pump 1
-2 is train B powered from bus G. Distractors have pump 1 powered from either F or G for plausibility, as no #1 pumps are powered from H.
D. Correct. Pump 1-1 is train A powered from bus F. Pump 1
-2 is train B powered from bus G. Distractors have pump 1 powered from either F or G for plausibility, as no #1 pumps are powered from H.
Technical References
- OIM J-1-1 8 References to be provided to applicants during exam: None Learning Objective
- State the power supplies to ASW System components. (53
- 39) Question Source:
Bank # P-2532 5 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.7 Difficulty:
2.0 DCPP L 1 61 Exa m Rev 0 Examination Outline Cross
-Reference Level RO 078 A4.01
- Ability to manually operate and/or monitor pressure gauges in the control room.
Tier # 2 Group # 1 K/A # 07 8 A4.01 Rating 3.1 Question 26 Both units are at full power.
Instrument air pressure on PI-380 on VB4 is approximately 105 psig. A loss of all offsite power occurs on Unit 1.
Unit 2 is not affected.
10 minutes later, which of the following would be an expected air pressure indication on PI
-380? A. Stable at approximately 105 psig B. Stable at approximately 9 5 psig C. 90 psig and lowering slowly D. 60 psig and lowering rapidly Proposed Answer:
A, Stable at approximately 105 psig Explanation:
A. Correct. Because both units IA systems are cross tied, Unit 2 rotary compressor will supply IA to unit 1. Air pressure should be normal.
Rotary AC 0
-5 is powered from Unit 2 (06 and 07 are powered from Unit 1)
B. Incorrect. This would be normal if its thought that pressure will fall until the standby air compressors start (approximately 93 to 100 psig)
C. Incorrect. Possible if the systems were not cross tied. Pressure, because of low usage, would fall slowly.
D. Incorrect. If the units were not cross
-tied, its possible to think pressure would fall rapidly, especially, since there are accumulators on some IA operated valves to aid in maintaining operation.
. Technical References
- LK-1, STG K-1 References to be provided to applicants during exam: None Learning Objective
- State the normal indications associated with the Compressed Air System. (40972) Question Source:
Ba nk # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 103 A2.03
- 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
- Phase A and B isolation.
Tier # 2 Group # 1 K/A # 103 A2.03 Rating 3.5 Question 27 GIVEN: Unit 1 was at 100% power A large break LOCA occurred RCS pressure is 40 psig Containment pressure is 25 psig and rising AFW flow is 0 gpm Narrow range levels in all steam generators are 3 8% and lowering slowly The crew has entered E
-1, Loss of Reactor or Secondary Coolant For the current plant conditions, what action should be taken by the crew and what is the basi s for the action?
A. Trip the RCPs to prevent severe core uncovery if the RCPs trip later in the event.
B. Go to FR-H.1, Response to Loss of Secondary Heat Sink to attempt to restore AFW flow
. C. Trip the RCPs to prevent damaging the RCPs due to a loss of cooling to the motor bearing oil coolers. D. Go to FR-H.1, Response to Loss of Secondary Heat Sink to ensure a heat sink is not required for the current plant conditions
. Proposed Answer:
C. Trip the RCPs to prevent damaging the RCPs due to a loss of cooling to the motor bearing oil coolers.
Explanation:
Applicability: In the scenario of this question, Phase A isolation occurs on SI, and Phase B isolation occurs at 22 psig containment pressure (malfunction on containment system). Phase B isolation closes containment isolation valves associated with RCP cooling (also an "operation" of a containment system). These valves are listed in procedure STP V
-11 "Containment Isolation Phase B Valves FCV
-355, FCV-356, FCV-357, FCV-363, FCV-749, and FCV
-750" as meeting Technical Specification 3.6.3 (Containment Isolation Valves) SR 3.6.3.8. As such, they are included in the "Containment" Technical Specification family and meet the K/A system designator for Containment (#103).
A. Incorrect because the RCPs are being secured due to losing cooling. Plausible because t h e severe core uncovery concern is related to small break LOCA response. B. Incorrect. Entry conditions not met for RED path FR
-H.1. Plausible as there is no AFW flow the adverse containment number for adequate SG level is higher (25% vs. 15%).
C. Correct. Loss of RCP motor cooling occurs when Phase B occurs (22 psig). RCPs are stopped to prevent damage to the motors due to overheating.
D. Incorrect. Entry conditions not met for RED path FR
-H.1. Plausible as there is no AFW flow the adverse containment number for adequate SG level is higher (25% vs. 15%) and for a LOCA, the proper response is to check RCS and SG pressures and exit the procedure if RCS pressure is below SG pressure (heat sink not required)
DCPP L 1 61 Exam Rev 0 Technical References
- E-1 "Reactor Trip or Safety Injection
", Foldout Page
-(various)", F-0 References to be provided to applicants during exam: None Learning Objective
- Describe system interrelationships between the RCP and other plant systems. (35742)
Question Source:
Bank # L081 RO Exam Question 41 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.5 / 43.5 / 45.3 / 45.13 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 103 A2.05
- 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: Emergency Containment entry.
Tier # 2 Group # 1 K/A # 103 A2.05 Rating 2.9 Question 28 GIVEN: PK10-10 "Fire Detected" has alarmed Temperatures are slowly rising in the Unit 1 Containment A fire in containment has been confirmed An Alert has been declared The Fire Brigade is about to make an emergency containment entry to initiate combatting the fire and an emergency exposure for protecting/saving plant equipment is to be authorized.
Per EP RB-2, Emergency Exposure Guidelines, what emergency exposure limit would apply
? (Assume there is no immediate threat to compromising plant safety.)
A. 5 rem B. 10 rem C. 15 rem D. 25 rem Proposed Answer:
B. 10 rem Explanation:
For a fire in containment, an emergency entry would be made (first part of the KA). However, that does not mean there are no exposure limits. Procedure EP RB
-2 (the second part of the A2 KA), outlines what exposures can be authorized for different events.
NOTE: SRO JPM not a conflict as the JPM is to determine the appropriate exposure based on conditions, none of which are a fire.
A. I ncorrect. 5 rem is the limit for exposures such as sampling B. Correct. Per EP RB
-2, "Property Saving, for example, might be dispatching the Fire Brigade to extinguish a fire in a Very High Radiation Area to protect plant equipment though no immediate threat exists to compromising Plant Safety" C. Incorrect. This is a limit for lens of the eye for the sampling limit
. D. Incorrect. This is the limit for dose saving to population (activities that justify a potential overexposure to a few workers in order to save even a small average dose in a large population) or lifesaving to individual.
Technical References
- EP RB-2 References to be provided to applicants during exam: None Learning Objective: State the emergency dose limits.
(7954) Question Source:
Bank # 72 DCPP NRC exam 11/2012 X (note changes; attach parent)
Modified Bank #
New Previous NRC Exam Yes Question History:
Last Two NRC Exam s No DCPP L 1 61 Exam Rev 0 Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.12 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 002 K6.06
- Knowledge of the effect or a loss or malfunction on (of) the following RCS components:
sensors or detectors Tier # 2 Group # 2 K/A # 002 K6.06 Rating 2.5 Question 29 Unit 1 reactor power is 15% and the crew is shutting down the unit to MODE 5
. PT-403A, Loop 4 Hot Leg Wide Range RCS Pressure, fails HIGH. What is the effect of this failure on the plant? A. Pressurizer PORV-456 opens B. RVLIS Dynamic Range will indicate higher than actual C. Subcooled Margin Monitor on VB2 will indicate higher than actual D. RHR-8702 RCS Loop 4 hot leg to RHR loop isolation, will not be able to be opened.
Proposed Answer:
D. RHR-8702 RCS Loop 4 hot leg to RHR loop isolation will not be able to be opened.
Explanation: A. Incorrect because Pressurizer NR Pressure is used to open PORVs for overpressure protection during normal operation. Plausible because PT
-403A will cause the PORV to open when LTOP is armed (PCV-455C). B. Incorrect because PT
-403A does not feed RVLIS. Plausible because PT
-403 does feed RVLIS and if failed, would cause the indication to be erroneous.
C. Incorrect because PT
-403A does not feed the SC CM Monitor. Plausible because PT
-403 does feed the S CCM Monitor and if failed would cause the indication to be erroneous.
D. Correct. PT
-403A provides an interlock to prevent opening the RHR loop isolation valve unless RCS pressure is below 390 psig to protect the RHR loop.
Technical References
- OIM A-1-2, A-2-2, A-4-7 References to be provided to applicants during exam: None Learning Objective
- Describe controls, indications, and alarms associated with the RCS. (35017) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 / 45.7 Difficulty: 3.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 015 K2.01
- Knowledge of bus power supplies to NIS channels, components, and interconnections.
Tier # 2 Group # 2 K/A # 015 K2.01 Rating 3.3 Question 30 Unit 1 is operating at 6% power Panel PY-1 3 is d e-energized. Which of the following excore instrumentation channels have lost power, and whether or not a n automatic reactor trip will occur? A. A power range only, no reactor trip will occur B. An intermediate range and a power range
- no reactor trip will occ ur C. A power range only, reactor trip will occur D. An intermediate range and a power range; reactor trip will occur Proposed Answer
A. A power range only
- no reactor trip will occur Explanation
A. Correct. PY-1 3 powers the 3 r d column of the NIS cabinet in the control room, with PR N4 3. Reactor power is below P10, and 25% (IR and PR trip setpoint), however, the PR logic is 2 of 4 and no IR has been lost
. B. Incorrect because the IR channels are powered from PY
-11 and 12. Plausible because it could be the IR channels are not all on the same channels as the SR channels. If an IR lost power, the reactor would trip.
C. Incorrect because there is no reactor trip. Plausible as only a PR is lost.
D. Incorrect because there is no reactor trip and only a PR is lost. Plausible as the power supplies must be known and if IR had lost power, this would be the correct answer.
Technical References
- LB4 OIM B-6-4a References to be provided to applicants during exam: None Learning Objective
- State the power supplies to Excore Nuclear Instrumentation System components. (40940)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
S-32224 X New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 2.5
DCPP L161 Exam Rev 0 Examination Outline Cross-Reference Level RO 016 K3.01 - Knowledge of the effect that a loss or malfunction of the NNIS will have on the RCS.
Tier # 2 Group # 2 K/A # 016 K3.01 Rating 3.4 Question 31 Unit 1 is operating at 75% reac tor power, steady state.
TE-130, the temperature element input to Letdown Temperature Control Valve TCV-130, fails high.
With no operator action, what will be the effect of this failure?
A. TCV-130 will close, RCS Tavg will lower.
B. TCV-130 will open, RCS Tavg will lower.
C. TCV-130 will open, RCS Tavg will rise.
D. TCV-130 will close, RCS Tavg will rise.
Proposed Answer: C. TCV-130 will open, RCS Tavg will rise.
Explanation:
A. Incorrect because TCV-130 will open and Tavg will rise. Plausible because it may be thought that the response of TCV-130 to high temperature is reversed. If this was the case, the listed effect on Ta vg would be correct.
B. Incorrect because Tavg will rise. Plausible because it may be thought the effect letdown temperature has on boron removal by the demineralizer bed is reversed.
C. Correct. If indicated letdown temperature rises, TCV-130 will open to raise the CCW flow through the letdown heat exchanger. Letdown temperature will lower. The letdown demineralizers will remove more boron from the letdown flow, lowering the RCS boron concentration over time. The positive reactivity effect will cause reactor power to rise incrementally, raising RCS Tavg.
D. Incorrect because TCV-130 will open. Plausible because the effect on RCS Tavg is correct, but it may be thought that the response of TCV-130 to high temperature is reversed.
Technical References
- .OIM B-1-1, LPA-33 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 Two NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.41.7 Difficulty: 3.0 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 033 A1.01
- Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including:
Spent fuel pool water level Tier # 2 Group # 2 K/A # 033 A1.01 Rating 2.7 Question 32 Unit 1 is at full power. 1
-2 Spent Fuel Pool (SFP) pump is in service.
Safety Injection occurs.
4 kV vital buses transfer to startup.
Which of the following describes the status of spent fuel pool cooling?
SFP temperature will be _______________ because ________________.
A. rising; while there is a SFP pump running, there is no cooling aligned to the SFP heat exchanger B. rising; while there is cooling aligned to the SFP heat exchanger, there is no SFP pump running C. stable; the 1
-1 SFP pump is running with cooling aligned to the SFP heat exchanger D. stable; the 1
-2 SFP pump running with cooling aligned to the SFP heat exchanger Proposed Answer:
B. rising; while there is cooling aligned to the SFP heat exchanger, there is no SFP pump running Explanation:
A. Incorrect. There would be cooling to the heat exchanger
- phase A does not close the non
-vital CCW header, however, the 1
-2 pump is stopped by the phase A (and does not restart)
B. Correct. On phase A, the 1-2 pump is de
-energized and does not restart. Loss of flow will cause SFP temperature to rise.
C. Incorrect. There would be cooling, however, there is no swap or auto start of the 1
-1 pump when the 1
-2 pump is tripped (as for other pumps, such as CCW). If the 1
-1 pump had been the inservice pump, this would have been correct.
D. Incorrect.
The 1-1 pump would have kept running, it is the 1
-2 pump that is tripped, it will not be running.
Technical References
- OIM B-6-7, LB-7 References to be provided to applicants during exam: None Learning Objective
- Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.8 Difficulty: 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 034 A4.02
- (FHES) Ability to manually operate and/or monitor neutron levels in the control room.
Tier # 2 Group # 2 K/A # 034 A4.02 Rating 3.5 Question 33 GIVEN: Unit 1 core reloading is in progress Initial Source Range N31 counts - 1 6 cps Initial Source Range N32 counts
- 2 0 cps N31 and N32 counts begin to increase.
W hich of the following would be the FIRST level of source range counts at which the crew will have to stop core reload in accordance with OP B-8DS2, Core Loading
? N31 N32 A. 32 cps, 36 cps B. 36 cps, 40 cps C. 48 cps, 52 cps D. 56 cps, 60 cps Proposed Answer:
B. 36 cps, 40 cps Explanation:
OP B-8DS2 lists several criteria for suspending fuel movement. For NIS channels the two criteria are: ALL responding channels increase by a factor of 2, or ONE responding channel increases by a factor of 3.
N31 increase
- double - 32 cps , triple - 48 cps N32 increase
-double - 40 cp s, triple - 60 cps A. Incorrect because N3 1 double, but N3 2 did not. B. Correct because both have doubled
. C. Incorrect. N3 1 tripled - not the first time fuel handling must be stopped D. Incorrect because, although both channels tripled, it is not the FIRST that would cause fuel movement to be suspended
. Technical References
- OP B-8DS2, Core Loading, LB-4, Excore Nuclear Instrumentation System, Pg 57.
References to be provided to applicants during exam: None Learning Objective
- Describe the operation of the Excore Nuclear Instrumentation System. (36972) Question Source:
Bank (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.10 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 035 A3.01
- Ability to monitor automatic operation of the S/G including S/G water level control.
Tier # 2 Group # 2 K/A # 035 A3.01 Rating 4.0 Question 34 Given: Unit 2 is at 100% power PK14-19, STATO R WTR CLG SYSTEM, alarms Turbine runback is indicated on the Triconix display and Main Generator MW are decreasing with no reactor trip What will the effect be on steam generator narrow range water levels and feed flow during the ramp? Narrow Range levels w ill initially __________ and feed flow
___________
steam flow
. A. rise; will track B. r ise, initially rises but quickly lowers and will be less than C. lower; will track D. lower; initially lowers but quickly rises and will be greater than Proposed Answer:
D. lower; initially lowers but quickly rises and will be greater than Explanation:
A. incorrect. Levels will lower due to "shrink" An increase in steam flow would cause levels to rise. Plausible, as at power both are in agreement.
B. Incorrect. Levels will lower.
Plausible - The feed flow response would be correct for the dominant level signal if level increased
. C. Incorrect. Levels will lower. Plausible if its thought that feed flow will follow steam flow, which is lowering, this is correct
. D. Correct. Levels will lower due to lowering steam flow. The feed system will respond by increasing flow to restore levels to program.
Technical References
- L C-8B , LPA-25, sim response to stator runback References to be provided to applicants during exam: None Learning Objective
- Describe controls, indications, and alarms associated with the DFWCS. (37641) Question Source:
Bank # S-53384 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 / 45.5 Difficulty: 3.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 071 G2.1.30
- Waste Gas Disposal: Ability to locate and operate components, including local controls.
Tier # 2 Group # 2 K/A # 071 G2.1.30 Rating 4.4 Question 35 GIVEN: Gas Decay Tank (GDT) 1-1 is selected for "Purge" with pressure at 45 psig GDT 1-2 is selected to "standby" with pressure at 10 psig GDT 1-3 is "fill" with pressure at 74 psig The operator selects "OPEN" for the 1
-1 GDT Vent valve.
Which of the following describes the location of the control for the 1
-1 GDT Vent valve and the expected response when "OPEN" is selected.
The 1-1 GDT Vent valve is operated from the ________________.
If the operator selects "OPEN" for the 1
-1 GDT Vent valve, the 1
-1 GDT Purge valve will
______________________.
A. Control Room close immediately B. Aux Board close immediately C. Control Room remain open until GDT 1-1 pressure is < 15 psig D. Aux Board remain open until GDT 1-1 pressure is < 15 psig Proposed Answer:
B. Aux Board close immediately Explanation:
A. Incorrec t because the vent valve is operated from the HMI at the Aux Board in the Aux Building. Plausible because the Purge Valve will immediately close and there many HMI controls (such as ventilation panels POV1 and 2, Main Turbine) in the control room
. B. Correc t. FCV-407, FCV-408, and FCV
-409 automatically close or are prevented from opening when any one of the following occur:
- 1) pressure in the GDT being purged drops below 15 psig. 2) the vent valve on the tank being purged is opened, or 3) the tank being purged is also in standby and automatically transfers to the fill mode.
C. Incorrect because the Vent Valve is operated from the HMI at the Aux Board in the Aux Building. Plausible because there are controls for various systems on HMI screens in back of the control boards, and plausible because there is an interlock that will also close the Purge Valve when GDT pressure goes below 15 psig.
D. Incorrect because the Purge Valve will immediately close. Plausible because the Vent Valve is operated from Aux Board, and because there is an interlock that will also close the Purge Valve when GDT pressure goes below 15 psig.
DCPP L 1 61 Exam Rev 0 Technical References
- LG-2 Gaseous Radwaste Systems, page 24, 19 References to be provided to applicants during exam: None Learning Objective
- Identify the location of components associated with the Gaseous Radwaste System. (40935), Describe controls, indications, and alarms associated with the Gaseous Radwaste System. (37706)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.5 Difficulty: 4.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 072 K4.01
- Knowledge of ARM system design feature(s) and/or interlock(s) which provide for containment ventilation isolation.
Tier # 2 Group # 2 K/A # 072 K4.01 Rating 3.3 Question 36 GIVEN: Containment Purge is in progress.
The following sequence of events occur:
o AR PK02-06, CONTMT VENT ISOLATION, alarms o It is determined that the alarm was due to Containment Purge Radiation Monitor, RE
-44A, failing high The operator presses the RESET pushbuttons on VB1. The signal resets:
A. and is now available to actuate again if RE
-44B detects high radiation.
B. however, a subsequent high radiation condition will not cause isolation to occur.
C. but immediately occurs again once the operator releases the reset pushbuttons
. D. however, the containment purge cannot be re
-established until the signal is cleared.
Proposed Answer:
B. however, a subsequent high radiation condition will not cause isolation to occur. Explanation:
A. Incorrect. Auto CVI is blocked.
Plausible because signals that do not have Retentive latch, such as FWI or P
-11 will occur again after reset.
B. Correct. Auto CVI from RE
-44B is blocked. 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
. C. 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.
Plausible because if the signal reset was the the "latch" or seal in type, this would be the response.
D. Incorrect. With the signal reset, purge can be re
-established.
Plausible if its thought the signal must be clear to operate the valves.
Technical References
- OIM B-6-9a. References to be provided to applicants during exam: None Learning Objective
- 5119 - Analyze automatic features and interlocks associated with the Containment Purge System.
Question Source:
Bank # L121 Question 34 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s Yes Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.3 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 079 K1.01
- Knowledge of the physical connections and/or cause effect relationships between the SAS and IAS.
Tier # 2 Group # 2 K/A # 079 K1.01 Rating 3.0 Question 37 When the Service Air Compressors are required to supply Instrument Air (IA), the preferred method is to use the:
___________valve which connects to IA ____________ of the IA after filters. A. Automatic; upstream B. Manual; downstream C. Automatic; downstream D. Manual; upstream Proposed Answer:
D. Manual; upstream Explanation:
Two crossties to Service Air (SA), not normally used.
Manual crosstie upstream of the after filters, normally closed, check valve only allows SA to supply Instrument Air.
Automatic cross tie (PCV
-114), normally isolated, allows Instrument Air to supply SA.
Starts to close if Instrument Air header pressure decreases to 95 psig, full closed at 90 psig.
Auto closes if all Compressors 0
-1 through 0
-6 are off.
Auto close feature defeated by SA Block Bypass switch A. Incorrect - not an automatic cross tie (automatic, service air refusal valve used to go from IA to SA). B. Incorrect
- connection is upstream of the filters
. Plausible
- the automatic valve is downstream.
C. Incorrect
- Plausible, the automatic cross tie, connection is upstream (automatic, service air refusal valve used to go from IA to SA)
D. Correct. The manual valve ties in upstream of the filters to ensure filtered SA is supplying IA, Technical References
- STG K1 Compressed Air, page 2.1
-2; OIM K-1-1; AP-9 References to be provided to applicants during exam: None Learning Objective
- Given an abnormal condition, summarize the major actions of OP AP-9 to mitigate an event in progress. (3477J)
Question Source:
Bank # NRC Exam 11/9/2009, Q# 38 X (note changes; attach parent) Modified Bank #
New Past NRC Exam Yes Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO 086 A2.04
- Ability to (a) predict the impacts of failure to actuate the FPS when required, resulting in fire damage on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of failure to actuate the FPS when required, resulting in fire damage: Tier # 2 Group # 2 K/A # 086 A2.04 Rating 3.3 Question 38 Given: Circulating Water Pump (CWP) 1-2 has tripped on overcurrent A Security Officer has call ed the Control Room and report ed that smoke is coming from the CWP 1-2 motor The CO 2 fire protection system failed to actuate automatically
. The operator can actuate CO 2 locally for CWP 1-2 by: A. Pulling the pin from the small CO 2 bottle at the west side of CWP 1
-2 motor and squeezing the handle to release CO
- 2. B. Pulling down the cover plate on the CO 2 actuation station on the wall near CWP 1
-2 and pressing the red button inside to release CO
- 2. C. Verifying the red light by the hose reel is on, pulling the nozzle from its holder, pointing the nozzle at the flames, then squeezing the handle to release CO
- 2. D. Ensure the abort valve is open, break the glass and open the pilot valve to release CO
- 2. Proposed Answer:
A. Pulling the pin from the small C O 2 bottle at the west side of CWP 1
-2 motor and squeezing the handle to release CO
- 2. Explanation:
A. Correct. Local manual actuation is accomplished by pulling the pin and squeezing the handle on the small local CO 2 bottle at each CWP.
This will operate a larger CO 2 bottle valve and release CO 2 from the bank
. B. Incorrect method. Plausible because this is similar to other fire protection actuation mechanisms.
C. Incorrect method. Plausible because this is similar to hose reel stations elsewhere on site.
D. Incorrect method. Plausible this is the method for actuating low pressure Cardox, for areas such as the Cable Spreading Room
. Technical References
- OP K2B:II, Low Pressure Cardox
- Manual Use Of The Cardox System, OP K-2E:IV High Pressure Carbon Dioxide Manual Actuation References to be provided to applicants during exam: None Learning Objective
- Describe the operation of the Cardox System. (3707)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 DCPP L 1 61 Exam Rev 0 Difficulty: 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO EPE 007 EK1.05 Knowledge of the operational implications of decay power as a function of time as they apply to the reactor trip. Tier # 1 Group # 1 K/A # EPE007 EK1.05 Rating 3.3 Question 39 The plant trips from full power.
The crew is checking if Source Range nuclear instruments should be energized.
The Source Ranges should be energized if its been approximately _____________ minutes since the reactor trip.
A. 3 to 4 B. 6 to 8 C. 12 to 14 D. 18 to 20 Proposed Answer: C. 12 to 14 Explanation:
A. Incorrect. Following a prompt drop in reactor power, a
-1/3 DPM SUR will cause power to decay until the Source Ranges energize in 14 minutes.
If power was lower, the minutes would be reasonable.
B. Incorrect. Following a prompt drop in reactor power, a
-1/3 DPM SUR will cause power to decay until the Source Ranges energize in 14 minutes. If its thought the either power drops further into the IR or unfamiliar with how many decades between the power range and the source range, then or the SUR is confused with increasing power during a startup (.75 DPM) its plausible to believe the source ranges may have just energized, this is approximately half the time it takes to reach the source ranges.
C. C orrect. Time to energize the source ranges should be approximately 14 minutes (between 4 and 5 decades at
-1/3 decade per minute)
D. Incorrect.
Time is too long, plausible if incorrect value used for IR or SUR is used
. Technical References
- LPE0, OIM B 3a References to be provided to applicants during exam: None Learning Objective
- Explain the plant response to reactor trip or SI (7388)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.2 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE008 AA1.02
- AA1.02 Ability to operate and / or monitor HPI (charging) pump to control PZR level/pressure as they apply to the Pressurizer Vapor Space Accident.
Tier # 1 Group # 1 K/A # APE 008 AA1.02 Rating 4.1 Question 40 GIVEN: A Pressurizer PORV fails open and cannot be isolated The plant trips and SI actuates 5 minutes after the reactor trip the crew enters E
-1, Loss of Reactor or Secondary Coolant Which of the following describes the expected plant conditions as the crew enters E
-1? Charging injection flow _______________ from the time of safety injection actuation and pressurizer level is ______________.
A. is unchanged; on
-scale and rising B. is unchanged; off
-scale low C. has risen; on
-scale and rising D. has risen; off
-scale low Proposed Answer:
C. has risen; on
-scale and rising Explanation:
A. Incorrect. Due to lowering RCS pressure, injection flow will have risen. If its thought that the injection throttle valves limit flow, this is plausible. Due to the location of the break, pressurizer level will be rising.
B. Incorrect. Due to lowering RCS pressure, injection flow will have risen. If its thought that the injection throttle valves limit flow, this is plausible. Unlike a vapor space break, a LOCA on a RCS loop would cause level to be offscale low.
C. Correct. Due to lowering RCS pressure, injection flow will have risen. Due to the location of the break, pressurizer level will be rising
. D. Incorrect. Due to lowering RCS pressure, injection flow will have risen. Unlike a vapor space break, a LOCA on a RCS loop would cause level to be offscale low.
Technical References
- LMCD-FRC 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 LOCA Question Source:
Bank # (note changes; attach parent)
Modified Bank #39 DCPP L111 NRC Exam 11/2012 X New Past NRC Exam Yes Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamenta l Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.5 Difficulty: 3.3
DCPP L161 Exam Rev 0 Examination Outline Cross-Reference Level RO EPE009 EA2.38 - Ability to determine or interpret the following as they apply to a small break LOCA: Existence of head bubble Tier # 1 Group # 1 K/A # EPE 009 EA2.38 Rating 3.9 Question 41 GIVEN: Following the performance to E-1.2, Post-LOC A Cooldown and Depre ssurization, the crew is performing FR I.3, Response to Voids in Reactor Vessel Charging and Letdown are matched There is a void in the reactor vessel head Which of the following would confirm that the vo id is a steam bubble in the reactor vessel head?
Pressurizer level lowers as char ging is ______________ or pressure is ____________.
A. lowered; raised B. raised; raised C. raised; lowered D. lowered; lowered Proposed Answer: B. raised; raised Explanation: System response is opposite of a normal response to changes in charging. If charging is raised, the bubble shrinks, causing level to lower. As char ging lowers, pressure lowers, causing level to rise. A. Incorrect. Charging lowered causes pressure to lower and the void to expand, raising pressurizer level. Plausible because raising pressure causes level to lower as the bubble collapses.
B. Correct. Level will lower for raising charging (raises pressure) or pressure is raised both cause the bubble to collapse.
C. Inorrect. Plausible - Raising ch arging lowers level however, lowering pressure also causes the bubble to expand, raising level.
D. Incorrect. Plausible, opposite effect - Loweri ng charging or pressure causes pressure to lower, and level will rise.
Technical References
- LPE-ZI References to be provided to applicants during exam:
None Learning Objective: Explain the basis of emergenc y procedure steps. (7920Q, R) Question Source: Bank # (note changes; attach parent) Modified Bank #
New X Question History: Last Two NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.41.2 Difficulty: 3.0 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO EPE 011 EA2.13
- EA2.13 Ability to determine or interpret the difference between overcooling and LOCA indications as they apply to a Large Break LOCA.
Tier # 1 Group # 1 K/A # EPE011 EA2.13 Rating 3.7 Question 42 According to the Diablo Canyon FSAR, which of the following completes the following statements? The highest containment pressure occurs if the LOCA occurs on the _________________.
This peak containment pressure for the worst case LOCA is ________________ than for a worst case steam break in containment.
A. RCS hot leg; lower B. RCS cold leg; lower C. RCS hot leg; higher D. RCS cold leg; higher Proposed Answer:
A. RCS hot leg; lower Explanation:
The candidate must determine which LOCA is the worst case and compare that response to the response of a steam break (overcooling event).
A. Correct. This is the worst case LOCA and the pressure peak for a LOCA is lower than the steam break (steam break has the highest pressure)
. B. Incorrect. The break on the RCS cold leg is not as severe as the hot leg. Plausible
- Containment pressure is higher for a steam break (lower for LOCA)
. C. Incorrect. Plausible, Location correct, pressure response incorrect.
D. Incorrect. Plausible, RCS is at higher pressure and temperature than the secondary. Both are incorrect.
Technical References
- LMCD FRZ References to be provided to applicants during exam: None Learning Objective
- Describe the limiting analysis for the Containment Critical Safety Function: (11331)
- a. Loss of Coolant Accide nt b. Steam Line Break Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.5 Difficulty: 3.2
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE015/017 AK3.01 Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow)
- Potential damage from high winding and/or bearing temperatures Tier # 1 Group # 1 K/A # APE 015/017 EK3.01 Rating 2.5 Question 43 Unit 1 is at full power. PK05
-01, RCP No , 1-1, is in alarm.
RCP 1-1 parameters:
Motor Stator Winding temperature: 280°F and rising at 5°F/minute Motor Radial Bearing temperatures:
175°F and rising at 3°F/minute
- 1 Seal return flow is 3.5 gpm and rising at 0.4 gpm Pump shaft vibration is 4.2 mils and rising at 0.
2 mils/minute Within the next 5 minutes, the RCP will exceed the operating limit designed to protect the:
A. RCP motor windings B. RCP motor bearings C. RCP #1 Seal D. RCP Pump Shaft Proposed Answer:
A. RCP motor windings Explanation:
A. Correct. In 5 minutes, stator winding temperature will be 305
°F. Limit is 300
°F to protect the stator windings.
B. Incorrect. Plausible
- Reason is correct, however, the limit of 200°F will not be exceeded in 5 minutes (190°F). C. Incorrect. Plausible
- Reason is correct, however limit of 8.0 gpm (6.0 if pump bearing temperatures rising) will not be exceeded in 5 minutes (5.5 gpm) D. Incorrect. Plausible
- Reason is correct, however, the vibration, while very high is less than the limit of 15 mils in 5 minutes (5.2 mils) Technical References
- LA-6, OP AP-28 References to be provided to applicants during exam: None Learning Objective
- Explain the effects of RCP operation during abnormal events (4897)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Conten t: 55.4 1.3 Difficulty: 2.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 022 AK1.03 Knowledge of the operational implications of the relationship between charging flow and PZR level as they apply to Loss of Reactor Coolant Makeup.
Tier # 1 Group # 1 K/A # APE022 AK1.03 Rating 3.0 Question 44 The plant is at full power, in a normal steady state lineup.
Without operator action, which of the following VCT level transmitter failures would cause pressurizer heaters to de
-energize due to low pressurizer level?
A. LT-112 fails low B. L T-112 fails high C. LT-114 fails low D. LT-114 fails high Proposed Answer:
B. LT-112 fails high Explanation:
Question addresses a "loss of makeup" (charging) from the perspective of the VCT makeup failure that leads to a loss of charging by causing VCT to drain and there is no swapover to RWST. Ultimately, there could be a loss of all charging due to gas intrusion from the VCT to the suction of the charging pumps.
To answer the question it must be understood that the loss of charging is the cause of the lowering pressurizer level and loss of heaters.
A. Incorrect.
Plausible if failure is misconstrued.
LT-112 LOW would initiate continuous makeup and the VCT fills and diverts. Charging and letdown unaffected.
B. Correct. Auto makeup is lost. VCT level will decrease. No auto swapover to the RWST, at 5% will occur (requires both channels, LT112 and LT114). VCT empties, charging flow decreases to 0 gpm. Pressurizer level decreases, letdown isolates and heaters de
-energize.
C. Incorrect. VCT level would be maintained by LT11
- 2. If level on LT
-112 were to also lower to 5%, auto swapover to RWST. Charging flow would be maintained.
Plausible if the functions for LT112 are assumed for LT114.
D. Incorrect. This fully opens LCV-112A. VCT level decreases to 14%, auto makeup from LT-112, makeup is greater than letdown.
Plausible if its thought auto makeup is defeated (as for LT112).
Technical References
- OIM B-1-4. OP AP-19 appendix A References to be provided to applicants during exam: None Learning Objective
- 40 449 - Discuss abnormal conditions associated with the CVCS Question Source:
Bank #43 DCPP L061C NRC exam 02/2009 X (note changes; attach parent)
Modified Bank #
New Past NRC Exam 02/2009 Yes Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.6 DCPP L 1 61 Exam Rev 0 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 025 AA1.23 Ability to operate and / or monitor RHR heat exchangers as they apply to the Loss of Residual Heat Removal System Tier # 1 Group # 1 K/A # APE025 AA1.23 Rating 2.8 Question 45 Unit 1 is in MODE 4. Both trains of RHR are in service.
Which of the following would be indicative of tube failure in an RHR heat exchanger?
CCW surge tank will __________; and flow to the RCS cold legs from the heat exchanger will
____________.
A. r is e; remain unchanged B. r is e; lower C. lower; remain unchanged D. lower; rise Proposed Answer:
B. rise; lower Explanation:
A tube failure results a loss of RHR due to leakage into CCW.
A. Incorrect. Unlike flow control such as charging, the flow is set by manual operation of the heat exchanger outlet valves. Loss of RHR flow to the CCW system will result in lower flow to the cold legs.
Plausible if its thought that like systems such as charging, flow is controlled by automatic valve control, and surge tank level will rise.
B. Correct. Flow will be from RHR to CCW. Because valve position is set by the operator, flow to the cold legs will lower.
C. Incorrect. Flow is into CCW not into RHR. Plausible if its believed CCW pressure higher than RHR, which is a low pressure system.
D. Incorrect. Flow is into CCW not into RHR. Plausible if its believed CCW pressure higher than RHR, which is a low pressure system.
Technical References
- OP AP-11, LB-2, OVID 106714-4, 1067 10-2 References to be provided to applicants during exam: None Learning Objective
- Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.3 Difficulty: 2.9
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 026 G2.
4.11 - Loss of CCW
- Knowledge of abnormal condition procedures Tier # 1 Group # 1 K/A # APE026 G2.4.11 Rating 4.0 Question 46 The crew has entered OP AP
-11, Malfunction of Component Cooling Water System.
In accordance with OP AP
-11, which of the following conditions would require the crew to trip the reactor and stop all RCPs? 1. Surge tank level off
-scale LOW 2. CCW temperature rises to 110°F
- 3. ASW is lost and cannot be recovered A. 1 only B. 2 and 3 C. 2 onl y D. 1 and 3 Proposed Answer:
D. 1 and 3 Explanation:
A. Incorrect. Of the three conditions listed, 1 and 3 require a reactor trip and stopping of all RCPs. Plausible because 3 states loss of ASW which may seem to imply its an action in OP AP-10 (Loss of ASW) not OP AP
-11. B. Incorrect. Trip and stopping of RCPs is required IF CCW exceeds 120°F. Plausible because 3 is a correct condition and temperature of 110°F is well above normal (less than 70°F).
C. Incorrect. Setpoint for this to be a correct condition is 120°F. Plausible as this is well above normal temperature (less than 70°F) and could be believed that a trip is required prior to exceeding system maximum temperature of 120°F.
D. Correct. Both condition 1 and 3 require a reactor trip and stopping RCPs per OP AP-11. Technical References
- OP AP-11 References to be provided to applicants during exam: None Learning Objective
- Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.10 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO EPE029 EK2.06
- EK2.06 Knowledge of the interrelations between breakers, relays, and disconnects following an ATWS.
Tier # 1 Group # 1 K/A # EPE029 EK2.06 Rating 2.9 Question 47 Unit 1 fails to automatically trip when required.
Which of the following failures would be the MINIMUM required to PREVENT a successful reactor trip and require the crew to perform FR
-S.1, Response to Nuclear Power Generation/ATWS?
A. One reactor trip breaker remains closed and one Rod Drive MG set remains energized.
B. One reactor trip breaker remains closed and both Rod Drive MG sets remain energized.
C. Both reactor trip breakers remain closed and one Rod Drive MG set remains energized
. D. Both reactor trip breakers remain closed and both Rod Drive MG sets are de
-energized. Proposed Answer:
C. Both reactor trip breakers remain closed and one Rod Drive MG set remains energized.
Explanation:
A. Incorrect. If one trip breaker is open, the reactor trips, regardless of the status of the MG sets. Plausible if it's the orientation of trip breakers and MG sets is not known.
B. Incorrect. If one trip breaker is open, the reactor trips, regardless of the status of the MG sets. Plausible if trip breakers, like the MG sets were in parallel.
C. Correct. This is the minimum, the MG sets are in parallel, only one running with the trip breakers closed will result in an ATWS.
D. Incorrect. Plausible, as this would prevent a trip, however, this is not the minimum.
Technical References
- OIM References to be provided to applicants during exam: None Learning Objective
- 9903 - Discuss abnormal conditions associated with the Rod Control System Question Source:
Bank #46 DCPP NRC Exam (L061C) 02/2009 X (note changes; attach parent)
Modified Bank #
New Past NRC Exam 02/2009 Yes Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.6 Difficulty: 2.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO EPE 038 EK1.02 Knowledge of the operational implications of leak rate vs. pressure drop as they apply to the SGTR.
Tier # 1 Group # 1 K/A # EPE038 EK1.02 Rating 3.2 Question 48 Which of the following actions taken in E
-3, Steam Generator Tube Rupture, is designed to prevent reinitiation of primary to secondary leakage due to uncontrolled depressurization of the ruptured steam generator
? A. Promptly terminating Safety Injection when conditions are satisfied.
B. Maintaining level in the ruptured steam generator above the top of the U
-tubes. C. Terminating depressurization if RCS subcooling lowers to less than 20°F during depressurization
. D. Cooling down the RCS to a temperature that allows subsequent depressurization to less than ruptured steam generator pressure.
Proposed Answer:
B. Maintaining level in the ruptured steam generator above the top of the U-tubes. Explanation:
All are plausible as they are all actions taken when dealing w/a steam generator tube rupture
. A. Incorrect. Terminating SI is important to control RCS pressure which, if allowed to rise above SG pressure will increase primary to secondary leakage but it will not cause uncontrolled depressurization of the steam generator.
B. Correct. It is also important to maintain the water level in the ruptured steam generator above the top of the U
-tubes. When the primary system is cooled in subsequent steps, the steam generator tubes in the ruptured steam generator will approach the temperature of the reactor coolant, particularly if reactor coolant pumps continue to run. If the steam space in the ruptured steam generator expands to contact these colder tubes, condensation will occur which would decrease the ruptured steam generator pressure. As previously demonstrated (see Step 3), this would reduce the reactor coolant subcooling margin and/or increase primary-to-secondary leakage, possibly delaying SI termination or causing SI reinitiation
. C. Incorrect. Loss of subcooling is indicative of a loss of RCS and would lead to entry into ECA-3.1. D. Incorrect.
This step is designed to allow subsequent depressurization and terminate break flow but it is not to prevent uncontrolled steam generator depressurization.
Technical References
- E-3 Background References to be provided to applicants during exam: None Learning Objective
- Explain basis of emergency procedure steps (E
-3 series) (7920F) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis
DCPP L 1 61 Exam Rev 0 10CFR Part 55 Content:
55.4 1.5 Difficulty: 3.25
DCPP L 1 61 Exam Re v 0 Examination Outline Cross
-Reference Level RO APE040 G2.4.2
- Steam Line Rupture
- Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.
Tier # 1 Group # 1 K/A # APE 040 G2.4.2 Rating 4.5 Question 49 Reactor power is 4%.
A large steam break inside containment occurs. Two minutes later the reactor is tripped, all Safety Injection actuation setpoints have been reached and the MSIVs are closed. Containment Spray has not actuated.
- 1. Which of the following could have caused the reactor to trip? (RCS low pressure trip and/or Safety Injection )
- 2. Which of the following could have caused the MSIVs to close? (High Containment Pressure or Low Steam Line Pressure)
A. 1. RCS low pressure trip or Safety Injection. 2. Low Steam Line pressure only.
B. 1. RCS low pressure trip or Safety Injection. 2. High Containment pressure or Low Steam Line pressure.
C. 1. Safety Injection only. 2. Low Steam Line pressure only.
D. 1. Safety Injection only. 2. High Containment pressure or Low Steam Line pressure.
Proposed Answer:
C. 1. Safety Injection only. 2. Low Steam Line pressure only.
Explanation:
A. Incorrect. RCS low pressure trip blocked at less than 10% power.
Plausible if the P
-7 block is missed and SI does cause a trip and 2 is correct.
B. Incorrect. High Containment pressure, although causing Phase A (and SI) to isolate containment, does not close the MSIVs.
Plausible as High
-High Containment pressure will cause MSIVs to close.
C. Correct. SI will cause a reactor trip. The MSIVs are closed by the low steam line pressure signal. D. Incorrect. High containment pressure does not close the MSIVs.
Plausible as except for high containment pressure answer is correct. High
-High Containment pressure will cause MSIVs to close, not High Containment pressure
. Technical References
- OIM B-6-4b, B-6-5, B-6-6, B-6-8 and B-6-10 References to be provided to applicants during exam: None Learning Objective
- 41313 - Analyze automatic features and interlocks associated with the Eagle-21/SSPS Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
DCPP L 1 61 Exam Re v 0 10CFR Part 55 Content:
55.4 1.7 Difficulty: 2.5
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 054 EK3.03 Knowledge of the reasons for manual control of AFW flow control valves as they apply to the Loss of Main Feedwater (MFW).
Tier # 1 Group # 1 K/A # APE054 AK3.03 Rating 3.8 Question 50 GIVEN:
- A total loss of Main Feedwater and Auxiliary Feedwater occurs
- The crew is performing the actions of FR
-H.1, Response to Loss of Secondary Heat Sink
- Bleed and Feed has been initiated
- All steam generators are "dry"
- Core Exit Thermocouples are 555°F, rising slowly The capability to feed all steam generators using the TDAFW pump has been restored.
What flow rate will be established and what is the reason for the flowrate?
A. Fully open one TDAFW LCV to a steam generator due to the urgent need to restore a heat sink. B. Fully open all TDAFW LCVs to the steam generator s due to the urgent need to restore a heat sink C. Throttle open one TDAFW LCV to establish approximately 100 gpm to one steam generator to ensure heat removal capability will be greater than decay heat. D. Throttle open all TDAFW LCVs to establish at total AFW flow of approximately 100 gpm to the steam generators to ensure heat removal capability will be greater than decay heat.
Proposed Answer:
A. Fully open one TDAFW LCV to a steam generator due to the urgent need to restore a heat sink
. Explanation:
A. Correct. With core exit temperatures increasing, maximum flow to restore a heat sink as quickly as possible is necessary
. B. Incorrect because only one steam generator is used in order to limit potential faults to only that steam generator. Plausible because the need to re
-establish a heat sink is urgent and the TDAFW pump would normally be used to feed all steam generators.
C. Incorrec t because maximum flow to one steam generator is to be used if core exit thermocouples are rising. Plausible because 100 gpm is the top end of the allowable flow band (25-100 gpm) to one dry steam generator i f core exit temperatures are lowering.
D. Incorrect because only one steam generator is used in order to limit potential faults to only that steam generator. Plausible because the TDAFW pump would normally be used to feed all steam generators, and the restricted flow band of 25
-100 gpm would be used if core exit temperatures were stable or lowering. Technical Reference s: FR-H.1 foldout page, FR
-H.1 background section 2.4 References to be provided to applicants during exam: None Learning Objective
- Explain feeding a dry S/G including: (6375)
Definition of dry S/G Effects of feeding a dry S/G Strategy for feeding a dry S/G DCPP L 161 Exam Rev 0 Question Source:
Bank #9 L091, 08/2011 X (note changes; attach parent)
Modified Bank #
New Past NRC Exam Yes Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.4 1.5 Difficulty 3.8
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 056 AA2.21 Ability to determine and interpret the following as they apply to the Loss of Offsite Power: ED/G frequency and voltage indicators Tier # 1 Group # 1 K/A # APE056 AA2.09 Rating 3.6 Question 51 A loss of 230 kV and 500 kV power has occurred. All 4 kV vital buses are energized from their respective emergency diesel generator.
Which of the following describes the effect, if any, on indicated bus frequency and voltage if an operator takes an emergency diesel generator "D/G Voltage Control Switch" to "RAISE"? A. Both voltage and frequency will rise.
B. No change in either voltage or frequency indication.
C. Frequency indication will rise, no change in voltage indication.
D. Voltage indication will rise, no change in frequency indication.
Proposed Answer:
D. Voltage indication will rise, no change in frequency indication.
Explanation:
A. Incorrect because when started due to a loss of power, the EDG is in the isochronous mode of operation. Speed and frequency are set by the engine governor and do not change. Plausible because the operator can raise voltage, and it may be thought that doing so may affect frequency. B. Incorrect because even in Isochronous, voltage will rise in response to the Voltage Control Switch. Plausible because frequency is not adjustable, and it may be thought that voltage is set as well
. C. Incorrect because frequency is not adjustable. Plausible because which parameter is adjustable in isochronous and which is not may be misunderstood, and both speed and voltage control switches have raise and lower positions and are active in the droop mode
. D. Correct. Taking the voltage control switch to raise will cause voltage to rise, and will not affect frequency. Frequency is held constant by the engine governor in isochronous mode.
Technical References
- LJ-6B References to be provided to applicants during exam: None Learning Objective
- Describe the operation of the Diesel Generator System. (6437) Question Source:
Bank # (note changes; attach parent) Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.7 Difficulty 3.0
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE057 G2.4.3 - Loss of Vital AC Instrument Bus
- Ability to identify post
-accident instrumentation.
Tier # 1 Group # 1 K/A # APE057 G2.4.3 Rating 3.7 Question 52 Unit 1 is at full power.
A loss of a vital 120 VAC instrument bus occurs, resulting in one Power Range channel , one Intermediate Range channel
, and one Gamma Metrics channel being de-energized.
Which of the de
-energized nuclear instrumentation is Post Accident Monitoring (PAM) instrumentation?
A. All three are PAM instruments B. Only the Power and Intermediate Range channels C. Only the Power Range and the Gamma Metrics channels D. Only the Gamma Metrics channel Proposed Answer:
D. Only the Gamma Metrics channel Explanation:
A. Incorrect because the Power and Intermediate Range channels are not PAM instrumentation. Plausible because they are referred to by operators in post
-trip activities and perform important functions such as indicate on SPDS.
B. Incorrect because the Power and Intermediate Range channels are not PAM instrumentation, while the Gamma Metrics channels are. Plausible because Power and Intermediate Range channels are important to reactor protection and feed SPDS, and because the Gamma Metrics can be used during refueling activities.
C. Incorre ct because Power Range channels are not PAM instruments
. Plausible because Gamma Metrics are PAM, and Power Range channels provide Reactor Power information to SPDS for the Subcriticality CSFST
. D. Correct. Of the three instruments, only the Gamma Metrics channel is a qualified PAM instrument.
PAM instrumentation is designed to be available in harsh environments, such as a LOCA. While the Power and Intermediate Range instruments feed instrumentation such as SPDS, they are not PAM instrumentation.
Technical References
- LB-4, LB-10 References to be provided to applicants during exam: None Learning Objective
- Describe PAMS components. (40462) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.2 Difficulty 2.8
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE062 EK3.03
- Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water
- Guidance actions contained in EOP for Loss of nuclear service water (*Aux Saltwater)
Tie r # 1 Group # 1 K/A # APE 062 EK3.03 Rating 4.0 Question 53 Unit 1 is at full power.
The following occurs:
PK01-01, ASW SYS HX DELTA P/HDR PRESS alarms due to input 85, Aux Salt Wtr To CCW Ht Exch 1
-2 Press Lo The operator reports 1
-2 ASW pump amps are high and out of the normal ban d CCW temperature is 75°F and rising slowly What is the reason the crew will place the idle ASW loop in service?
A. T he running pump is cavitating.
B. There are signs of a piping rupture. C. To provide additional system cooling
. D. There are signs of fouling of the CCW heat exchanger
. Proposed Answer:
B. There are signs of a piping rupture.
Explanation:
A. Incorrect because the signs of cavitation are amp swings, and that is not indicated. Plausible because amps are not normal, and in the event of cavitation, the other ASW pump would be started.
B. Correct. Per OP AP
-10, rupture is indicated by low header press AND high pump amps. The idle loop is started in response
. C. Incorrect because, while CCW temperature is elevated, the action for high temperature is to reduce loads.
Plausible because temperature is high and rising and improved cooling is needed. D. Incorrect because fouling is indicated by high CCW HX diff pressure AND low pump amps. Plausible because temperature is rising, therefore the in-service HX is not being effective due to reduced flow. It may be thought that the reduced flow is due to fouling.
Technical References
- OP AP-10 References to be provided to applicants during exam: None Learning Objective
- State the purpose of actions when restoring ASW after malfunctions. (5354) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty 2.3
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 065 AA1.05 - Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air. (CFR 41.7 / 45.5 / 45.6) RPS Tier # 1 Group # 1 K/A # APE065 AA1.05 Rating 3.3 Question 54 Unit 1 is at full power.
A rupture in the instrument air system causes instrument air pressure to lower rapidly.
In a short period of time, the reactor will trip due to __________________.
A. High-high steam generator level B. Low-low steam generator level C. High Pressurizer level D. Low Pressurizer pressure Proposed Answer:
B. Low-low steam generator level Explanation:
A. Incorrect because the Main Feedwater (MFW) Reg valves fail closed causing low steam generator level. Plausible if it is thought the Main Feedwater Reg valves fail open on loss of air. B. Correct. the Main Feedwater Reg valves fail closed on loss of air, and the reactor very soon will trip on low
-low steam generator levels. C. Incorrect because the reactor will trip on low
-low steam generator level. Plausible because the combination of valves failing on loss of air will cause letdown and charging to isolate, but flow to continue to the RCP seals causing pressurizer level to slowly rise. However, the immediate concern is low steam generator level
- s. (FCV-584 (Instrument Air to Containment
) fails close d, FCV-128 (Charging flow) fails open, HCV
-142 (Charging Backpressure Control Valve) fails closed forcing more flow to RCP seals
) D. Incorrect because the reactor will trip on low
-low steam generator level. Plausible because it may be thought that the Pressurizer spray valves would fail open on loss of air, causing pressure to rapidly lower. However, they fail closed.
Technical References
- OP AP-9 References to be provided to applicants during exam: None Learning Objective
- 3541 - List the effects that a loss of Instrument Air would have on the plant Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty 2.8
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO E05 EK2.1 Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Tier # 1 Group # 1 K/A # EK2.1 Rating 3.7 Question 55 GIVEN: Green lights are lit for the reactor trip breakers Red lights on Monitor Light Box C are lit:
o Safety Injection o FW Isol o SG Lvl AFW flow is 0 gpm RCS pressure is 1900 psig and rising slowly PK09-11, FEEDWATER ISOLATION, is lit Wide range steam generator levels are approximately 44%
and lowering slowly The crew has reached step 7, TRY To Establish Mn Fdwtr Flow To At Least One S/G, in FR
-H.1, Loss of Secondary Heat Sink.
The step has the crew take the following actions:
- 1. Block Low Pressurizer pressure and Main Steamline low pressure SI
- 2. Reset SI 3. Cycle the reactor trip breakers
- 4. Reset Feedwater Isolation For the current plant conditions, which of the actions were the minimum required in order to open the Main Feedwater Reg valves? A. 4 only B. 2 and 4 only C. 2 and 3 and 4 only D. 1 and 2 and 3 and 4 Proposed Answer:
C. 2 and 3 and 4 only Explanation:
Restoring feedwater is essential for a LOHS. The question tests the components, interlocks and signals that must be addressed in order to restore a heat sink.
A. Incorrect because the SI and Reactor Trip (P4) signals being present will prevent resetting the FW Isolation in this case. Plausible because Reactor Trip with low Tavg causes FW Isolation, and does not require other actions before FW Isolation can be reset. The requirement to reset SI and cycle the reactor trip breakers may be misunderstood.
B. Incorrect because is necessary in this case to cycl e the reactor trip breakers. Plausible because reset of SI is one of the required actions. C. Correct. SI can be reset and the breakers cycled. Although the step has Auto SI blocked, at this time, SI would reset because there is not an active signal (Pressurizer Pressure is above the SI setpoint of 1850 psig.)
D. Incorrect because block of SI, for the current plant conditions, is not required to open the valves. Plausible because it may be thought that blocking SI is required to reset SI.
DCPP L 161 Exam Rev 0 Technical References
- FR-H.1, LB-6A References to be provided to applicants during exam: None Learning Objective
- Explain basis of emergency procedure steps (FR-Hs) including: (7920 N) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty 4.0
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 077 AK2.05 Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and Pumps.
Tier # 1 Group # 1 K/A # APE077 AK2.05 Rating 3.1 Question 56 Unit 1 is at full power.
A large seismic event causes a loss of the 500 kV switchyard. Additionally, 230 kV instability causes 4 kV vital bus voltage to lower and energize Second Level Undervoltage Relays (SLUR). Which of the following describes the response of pump s running on a 4 kV vital bus
? Pump speed lowers and amps
_______ until the ___________ is/are opened by the load shed relay. A. rise; startup feeder breaker B. lower; startup feeder breaker C. rise; pump breakers D. lowe r; pump breakers Proposed Answer:
C. rise; pump breakers Explanation:
A. Incorrect because the SLUR will energize the load shed relay and open the breaker for all loads on the bus (except CFCUs and 480V MCCs) not the feeder breaker
. Plausible because startup and auxiliary feeder breakers are opened on a transfer to diesel, however by that time the pump motor breakers have been opened by the load shed relays.
B. Incorrect because as frequency lowers, motor amps will rise due to the reduction of counter EMF. Also incorrect because the load shed relays will open the pump motor breakers, not the startup feeders. Plausible because it may be thought that as pumps slow down and do less work, they will draw less current. Also plausible because the startup feeder breakers will be opened, but not by the load shed relays.
C. Correct. Speed will lower as voltage lowers, causing amps to rise. The load shed relay will strip the bus.
D. Incorrect because motor amps will rise.
Plausible because it may be thought that motor amps will lower as frequency lowers, and it is correct that the load shed relay will strip the bus. Technical References
- OIM J-5-1e References to be provided to applicants during exam: None Learning Objective
- Analyze automatic features and interlocks associated with the Electrical Power Transfer System.
4 kV Vital Bus Shed Loads Relay Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No DCPP L 161 Exam Rev 0 Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.8 Difficulty 3.3 DCPP L 161 Exam Re v 0 Examination Outline Cross
-Reference Level RO APE 028 AK3.02 Knowledge of the reasons for the relationship between PZR pressure increase and reactor makeup/letdown imbalance as they apply to the Pressurizer level Control Malfunctions.
Tier # 1 Group # 2 K/A # APE028 AK3.02 Rating 2.9 Question 57 Unit 1 is at 50%. Initially, all pressurizer level channels are indicating the same level
. Subsequently, the operator reports the following:
PK05-21, PZR LEVEL HI/LO, is in alarm There are no other alarms RCS pressure is 2240 psig and rising Backup pressurizer heaters are energized Charging is 75 gpm Which of the following explains the reason(s) for these indications
? 1. Median pressurizer level channel 5% above reference level
- 2. Highest pressurizer level channel 5% above reference level
- 3. A Pressurizer level channel fails high A. 1 only B. Either 1 or 3 C. 2 only D. Either 2 or 3 Proposed Answer:
A. 1 only Explanation:
A. Correct. If the median signal select signal is 5% above reference the heaters will energize. This will raise RCS pressure, and the rising level above program will cause charging to lower to turn level.
B. Incorrect because i f a channel fails high, it is thrown out, and it would also cause PK06-21, PCS Rack Trouble Alarm Inputs and Response. Plausible because the responses listed are for indicated level being high and it may be thought that the channel failing high is the cause. C. Incorrect because it is the median channel that controls. Plausible because there are control systems, such at Tave
, which are auctioneered high, and it may be thought that a failed high level channel could cause the responses listed
. D. Incorrect because it is the median channel that controls. Plausible because there are control systems, such at Tave, which are auctioneered high, and it may be thought that the highest channel, or a failed high level channel, could cause the responses listed.
Technical References
- OIM A-4-2a, A-4-3, PK05-21 References to be provided to applicants during exam: None Learning Objective
- Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X DCPP L 161 Exam Re v 0 Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty 2.5
DCPP L161 Exam Rev 0 Examination Outline Cross-Reference Level RO APE 036 AA2.01 - Ability to determine and interpret the following as they apply to the Fuel Handling Incidents: ARM system indications Tier # 1 Group # 2 K/A # APE036 AA2.01 Rating 3.2 Question 58 GIVEN: Unit 1 is at full power Fuel assemblies are being move d in the Spent Fuel Pool An operator reports the following indications for RM-58 o Red Light is ON for Trip 2 o Amber Light is ON for Trip 1 o Green Operate Light is ON Which of the following POV1 (2) indications would be consistent with the RM-58 indication?
- 1. No supply fans running and Exhaust Fan E-5 running
- 2. Supply Fan S-2 and Exhaust Fan E-5 running
- 3. Red "Iodine Removal" indi cation on POV1 and POV2 A. 1 only B. 2 only C. 1 and 3 D. 2 and 3 Proposed Answer: D. 2 and 3 Explanation:
A. Incorrect because a supply fan will remain running if an exhaust fan is running. Plausible because it may be thought that supply fans should be stopped to force a negative pressure to prevent radioactivity from the room from exiting anywhere except out the ventilation system past effluent radiation monitors. However, fan capacities ensure a negative pressure with one supply and one exhaust fan running.
B. Incorrect because the iodine removal light on the POV panels will also be lit. Plausible because it may be thought that the Iodine Removal Mode is selected strictly manually.
C. Incorrect because it may be thought that supply fans should be stopped to maintain a negative pressure. Plausible because the Iodine Removal light will be lit.
D. Correct. If RM-58 is in high alarm, FHBVS shifts to Iodine Removal and the exhaust shifts to E-5 or E-6 to align the exhaust th ru a charcoal filter. One supply fan runs, the exhaust is thru E-5 or E-6. Because the exhaust is greater than supply a slight negative pressure is maintained in the FHB. High alarm causes ventilation to shift to Iodine Removal and the red light would be lit on the POV panels. .
Technical References
- LH-7 References to be provided to applicants during exam:
None Learning Objective: Discuss abnormal conditions associated with the Fuel Handling Building Ventilation System. (40721) 10 4 10 3 10 2 10 1 10 0 10 -1 Slope BiasMete rTrip 1TripRM 1RM 2TP 3TP 2TP 1Adj Adj Adj Re f SetOperateTrip A djCheck Trip 2 TP 4 HV SetOperateTrip 1Trip 2 Red Ambe rGreen DCPP L 161 Exam Rev 0 Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.11 Difficulty 3.3
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 059 AA1.02 Ability to operate and / or monitor the ARM* system as they apply to the Accidental Liquid Radwaste Release.
(* At DCPP the PRM system performs this function.)
Tier # 2 Group # 1 K/A # APE0 59 AA1.02 Rating 3.3 Question 59 Unit 1 is at full power.
PK11-17, SG BLOW DOWN HI RAD, input 508, Steam Gen Blowdown Sample Hdr Hi Rad , alarms. Which of the following should have occurred?
- 1. Inside SG Blowdown Isolation valves closed
- 3. FCV-498, Disch Tunnel, closed and FCV
-499, Equip Drn Rcvr, opened A. 1 only B. 1 and 3 C. 2 only D. 2 and 3 Proposed Answer:
D. 2 and 3 Explanation:
A. Incorrect because the outside containment valves automatically close on high blowdown activity. Plausible because the inside valves automatically close on main steamline isolation (MSI).
B. Incorrect because the outside containment valves automatically close on high blowdown activity. Plausible because the inside valves automatically close on main steamline isolation (MSI), and the blowdown effluent flow does swap to the Equipment Drain Receiver on high radiation
. C. Incorrect because the blowdown effluent flow swaps to the Equipment Drain Receiver on high radiation. Plausible because the outside containment isolation valves close on high radiation and that would terminate flow into the blowdown system.
D. Correct. Either RE-19 or 23 will isolate the system by closing the outside containment, sample valves and terminate blowdown effluent flow to the discharge tunnel by aligning to the Equipment Drain Receiver
. Technical References
- LD- 2, AR PK11
-17 References to be provided to applicants during exam: None Learning Objective
- 8724 Analyze automatic features and interlocks associated with the SGBD system Question Source:
Bank # DCPP bank CR-0074 X (note changes; attach parent)
Modified Bank #
New Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental
DCPP L 161 Exam Rev 0 Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.7 Difficulty 2.3
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO APE 069 G2.1.7 - Loss of Containment Integrity: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Tier # 1 Group # 2 K/A # APE069 G2.1.7 Rating 4.4 Question 60 GIVEN: The Shift For eman has entered FR
-Z.1, Response to High Containment Pressure The operator reports the following:
o All MSIVs are closed o Pressure in all steam generators is approximately 3 50 psig and lowering rapidly o Containment pressure is rising o Narrow range level in all steam generator is 5% and lowering o Total AFW flow is 300 gpm Per FR-Z.1, AFW flow should be _________.
A. raised to greater than 435 gpm total flow to all steam generators B. lowered to approximately 25 gpm to each steam generator C. isolated to three steam generators and raised to greater than 435 gpm to one steam generator D. isolated to three steam generators and lowered to approximately 25 gpm to one steam generator Proposed Answe r: B. lowered to approximately 25 gpm to each steam generator Explanation:
A. Incorrect.
With narrow range levels less than required, normal action is to maintain (raise) total AFW flow to greater than 435 gpm.
B. Correct. All steam generators are faulted. Action is to establish (greater than) 25 gpm to each steam generator to maintain components "wet".
C. Incorrect. This would be appropriate if one was being maintained for RCS cooldown, however, all are faulted.
D. Incorrect. would maintain components in one steam generator "wet" could be thought this would maintain the generator "available for RCS cooldown".
Technical References
- FR-Z.1 step 7 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 Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.5 DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO EPE 074 EK3.08 Knowledge of the reasons for the Securing RCPs as they apply to the Inadequate Core Cooling.
Tier # 1 Group # 2 K/A # EPE074 EK3.08 Rating 4.1 Question 61 The crew is performing FR
-C.1, Response to Inadequate Core Cooling. The crew has performed an initial secondary depressurization to inject accumulators.
Before beginning a subsequent secondary depressurization to atmospheric pressure, the procedure directs stopping all RCPs.
Which of the following describes why the RCPs are secured?
A. To remove the heat addition associated with RCP operation
. B. If they trip during the depressurization, a deeper core uncovery could occur
. C. To uncouple the steam generators, which are a heat source at this time, from the RCS. D. It is anticipated number 1 seal requirements will be lost and operation could damage the seals. Proposed Answer:
D. It is anticipated number 1 seal requirements will be lost and operation could damage the seals.
Explanation:
A. Incorrect because they are stopped to protect the #1 seals from damage. Plausible because they are secured for this reason in FR
-H.1, Loss of Secondary Heat Sink.
B. Incorrect because they are stopped to protect the #1 seals from damage. Plausible because this is a reason in the executive volume related to early RCP stop/restart.
C. Incorrect because they are stopped to protect the #1 seals from damage. Plausible because in situations where RCS pressure is lower than steam generator pressure, the steam generators can be a heat source. However, in this situation, the RCS pressure will still be higher than steam generator pressure and therefore the steam generators are still heat sinks
. D. Correct. The depressurization is going to lower pressure in the RCS to less than required for RCP number 1 seal. Loss of adequate seal DP could damage the seals if the pumps are allowed to continue to operate.
Technical References
- FR-C.1 and background, Executive Volume RCP Trip/Restart, H.1 and C.2 References to be provided to applicants during exam: None Learning Objective
- Explain basis of emergency procedure steps (FR
-Cs). (7920M)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.5 Difficulty 3.0
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO E02 EK1.2 Knowledge of the operational implications of the Normal, abnormal and emergency operating procedures associated with (SI Termination).as they apply to the (SI Termination)
Tier # 1 Group # 2 K/A # E02 EK1.2 Rating 3.4 Question 62 GIVEN: Containment pressure is 2.5 psig after peaking at 4 psig RCS subcooling is 23°F RCS pressure is 1700 psig and stable Steam Generator Narrow Range levels:
o 22%, rising slowly o 18%, rising slowly o 13%, lowering slowly o 10%, lowering slowly 1-2 AFW pump is running Total AFW flow is 400 gpm Pressurizer level is 1 8%, rising slowly The crew is checking if Safety Injection, (SI), can be terminated in E
-1, Loss of Reactor or Secondary Coolant.
Which of the following describes the status of SI termination criteria for Secondary Heat Sink and Pressurizer Level
? Secondary Heat Sink Pressurizer Level A. Met Met B. Met Not Met C. Not Met Met D. Not Met Not Met Proposed Answer:
A. Met Met Explanation:
A. Correct. Once containment pressure is below 3 psig, then the following are necessary to meet SI termination criteria:
RCS Subcooled based on core exit T/Cs
- GREATER THAN 20°F Total AFW flow greater than 435 gpm OR ONE intact SG level greater than 15%
RCS pressure STABLE or rising Pressurizer level greater than 12%
All are met at this time.
B. Incorrect because Pressurizer level criteria is met. Plausible because it may be misunderstood when to apply adverse containment parameter values. Containment pressure was above the criterion (3.0 psig) for use of adverse containment values (Pressurizer level DCPP L 161 Exam Rev 0 min 40%), but is now less than that criterion.
C. Incorrect because the steam generator level criterion is met (at least one steam generator NR level is >15%)
. Plausible because it may be misunderstood when to apply adverse containment parameter values. Containment pressure was above the criterion (3.0 psig) for use of adverse containment values (Pressurizer level min 40%), but is now less than that criterion.
The adverse containment value for steam generator level is 25%, and would be Not Met.
D. Incorrect because both steam generator and pressurizer level criteria are met
. Plausible if adverse containment parameter values are used. In that case, both would be Not Met
. Technical
References:
E-1 step 8 References to be provided to applicants during exam: None Learning Objective
- Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.5 Difficulty 3.3
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO E03 EK2.2 Knowledge of the interrelations between the (LOCA Cooldown and Depressurization) and the facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Tier # 1 Group # 2 K/A # E03 EK2.2 Rating 3.7 Question 63 GIVEN: The crew is performing E
-1.2, Post-LOCA Cooldown and Depressurization Narrow range steam generator levels are approximately 45%
Pressurizer level is 35%
RVLIS level is 85%
The crew has just started a RCP. Which of the following describes the expected pressurizer level response?
A. Rise due to an increase in core heat removal.
B. Lower due to an increase secondary heat removal
. C. Rise due to transfer of voids in the reactor vessel head to the pressurizer
. D. Lower due to collapse of voids in the reactor vessel head.
Proposed Answer:
D. Lower due to collapse of voids in the reactor vessel head.
Explanation
- A. Incorrect because voids in the head will be removed by the forced flow. The void volume will be filled in by liquid volume from the pressurizer. Plausible because forced flow will increase the ability to remove core decay heat.
B. Incorrect because voids in the head will be removed by the forced flow. The void volume will be filled in by liquid volume from the pressurizer. Plausible because an increase in secondary heat removal could cause a cooldown, which would eventually result in shrinkage of the RCS and pressurizer level to lower.
C. Incorrect because voids in the head will be removed by the forced flow. The void volume will be filled in by liquid volume from the pressurizer. Plausible because voids will cause a change in pressurizer level, and the exact response may be misunderstood. Cautions in many EOPs note that creation of a void causes pressurizer level to rise. In this case, the void will collapse.
D. Correct. Starting the RCP will collapse voids and cause pressurizer level to lower (this is a reason level is raised prior to starting the RCP.
Technical References
- E-1.2, LPE-1B References to be provided to applicants during exam: None Learning Objective
- Explain basis of emergency procedure steps (E
-1.1, E-1.2). (7920S)
Question Source:
Bank #26 DCPP NRC exam 12/2007 X (note changes; attach parent)
Modified Bank #
New Past NRC Exam X Question History:
Last Two NRC Exam s No DCPP L 161 Exam Rev 0 Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.5 Difficulty 2.8 DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO E13 EA2.1 Ability to determine and interpret the facility conditions and selection of appropriate procedures during abnormal and emergency operations as they apply to the (Steam Generator Overpressure)
Tier # 1 Group # 2 K/A # E13 EA2.1 Rating 2.9 Question 64 The crew is performing E
-0.2, Natural Circulation Cooldown.
A stagnant loop develops and pressure in one of the steam generators rises to 1120 psig. The STA reports the Critical Safety Function Status Heat Sink is YELLOW for H.2, Response to Steam Generator Overpressure.
What should be the status of the steam generator and the action that should be taken by the crew?
A. All steam generator safeties should be open and the crew must stop performing E
-0.2 and go to FR-H.2. B. One steam generator safety should be open and the crew must stop performing E
-0.2 and go to FR-H.2. C. All steam generator safeties should be open and the crew may, at the discretion of the Shift Foreman, stop performing E
-0.2 and go to FR
-H.2. D. One steam generator safety should be open and the crew may, at the discretion of the Shift Foreman, stop performing E
-0.2 and go to FR
-H.2. Proposed Answer:
C. All steam generator safeties should be open and the crew may, at the discretion of the Shift Foreman, stop performing E
-0.2 and go to FR
-H.2. Explanation:
The steam generator safety valves open at: 1065 psig, 1078 psig, 1090 psig, 1103 psig and 1115 psig. A. Incorrect because FR
-H.2 is a yellow path and not mandatory. Plausible because a ll related steam generator safeties should be open (greater than 1115 psig)
. B. Incorrect because FR
-H.2 is a yellow path and not mandatory. Plausible because pressure is above the setpoint for the lowest set safety valve. C. Correct. All safeties should be open.
A Yellow CSF procedure may be performed at the discretion of the Shift Foreman.
D. Incorrect because all related steam generator safety valve are open. Plausible because pressure is above the setpoint for the lowest set safety valve, and a Yellow CSF procedur e may be performed at the discretion of the Shift Foreman.
Technical References
- F-0, LC-2A References to be provided to applicants during exam: None Learning Objective
- Describe Main Steam System components. (37594)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X
DCPP L 161 Exam Rev 0 10CFR Part 55 Content:
55.4 1.8 Difficulty 2.8
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO E16 EK2.1 Knowledge of the interrelations between the (High Containment Radiation) and the Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Tier # 1 Group # 2 K/A # E16 EK2.1 Rating 3.0 Question 65 GIVEN: Unit 1 is in MODE 6 PK02-11, MODE 6 CVI CUT
-IN, is lit, due to input 1127, Mode 6 CVI Cut
-In Train B The Train A input to PK02
-11 is NOT in High radiation is detected by both Containment Purge Exhaust Duct radiation monitors, RM
-44A and RM-44B. What will be the status of Containment Ventilation Isolation, CVI? A. All the open CVI valves have closed. B. Only the CVI valves associated with Train A have closed, any open Train B valves are still open. C. Only the CVI valves associated with Train B have closed, any open Train A valves are still open. D. All of the open CVI valves are still open
. Proposed Answer:
A. All the open CVI valves have closed. Explanation
- In Mode 6, all ESF actuations are blocked from causing CVI, and ONLY high rad will cause CVI. In this question, only one of the trains is in Mode 6 and the other is not (one input in alarm, the other is not).
Regardless of whether Mode 6 is Cut
-In, if there is high radiation, BOTH trains still respond, and EITHER train would cause ALL CVI valves to close. A. Correct. All CVI valves will respond and close on high radiation, whether the train is in Mode 6 or not. B. Incorrect because all valve will respond. Plausible if the function of placing the rad monitor in Mode 6 is not understood, and it is thought that with Train B Mode 6 Cut
-In the radiation monitor signal to CVI to that train is defeated. Either train will cause all valves to close. C. Incorrect because all valve will respond. Plausible if the function of placing the rad monitor in Mode 6 is not understood, and it is thought that only with Mode 6 Cut
-In will the valves respond to a high radiation signal.
Either train will cause all valves to close.
D. Incorrect because both trains will respond. Plausible if it is believed a signal from both train s is required to initiate CVI, and it will not occur with one train in Mode 6 and the other not. However, regardless of Mode 6 status, either train will cause all valves to close.
Technical References
- OIM B-6-9a References to be provided to applicants during exam: None Learning Objective
- Analyze automatic features and interlocks associated with the Digital Radiation Monitoring System.
RM-44A and RM
-44B Containment Exhaust Monitors Question Source:
Bank #
DCPP L 161 Exam Rev 0 (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.4 1.10 Difficulty 3.8
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.1.15 - Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.
Tier # 3 Group # 1 K/A # G2.1.15 Rating 2.7 Question 66 According to OP1
.DC31, Dissemination of Operations Information , h ow is a shift order that discusses operating experience of an incident summary transmitted to an operator?
The shift order shall be ______________
- 1. covered by the Shift Manager at shift bri ef 2. placed in the Shift Foreman Shift Turnover notes for review at shift turnover
- 3. given to each operator for review and signature A. 1 only B. 1 and 2 C. 2 and 3 D. 3 only Proposed Answer:
A. 1 only Explanation:
Shift order book shall contain two types of information: standing orders, and shift orders.
The purpose of the incident summary report is to transmit to the shift operators a concise review of any incident and its cause that the operations manager may deem important. The incident summary shall be reviewed with the crew at the shift briefing.
A. Correct. The incident shall be covered at shift brief.
B. Incorrect because it is only required to be covered by the Shift Manager (SM) at the shift briefing. It is forwarded to the CRA when review is complete, and not kept in the Shift Order book.
Plausible because it may seem reasonable for the SM and SFM to both have for turnover to the crews.
C. Incorrect because it is only required to be covered by the SM at the shift briefing. Plausible because other information is required to be signed for by each operator.
D. Incorrect because it is only required to be covered by the SM at the shift briefing. Plausible because other information is required to be signed for by each operator.
Technical References
- OP1.DC31 References to be provided to applicants during exam: None Learning Objective
- Discuss Operating Experience associated with Operations Department Policies and Administrative Procedures. (46416, 46639)
Question Source:
Bank # (note changes; attach parent)
Modified Bank # New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.10 Difficulty 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.1.32 - Ability to explain and apply system limits and precautions.
Tier # 3 Group # 1 K/A # G2.1.32 Rating 3.8 Question 67 The crew has just entered an Emergency Operating Procedure, such as FR-S.1, Response to Nuclear Power Generation/ATWS.
There is a caution that appears prior to step 1.
Because the caution appears before step 1, the caution ______________.
A. applies only to step 1 B. applies to step 1 and may apply to subsequent procedure steps C. informs the operator of an action that must be taken if conditions are met at any time while in the procedure.
D. informs the operator that there are time critical operator actions that must be taken Proposed Answer:
B. applies to step 1 and may apply to subsequent procedure steps Explanation:
As a generic item, the RO candidate must explain what a caution is and how it applies to a procedure as opposed to a specific system related precaution or note.
The caution in procedures such as FR-S.1, states that the RCPs should not be tripped if reactor power is greater than 5%. This caution, because it appears prior to step 1, applies not only to step 1 but to subsequent steps as well (while above 5% power).
A. Incorrect because it may apply also to subsequent steps. Plausible because notes prior to a step in the procedure typically apply to that step.
B. Correct. The caution can appl y to more than step 1.
C. Incorrect because that type of guidance is located in the foldout page or continuous action step, or possibly a note. Plausible because it may seem like it is giving direction, when (as in the case of the example caution above) it is actually cautioning against taking an action as long as a specific condition is present (>5% power)
. D. Incorrect because steps to meet TCOAs are identified with diamonds, not with cautions. Plausible because it is reasonable to identify time critical items in procedures, however it is accomplished by other means than cautions.
Technical References
- LPERULE, WOG Users Guide References to be provided to applicants during exam: None Learning Objective
- State rules of usage for Emergency and Abnormal Operating Procedures as specified in the DCPP EOP User's Guide and OP1.DC1 0, including the following: (5435
) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.10 Difficulty 3.0
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.1.45 -Ability to identify and interpret diverse indications to validate the response of another indication.
Tier # 3 Group # 1 K/A # G2.1.45 Rating 4.3 Question 68 PK 05-16 , PZR PRESSURE HI/LO, alarms. In accordance with OP1.DC10, Conduct of Operations, to aid in determining if a transient is in progress or there is a pressure control problem, the operator using the diagnostic model would then check the next higher priority parameter, _________.
A. RCS temperature B. Reactor power C. Pressurizer level D. Subcooling Proposed Answer:
C. Pressurizer level Explanation:
The key parameters in order are: power/temperature/pressurizer level/pressure/flow. If pressure is alarming, the operator should check level. If level is constant, the problem is with pressure control, if level is changing, the operator checks temperature etc.
A. Incorrect because pressurizer level should be checked next. Plausible because these parameters are linked by cause and effect and it may be thought that temperature is the next parameter to check because a temperature transient would affect pressure.
B. Incorrect because pressurizer level should be checked next. Plausible because these parameters are linked by cause and effect
, and it may be thought that since power will ultimately affect most parameters downstream, it is a good place to start.
C. Correct. If pressure is changing, level is checked. If it is changing, there is a problem that needs further checking of the parameters. If level is constant, the problem is with pressure control, (PORV, spray, controller, etc)
D. Incorrect because pressurizer level should be checked next. Plausible because Subcooling is a key parameter checked in EOPs for verification of sufficient core cooling.
However, i t is not part of the PTLPF model.
Technical References
- OP1.DC10 References to be provided to applicants during exam: None Learning Objective
- Discuss the STAR
-T diagnostic model. (56221)
Question Source:
Bank # (note changes; attach parent)
Modified Bank # New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.14 Difficulty 2.5
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.2.43 - Knowledge of the process used to track inoperable alarms. Tier # 3 Group # 2 K/A # G2.2.4 3 Rating 3.0 Question 69 Per OP1.DC24, Control of Annunciator System Problems, how often is the Control Operator to initiate and review "main annunciator removed from scan" report? A. Daily B. Weekly C. Monthly D. Quarterly Proposed Answer:
B. Weekly Explanation:
A. Incorrect because the report is initiated and reviewed by the control operator weekly
. Plausible because logs are taken daily and may by practice be reviewed more often
. B. Correct. Per OP1.DC24, the report is initiated and reviewed weekly by the control operator in conjunction with the SM audit of annunciators.
C. Incorrect because the report is initiated and reviewed by the control operator weekly. Plausible because monthly is a reasonable time and the periodicity of many reports.
D. Incorrect because the report is initiated and reviewed by the control operator weekly. Plausible because quarterly is a reasonable time and the periodicity of many reports and typically the number of defeated alarms is low.
Technical References
- OP1.DC24 References to be provided to applicants during exam: None Learning Objective
- Question Source:
Bank # (note changes; attach parent) Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.10 Difficulty 2.5 DCPP L161 Exam Rev 0 Examination Outline Cross-Reference Level RO G2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
Tier # 3 Group # 2 K/A # G2.2.44 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.
According to OP1DC10, Conduct of Oper ations, 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 ope rator repeating back the high level action.
- 2. Specific foldout page parameters and values are not required to be repeated back.
- 3. A copy of the foldout page should be give n to any operator with an assignment.
A. 1 and 2 only B. 1 and 3 only C. 2 and 3 only D. 1 and 2 and 3 Proposed Answer: D. 1 and 2 and 3 Explanation:
A. Incorrect because a copy of the foldout page is given to the operators. Plausible because it may be thought a copy of the foldout page is not needed if the opera tor repeats back the high level action.
B. Incorrect because a detailed repeat back is NOT required. Plausible because the normal practice when given direction is to repeat back the details.
C. Incorrect because a repeat back of the high le vel action is required. Plausible because it may be thought that a repeat b ack is not required since a copy of the foldout page is given to the operator.
D. Correct. All three listed items define how foldout page items are handled for non-TCOA procedure entries (when a brief is held).
Technical References
- OP1.DC10 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 #70 DCPP L111 NRC Exam 11/2012 X New Past NRC Exam X Question History: Last Two NRC Exams No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.41.10 Difficulty 3.3 DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.3.11 - Ability to control radiation releases.
Tier # 3 Group # 3 K/A # G2.3.11 Rating 3.8 Question 71 Which of the following is required prior to initiating a liquid or gaseous radwaste discharge? 1. Sufficient dilution flowrate
- 2. Shift Foreman review and approval of the discharge permit
- 3. Shift Manager review and approval of the discharge permit A. 2 only B. 1 and 2 only C. 3 only D. 1 and 2 and 3 Proposed Answer:
B. 1 and 2 only Explanation:
A. Incorrect because both a gas and liquid discharge require adequate dilution flow prior to initiating the discharge.
Plausible because it may be thought that since review and approva l is needed prior to initiating the release, and there must be an operable rad monitor (or comp actions in place), the review covers the need for dilution flow. B. Correct. Approval by the SFM is required, additionally, there must be adequate flow rate (ASW or Circ water flow) for the discharge to occur.
C. Incorrect because review and approval is required, but not from the SM.
Plausible because the SM is responsible for both units and radwaste release is a common function
. D. Incorrect because SM approval not required.
Plausible because the SM is responsible for both units and a radwaste release is a common function.
Technical References
- OP G-2:V, OP G-1:II References to be provided to applicants during exam: None Learning Objective
- 7431 - Explain the actions to be taken inside the control room for a GDT rupture Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.1 2 Difficulty 2.8
DCPP L161 Exam Rev 0 Examination Outline Cross-Reference Level RO G2.3.15 - Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel moni toring equipment, etc.
Tier # 3 Group # 3 K/A # G2.3.15 Rating 2.9 Question 72 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 because a PED measures beta and gamma. Plausible because gamma is the most plentiful radiation present in a PWR aux building. B. Incorrect because a PED measures both beta and gamma. Plausible because beta is a common radiation present where leakage from radioactive systems occurs.
C. Correct. A PED measures both beta and gamma, but not neutron radiation. D. Incorrect because a PED will not measure neutron radiation. Plausible because a TLD that can measure all 3 types can be issued, and it could be confused with the PED issued at the access point. Technical
References:
Fundamentals - LFC7S References to be provided to applicants during exam:
None Learning Objective: DESCRIBE the characteristics and pr inciples of operation of each of the following types of portable/personal radiation monitoring instruments. (66027) Personal electronic dosimeters (PED) Question Source: Bank #73 DCPP L111 NRC Exam 11/2012 X (note changes; attach parent) Modified Bank #
New Question History: Last Two NRC Exams N Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
CFR: 55.41.11 Difficulty 3.3 DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.4.4 - Ability to recognize abnormal indications for system operating parameters that are entry
-level conditions for emergency and abnormal operating procedures.
Tier # 3 Group # 4 K/A # G2.4.4 Rating 4.5 Question 73 Which of the following conditions would require the crew to initiate emergency boration? 1. Unexplained lowering of reactor power
- 2. One control rod fails to insert following a reactor trip, without SI
- 3. Normal boration using the VCT makeup system is unavailable A. 1 and 2 B. 1 and 3 C. 2 only D. 3 only Proposed Answer:
D. 3 only Explanation:
The crew must recognize that the conditions to enter the abnormal procedure to emergency borate are only met for condition 3.
A. Incorrect because lowering of power does not required emergency boration. Plausible because an unexplained RISE in power requires emergency boration, and it could be the reactivity effects are incorrectly applied or the answer is misread.
B. Incorrect because lowering of power does not required emergency boration. Plausible because emergency boration is required if normal boration is unavailable (3 is an entry condition).
C. Incorrect because emergency boration is not required if only one rod fails to insert on a reactor trip. Plausible because emergency boration is required per E
-0.1 if 2 or more rods fail to insert. D. Correct. This is the only one listed that is an entry condition for emergency boration.
Technical References
- OP AP-6, LPA6, ARPK03
-13, 03-14 References to be provided to applicants during exam: None Learning Objective
- Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event (3478)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.1 0 Difficulty 3.0
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.4.32 - Knowledge of operator response to loss of all annunciators Tier # 3 Group # 4 K/A # G2.4.32 Rating 3.6 Question 74 Both Unit 1 and 2 are at full power, steady state.
An indication of a loss of power to all annunciators on Unit 2 would be PK15-22, MAIN ANNUN SYSTEM TROUBLE on
_________ and the crew should __________.
A. Unit 1; establish continuous surveillance of ALL parameters in the Control Room B. Unit 2; establish continuous surveillance of ALL parameters in the Control Room C. Unit 1; place rod control in MANUAL D. Unit 2; place rod control in MANUAL Proposed Answer:
A. Unit 1; establish continuous surveillance of ALL parameters in the Control Room Explanation:
A. Correct. The loss of annunciators on one unit causes an alarm on the opposite unit. The action is to stop load changes and increase surveillance of all parameters
. B. Incorrect because the alarm will annunciate on Unit 1.
Plausible because how the annunciator trouble alarm functions could be misunderstood, and the action is correct.
C. Incorrect because monitoring all parameters is what is required. Plausible because the alarm location is correct, and the action could be considered to be conservative. However, unlike other situations, such as a loss of DRPI or instrument failure, rods are not required to be placed in MANUAL.
D. Incorrect because the alarm will annunciate on Unit 1, and because the action is incorrect. Plausible because how the annunciator trouble alarm functions could be misunderstood , and placing rods in manual could be considered to be conservative
. Technical References
- AR PK15-22, References to be provided to applicants during exam: None Learning Objective
- Describe controls, indications, and alarms associated with the Annunciat or System. Main Annunciator Power System Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.4 1.10 Difficulty 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level RO G2.4.37 -Knowledge of the lines of authority during implementation of the emergency plan.
Tier # 3 Group # 4 K/A # G2.4.37 Rating 3.0 Question 75 A n ALERT has been declared
. The ____________ should direct the activation of VANS immediately after ______________.
A. Work Control Shift Foreman; notifying the County and State B. Work Control Shift Foreman; event classification C. Shift Man ager; notifying the County and State D. Shift Man ager; event classification Proposed Answer:
D. Shift Man ager; event classification Explanation:
A. Incorrect because the Shift Manager (SM) has command and control and would direct notifying site personnel of the emergency, and because VANS is activated immediately after classification to ensure facilities are manned as soon as possible. Plausible because the WCSFM is present in the control room and aids in EAL classification. It could also be thought that notification of local jurisdictions must be initiated promptly, causing VANS to
be delayed.
B. Incorrect because the Shift Manager (SM) has command and control and would direct notifying site personnel of the emergency. Plausible because VANS is initiated immediately following classification.
C. Incorrect because VANS is activated immediately after classification to ensure facilities are manned as soon as possible. Plausible because the S M will direct the phone talker
, but it is after classification not notification of county/state.
D. Correct. Shift Manager will make classification and then direct the phone talker to activate VANS. Technical References
- LEP-2, EP G-2 References to be provided to applicants during exam: None Learning Objective
- Given an approved Emergency Notification Form, use EP G
-2 to describe the process of making notifications to the County of San Luis Obispo and the State of California, including the timeliness and accuracy requirements.
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.4 1.1 0 Difficulty 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO APE 015/017
- AA2.1 1 Ability to determine and interpret when to jog RCPs during ICC as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow). (CFR 43.5 / 45.13) Tier # 1 Group # 1 K/A # APE 015/017 AA2.1 1 Rating 3.8 Question 76 GIVEN: Unit 1 Core Exit Thermocouples (CETs) are 800°F and rising slowly RVLIS Full Range level is 3 0% and lowering slowly Injection flow is 0 gpm RCS pressure is 1 300 psig and rising slowly 11 and 12 narrow range steam generator levels are 1 0% 13 and 14 narrow range steam generator levels are 3 0% For the current plant conditions and Function Restoration Guideline in effect, will RCPs be started at this time
? NOTE: FR-C.1, Response to Inadequate Core Cooling FR-C.2, Response to Degraded Core Cooling A. Yes, FR-C.2 directs of start of either 1 3 or 1 4 RCP to prevent inadequate core cooling conditions from developing.
B. No, FR-C.1 does not direct a RCP start unless core exit thermocouples exceed 1200
°F, then 11 or 12 RCP would be started to provide temporary core cooling.
C. Yes, FR-C.2 directs of start of any available RCP to prevent inadequate core cooling conditions from developing.
D. No, FR-C.1 does not direct a RCP start unless core exit thermocouples exceed 1200°F, then 13 or 14 RCP would be started to provide temporary core cooling.
Proposed Answer:
D. No, FR-C.1 does not direct a RCP start unless core exit thermocouples exceed 1200°F, then 1 3 or 1 4 RCP would be started to provide temporary core cooling.
. Explanation:
FR-C.1 entry conditions met
- w/no RCPs running, RVLIS level of less than 32%, CSF C is RED (MAGENTA if above 32% RVLIS level)
. A. Incorrect. Conditions are met for FR
-C.1. C.2 does not apply. Reason is plausible as there are steps/cautions designed to prevent conditions from degrading (such as C.2, caution pertaining to not addressing RCS integrity challenge while depressurizing. This caution is designed to prevent stopping on-going core cooling recovery actions, which could then lead to an extreme challenge to core cooling, FR
-C.1, developing).
B. Incorrect. The procedure in use is FR
-C.1. In C.1, RCPs are started, (one at a time, in loops with adequate steam generator level), only if CETs exceed 1200
°F. Plausible as starting 1 or 2 (preferred) is the usual EOP method of restoring RCS pressure control.
C. Incorrect. Conditions are met for entry into FR
-C.1 not C.2. Reason is plausible as there are steps/cautions designed to prevent conditions from degrading (such as C.2, caution pertaining to not addressing RCS integrity challenge while depressurizing. This caution is DCPP L 1 61 Exam Rev 0 designed to prevent stopping on
-going core cooling recovery actions, which could then lead to an extreme challenge to core cooling, FR
-C.1, developing).
D. Correct. Appropriate procedure is C.1 and only RCPs in loops with adequate steam generator level (above 15%) are started
. Technical References
- F-0, FR-C.1 , FR-C.2 References to be provided to applicants during exam:
none Learning Objective
- 5711 - Explain the effects of RCP operation on core exit temperature Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 SRO - requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure applies and coupled with the background knowledge of the conditions necessary to start RCPs and which RCPs would be started (step 19 is in essence a sub
-procedure in C.1)
Difficulty: 3.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO APE022 AA2.01 = Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Whether charging line leak exists Tier # 1 Group # 1 K/A # APE 022 AA2.01 Rating 3.8 Question 77 Unit 1 at full power.
The CO reports the following:
Normal letdown is in service Seal Injection is 8 gpm to each RCP Pressurizer level is lowering slowly Charging flow has risen to 110 gpm VCT level is lowering Containment parameters are unchanged and normal Regen Heat Exchanger outlet temperature has risen Which of the following states the location of the leak in progress and the appropriate Emergency Plan classification, if any, that should be made by the Shift Manager?
A. There is an RCS leak. No E-Plan declaration is required.
B. There is an RCS leak
. The Shift Manager will make an E
-Plan declaration due to exceeding the limits of SU6.1.
C. There is a leak on the charging line. No E
-Plan declaration is required.
D. There is a leak on the charging line. The Shift Manager will make an E
-Plan declaration due to exceeding the limits of SU6.1.
Proposed Answer:
C. There is a leak on the charging line. No E
-Plan declaration is required.
Explanation:
A. Incorrect. The leak is on the charging line, as evidenced by increasing Regen HX outlet temperature, lowering VCT level and stable Pressurizer level
. If it was RCS leakage, the leakage is would be unidentified leakage. Current value is above unidentified but below identified leakage.
B. Incorrect. The leak is on the charging line, as evidenced by increasing Regen HX outlet temperature, lowering VCT level and stable Pressurizer level. If it was RCS leakage, the leakage is would be unidentified leakage and classification would be appropriate
. C. Correct. The leak is on the charging line, as evidenced by increasing Regen HX outlet temperature, lowering VCT level and stable Pressurizer level. Because it is not RCS leakage, (can be isolated and outside containment), no classification is required
.EP App D for category S states the leakage that constitutes the EAL is RCS leakage. According to Tech Spec Bases B3.4.14, for IDENTIFIED RCS Leakage SU6.1 Unidentified or pressure boundary leakage > 10 gpm OR Identified leakage > 25 gpm LEAKAGE, "Identified LEAKAGE does not include LEAKAGE from portions of the Chemical and Volume RCS Leakage SU6.1 Unidentified or pressure boundary leakage > 10 gpm OR Identified leakage > 25 gpm
DCPP L 1 61 Exam Rev 0 Control System outside of containment that can be isolated from the RCS."
D. Incorrect. The leak is on the charging line and not RCS leakage, as evidenced by increasing Regen HX outlet temperature, lowering VCT level and stable Pressurizer level. If it is thought to be RCS leakage, the leakage is would current value is above unidentified and classification would be warranted
. Technical References
- E Plan App D category S, STP
-10C, Tech Spec Bases 3.4.11 References to be provided to applicants during exam:
none Learning Objective
- Question Source:
Bank # X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam
- 76 L121 08/2014 Yes Last Two NRC Exam s Yes Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.27 SRO must determine the location of the leakage is such that it is not RCS leakage. Also, SRO knowledge is that leakage that can be isolated from the RCS is not part of the criteria for making the E-Plan classification. It is important to make a proper classification or non
-classification to ensure the proper response of the ERO. Over response (or "conservative" classification) is not a desired outcome.
Per the SRO guidance of NUREG 1021, ES
-401, the emergency plan classification is Fuel Handling Facilities and Procedures (Emergency Classifications)
Difficulty: 2.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO EPE 029 G2.4.41
- Anticipated Transient Without Scram (ATWS): Knowledge of the emergency action level thresholds and classifications.
Tier # 1 Group # 1 K/A # EPE029 G2.4.41 Rating 4.6 Question 78 Following a failure of the reactor protection system to automatically trip the reactor on low-low steam generator level, the Shift Manager is reviewing Emergency Action Level (EAL) SA2.1. SA2.1 states:
An automatic trip failed to shut down the reactor AND Manual actions taken at the reactor control console successfully shut down the reactor as indicated by reactor power < 5%
Per E-Plan App D Cat S
- Diablo Canyon Power Plant Emergency Plan System Malfunction, in addition to "actuation of the reactor trip switches", which of the actions listed below are considered "manual actions taken at the reactor control console"? 1. Initiation of emergency boration
- 2. Deenergizing 480V Buses 13D and 13E
- 3. Actuation of the turbine trip switch
- 4. Safety Injection A. 1 and 4 onl y B. 1 and 2 only C. 2 and 3 only D. 3 and 4 only Proposed Answer:
C. 2 and 3 only Explanation:
A. Incorrect. Emergency boration and Safety Injection are not listed as an action. App D states: If any of the alternate recovery actions for emergency boration of the RCS listed in EOPs are required to reduce reactor power below 5%, the reactor trips have been unsuccessful.
Safety Injection is not listed, however, because it provides a reactor trip signal (at all power levels) and can be done at the center console, it is a plausible distractor.
B. Incorrect. Deenergizing 13D and E are listed, emergency boration is not.
Plausible, emergency boration is initiated from the control boards.
C. Correct. Appendix D states: The manual actions taken at the reactor control console in th e Control Room to trip the reactor cause rapid control rod insertion. For the purposes of evaluating this EAL, any of the following manual actions are considered successful if reactor power is reduced below 5%:
Actuation of the reactor trip switches.
Deenergization of 480V Buses 13D and 13E (52
-HD-13 and 52-HE-4) at the Control Room vertical board Actuation of the turbine trip switch
. D. Incorrect. Turbine trip switch is a listed action, safety injection is not. Plausible, SI is a reactor trip and initiated from the control boards.
DCPP L 1 61 Exam Rev 0 Technical References
- E-Plan App D Cat S References to be provided to applicants during exam:
none Learning Objective
- Given indications of an event, use EP G
-1 to classify the event with 100% accuracy within 15 minutes. (42285)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.7 Knowledge of the bases for EALs is SRO knowledge.
Difficulty:3.7
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO EPE038 EA2.03
- Ability to determine or interpret which S/G is ruptured as they apply to a SGTR Tier # 1 Group # 1 K/A # EPE038 EA2.03 Rating 4.6 Question 79 GIVEN: The crew is performing FR-S.1, Response to Nuclear Power Generation/ATWS RCS pressure is 1800 psig and lowering RCS temperature is 525
°F and lowering PK11-21, HIGH RADIATION, is LIT Containment pressure is 2.1 psig and rising Steam Generator pressures:
o 11 Steam Generator
- 980 psig and stable o 12 Steam Generator
- 980 psig and stable o 13 Steam Generator
- 980 psig and stable o 14 Steam Generator
- 10 15 psig and rising The operator has reduced AFW flow to 0 gpm to all steam generators Steam Generator narrow range levels are:
o 11 Steam Generator
- 62% and stable o 12 Steam Generator
- 64% and stable o 13 Steam Generator
- 6 4% and stable o 14 Steam Generator
- 70% and rising The crew has successfully opened the reactor trip breakers and is preparing to exit FR
-S.1. Which of the following describes the expected procedure flowpath
? NOTE: E-0, Reactor Trip or Safety Injection E-1, Loss of Reactor or Secondary Coolant E-3, Steam Generator Tube Rupture ECA-3.1, SGTR With Loss Of Reactor Coolant Subcooled Recovery Desired From FR-S.1 go to:
A. E-0 to E-1 B. E-1 C. E-0, to E-3 , to ECA-3.1 D. E-3 , to ECA-3.1 Proposed Answer:
C. E-0, to E-3, to ECA-3.1. Explanation:
The 14 steam generator is ruptured (level rising uncontrollably. The crew must first return to E
-0, perform the diagnostic steps and will got to E
-3 based on the uncontrollable level increase
. Its pressure is stable due to rising toward the 10% steam dump setpoint.
A. Incorrect. The proper procedure is E
-0 but 14 Steam Generator is ruptured and the proper transition from E
-0 is E-3. Plausible
- LOCA conditions met and may not recognize DCPP L 1 61 Exam Rev 0 ruptured SG B. Incorrect. The proper procedure is E
-0. This is plausible because if there were another red or magenta CSF a transition to E
-0 would not be made, also, the absence of a steamline rad monitor may cause misdiagnosis of even ts. C. Correct. This is the proper response if there i s a ruptured steam generator. Rising level (coupled with the increased pressure) is the indication the 14 SG is ruptured. The rising containment pressure and containment radiation alarm (LOCA) will cause the crew to have to go to ECA
-3.1 D. Incorrect. The proper procedure is E
-0. This is plausible because if there were another red or magenta CSF a transition to E
-0 would not be made, Technical References
- FR-S.1, E-0, E-3 References to be provided to applicants during exam:
none Learning Objective
- 7336 - State contents of foldout page Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 SRO must recognize the symptoms of a post trip tube rupture and then determine the correct action to take.
Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO W/E11 G2.4.6 - Loss of Emergency Coolant Recirc.
- Knowledge of EOP mitigation strategies.
Tier # 1 Group # 1 K/A # E 11 G2.4.6 Rating 4.7 Question 80 EOP ECA-1.1, Loss of Emergency Coolant Recirculation, is in effect.
Containment pressure rises and the Containment Critical Safety Function turns MAGENTA.
The Shift Foreman will:
A. Transition to FR
-Z.1, Response to High Containment Pressure, but direct the operators to operate Containment Spray Pumps as described by ECA
-1.1. B. Transition to FR
-Z.1, Response to High Containment Pressure, and direct the operators to operate Containment Spray Pumps as described by FR
-Z.1. C. Remain in ECA
-1.1, and direct the operator to verify all available CFCUs are running
. D. Remain in ECA-1.1, and direct the operator to start both Containment Spray Pumps to clear the Containment MAGENTA Critical Safety Function
. Proposed Answer:
A. Transition to FR
-Z.1, Response to High Containment Pressure, but direct the operators to operate Containment Spray Pumps as described by ECA-1.1. Explanation:
A. Correct. A transition is appropriate.
FR-Z.1 step 3 b. states: Operate Containment Spray as directed by EOP ECA 1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, Step 6. B. Incorrect. Plausible
- A transition is appropriate, however, the overriding concern to prolong RWST inventory overrides running spray pumps as directed by Z.1
. C. Incorrect. CFCUs will cool containment, spray pumps lower pressure. The MAGENTA path must be addressed. Spray pumps are operated to lower pressure but CFCUs are operated to conserve RWST inventory
. Plausible
- ECA-1.1 directs the operation of both the CFCUs and the spray pumps.
D. Incorrect. While in ECA
-1.1, the CS pumps are operated based on the number of CFCUs and containment pressure. A transition to address the MAGENTA path is appropriate however, Z.1 will inform the operator to operate the pumps in accordance with ECA
-1.1. Plausible
- ECA-1.1 directs the operation of both the CFCUs and the spray pumps.
Technical References
- EC A-1.1 and background References to be provided to applicants during exam:
none Learning Objective
- 42460 - Explain basis of emergency steps of ECA
-1.1 Question Source:
Bank # 80 DCPP NRC exam 02/2009 X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam Yes Last Two NRC Exam s No DCPP L 1 61 Exam Rev 0 Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 SRO must be knowledgeable of the strategy of ECA
-1.1, which is to conserve RWST level and the interrelationship between Z.1 and the ECA. Difficulty: 3.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO APE077 G2.4.30 Generator Voltage and Electric Grid Disturbances: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
Tier # 1 Group # 1 K/A # APE077 G2.4.30 Rating 4.1 Question 81 Unit 1 and Unit 2 are at full power.
Due to grid instability, the Shift Manager is in contact with the Grid Control Center (GCC) about an emergency backdown order
. Per Operations Policy, B
-1, Communications With Generation and Transmission Organizations , which of the following questions should be asked by the Shift Manager?
- 1. How many megawatts per unit?
- 2. How many megawatts need to be shed
? 3. How quickly
? A. 1 and 3 B. 1 and 2 C. 2 only D. 2 and 3 Proposed Answer:
D. 2 and 3 Explanation:
A. Incorrect. The information that must be known is: 1. Is it an emergency, 2. How many total MW and 3. How quickly (time). The SM decides how much per unit and ramp rate (max is 100 MW/unit)
. B. Incorrect. 3 correct, 2 is not. Plausible
- that it is predetermined how much each unit will shed. SM decides how much per unit
. C. Incorrect. In addition to the amount of the load reduction, how quickly must also be asked.
D. Correct. Of the questions listed, how quickly and the amount needs to be asked.
Technical References
- Operations Policy B
-1 References to be provided to applicants during exam:
none Learning Objective
- 3654 - Identify and discuss Operations Department policy statements Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.5 Ops Policy B
-1 is implemented by the Shift Manager. The knowledge of the information that should be obtained by the SM when notified of the need to perform an emergency backdown
. Difficulty: 2.7
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO APE005 G2.4.47 - Inoperable/Stuck Control Rod
- Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. Tier # 1 Group # 2 K/A # APE005 G2.4.47 Rating 4.2 Question 82 GIVEN: A plant shutdown is being performed in accordance with OP AP
-12D, Plant Shutdown With Immovable Control Rods Initially AFD at 100% power is 0 Power reduction from 100%, at a rate of approximately 1.0%/minute , has just commenced If AFD changing at a rate of 1.0% per minute, in ________ minutes, the Shift Foreman should direct the crew to ______________.
A. 1 4 to 1 6; stabilize the plant and restore AFD to within the AFD band B. 1 4 to 1 6; trip the reactor and go to E
-0, Reactor Trip or Safety Injection C. 1 9 to 21; stabilize the plant and restore AFD to within the Acceptable Operation region D. 1 9 to 21; trip the reactor and go to E
-0, Reactor Trip or Safety Injection Proposed Answer:
B. 14 to 16; trip the reactor and go t o E-0, Reactor Trip or Safety Injection Explanation:The core outlet temperature decreases significantly, and the core inlet temperature increases slightly. This increases the moderator density in the top of the core relative to the bottom, and causes better neutron thermalization in the top of the core. Also, because the water is more dense in the bottom of the core, more boron atoms are added to the bottom of the core, reducing the thermal utilization factor in the bottom of the core. Both of these effects cause the power to shift toward the top of the core.
A. Incorrect. AFD will move positive as more flux is produced in the top of txhe core.
In about 15 minutes, AFD will be at approximately
+1 5. The 85% power limit is 14.5%, however, the action is to trip. OP AP-12D states that actions should not be taken to stabilize the plant and once the downpower starts it should be completed.
Plausible to think that stabilizing the plant and possibly attempting to restore AFD is preferable to a reactor trip. B. Correct. In 15 minutes, AFD will be at approximately +15 and outside the "doghouse". Per the procedure, the proper action is to trip and go to E
-0. C. Incorrect. In approximately 20 minutes, AFD outside the "doghouse" IF its believed AFD will go negative. Also, plausible to think that actions should be taken to stabilize the plant
. D. Incorrect. Plausible because in approximately 20 minutes, AFD will be outside the "doghouse", if the incorrect assumption is made about how it will respond to the downpower. The action is correct. Technical References
- R 23-1F-1, OP AP-12D, GFES Reactor Theory chapter 8, LPA-12 References to be provided to applicants during exam:
R 23-1F-1 Learning Objective
- 59900 - Demonstrate the ability to shut down the reactor with control rods immovabl e. Given an abnormal condition, summarize the major actions of OP AP-12 to mitigate an event in DCPP L 1 61 Exam Rev 0 progress. (3477M)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.6 The SRO must assess the plant conditions, the rate of decrease and perform the diagnostic steps to determine the action to take either when outside the acceptable range (doghouse). Normally, if outside the band, action would be lower power while attempting to stabilize AFD. As the actions are not immediate actions, this is SRO knowledge.
The SRO must assess and then select the proper procedure (in this case, E-0)
Difficulty: 3.7
DC PP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO APE024 AA2.04
- Ability to determine and interpret the following as they apply to the Emergency Boration:
Availability of BWST Tier # 1 Group # 2 K/A # APE024 AA2.04 Rating 4.2 Question 83 GIVEN: Unit 1 tripped from 100% power RCS Tave is 547°F RCS boron concentration is 300 ppm 2 Control Rods are stuck out Boric Acid Storage Tank (BAST) #1 is in service and level is 96%
Boric Acid Storage Tank (BAST) #2 level is 96%
RWST level is 96%
The crew initiates emergency boration through the preferred flow path in accordance with OP AP-6, Emergency Boration.
When the required boration is complete, what ECG or Technical Specification action, if any, should be taken?
NOTE: assume a 1 ppm change requires 9 gallons of 4% boric acid A. Makeup to the BAST but no Technical Specification or ECG ACTION is required.
B. Makeup to the RWST but no Technical Specification or ECG ACTION is required.
C. Within one hour, restore the RWST to OPERABLE status
. D. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, restore the Boric Acid Storage system to OPERABLE status.
Proposed Answer:
D. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, restore the Boric Acid Storage system to OPERABLE status. Explanation:
The SRO must determine the following:
Amount of ppm change
- AP-6 states: Borate 400 ppm per stuck rod OR borate to 2000 - 2400 ppm boron concentration, whichever is less. (for current plant condition, 300 ppm + 800 ppm = 1100 final concentration).
What the "normal source" is in OP AP makeup using the CVCS makeup controller.
Thumb rule (GFES) for a ppm change - 9 gallons/ppm for BAST. (This correlates to approximately 27 gallons/ppm if using the RWST
- about a 3/1 ratio due to differences in boron concentration)
Amount of inventory required. #stuck rods x required ppm change x thumb rule = 2 x 400 ppm x
9 gals/ppm = 7200 gallons (21600 is assuming RWST)
Change in BAST level = (95% level) 14964 gallons
- 7200 = 7764 gallons remaining which is less than the requirement in ECG 8.9.
A. Incorrect. The combined levels of the BAST is 14964 gallons. If the conversion of ppm to gallons is not done and 800 is used, then the assumed level would be 14114 gallons. Still above the requirements of ECG 8.9 B. Incorrect. If the thumb rule for the BAST is used, then 7200 gallons would be the calculated change, which is still above the required 455300 gallons of LCO 3.5.4 DC PP L 1 61 Exam Rev 0 (463534 - 7200 = 456334 gallons)
C. Incorrect. If its assumed the RWST are the alternate source and the math correctly calculated, this would be the answer. 463354
- 21600 = 441934 gallons D. Correct. The required level of 14042 gallons will not be met and the LCO action must be taken to restore the BAST to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Technical References
- ECG 8.9, Tech Spec 3.5.4, STP C
-20, OP AP-6. References to be provided to applicants during exam: ECG 8.9 (LCO only), Tech Spec 3.5.4, BAST and RWST level tables from C
-20 Learning Objective
- Discuss significant Technical Specifications and Equipment Control Guidelines associated with the Reactor Makeup Control System.
- Apply TS 3.1 & 3.9 Technical Specification LCOs. (9697A/I)
- Apply the requirements of System 8 ECGs. (66041)
Question Source:
Bank # X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam 04/16 DCPP #82 Yes Last Two NRC Exam s Yes Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.2 SRO knowledge of STP C
-20 and bases of ECG and Tech Specs.
Difficulty: 3.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO APE033 AA2.10 Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear instrumentation: Tech Spec limits if both intermediate range channels have failed.
Tier # 1 Group # 2 K/A # APE033 AA2.10 Rating 3.8 Question 84 Unit 1 is at 6% power during a plant startup.
If both Intermediate Range Channels become inoperable, in accordance with Technical Specifications, 1. Which of the following is required?
- 2. What is the accident the Intermediate Range is designed to protect
?
A. 1. Enter TS 3.0.3
. 2. Protection against a reactivity excursion caused by an uncontrolled RCCA withdrawal.
B. 1. Reduce power to less than P
-6. 2. Protection against a reactivity excursion caused by a Main Steam Line Break.
C. 1. Enter TS 3.0.3.
- 2. Protection against a reactivity excursion caused by a Main Steam Line Break.
D. 1. Reduce power to less than P
-6. 2. Protection against a reactivity excursion caused by an uncontrolled RCCA withdrawal
. Answer: D. 1. Reduce power to less than P
-6. 2. Protection against a reactivity excursion cause d by an uncontrolled RCCA withdrawal.
Explanation:
A. Incorrect because TS 3.3.1 condition G addresses loss of both IR channels
. Entry into LCO 3.0.3 is not appropriate. Plausible because it may be thought that TS 3.3.1 covers loss of one IR channel, but not both. Also plausible because the IR provides protection from an uncontrolled RCCA withdrawal.
B. Incorrect because the IR provides protection from an uncontrolled RCCA withdrawal, not a MSLB accident
. Plausible because lowering power to less than P
-6 is the correct action.
C. Incorrect because TS 3.3.1 condition G addresses loss of both IR channels. Entry into LCO 3.0.3 is not appropriate. Plausible because it may be thought that TS 3.3.1 covers loss of one IR channel, but not both.
Also, a MSLB is worst at low power, adding positive reactivity. Because IR are only required below 10%, plausible that they provide protection from this accident.
D. Correct. TS 3.3.1 condition G addresses loss of both IRs. When below P
-6, the SR NIS provides protection against an uncontrolled RCCA withdrawal. When above P
-10 the PR NIS provides this protection. When between these power levels (including the stated 6%
power) the IR provides protection against this accident. TS 3.3.1 action G.2 directs with 2 IR channels inoperable to lower power to less than P
-6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Technical References
- LCO 3.3.1, B3.3 bases, References to be provided to applicants during exam:
None Learning Objective
- 9694C -Apply TS 3.3 Technical Specification bases.
9697C - Apply TS 3.3 Technical Specification LCOs.
Question Source:
Bank #
DCPP L 1 61 Exam Rev 0 (note changes; attach parent)
Modified Bank:
New X Question History:
Past NRC Exam No Las t Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 / 45.13 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO APE076 2.4.31
- High Reactor Coolant Activity: Knowledge of annunciator alarms, indications, or response procedures.
Tier # 1 Group # 2 K/A # APE 076 G Rating 4.1 Question 85 GIVEN: Unit 1 is at full power, operating with a recently identified fuel defect causing elevated RCS activity PK11-06, SJAE HI-RAD , is in high alarm The Shift Foreman is implementing OP O
-4, Primary to Secondary Steam Generator Tube Leak Detection RM-15 and 15R, condenser air ejector discharge radiation monitors, are confirmed to be at the high alarm setpoint and exceeding the action level for requiring a shutdown to MODE 3 Per OP O-4, 1. when evaluating the rise in RM
-15 count rate, the Shift Foreman should realize that due to the elevated RCS activity, the alarm setpoint probably corresponds to a _______________ leak rate than 30 gpd
. 2.once the power reduction begins, the unit must be in MODE 3 in no more than _____________ hours. A. lower; 3 B. lower; 6 C. higher; 3 D. higher; 6 Proposed Answer: A. lower; 3 Explanation:
A. Correct. The alarm will occur at a lower leak rate due to noble gases.
OP O-4 states: The high alarm setpoint of RM
-15 and RM-15R is set to the activity level that should be seen by the detector if the previously calculated RCS activity still exists and a total of 30 gpd is leaking from the RCS to the steam generators.
Once the power reduction begins, power must be less than 50% in one hour and MODE 3 in the following 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> total)
. B. Incorrect. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is the technical specification time to be in MODE 3
. C. Incorrect. Plausible to believe the higher activity makes the alarm less conservative if its not understood how the alarm setpoint is set, time is correct.
D. Incorrect. Plausible to believe the alarm is less conservative if its thought a higher activity corresponds to higher leak rate calculation and time corresponds to technical specification time Technical References
- OP O-4, PK11-06,Table T
-11C-2 (SJAE RM calibration data), LCO 3.4.13 References to be provided to applicants during exam:
none Learning Objective
- 5747 -Explain the Radiation Monitoring system response to S/G tube leakage Question Source:
Bank #
DCPP L 1 61 Exam Rev 0 (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.4 From O-4,3.1 The SFM is responsible for evaluating the RM
-15 or RM-15R rise in count rate and grab sample results for steam generator leakage changes.
Difficulty: 3.3
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO 004 G2.1.20
- CVCS: Ability to interpret and execute procedure steps.
Tier # 2 Group # 1 K/A # 004 G2.1.20 Rating Question 86 GIVEN: Unit 2 RCS temperature is 225°F RHR is in service The crew is performing OP AP
-18, Letdown Line Failure Letdown has been isolated The operator reports Pressurizer level and pressure are lowering slowly.
What action should be taken by the Shift Foreman?
A. Go to either OP AP
-1, Excessive RCS Leakage or OP AP-24, Shutdown LO CA. B. Go to either OP AP SD-0, Loss of/or Inadequate Decay Heat Removal or OP AP
-24, Shutdown LOCA.
C. Go to OP AP
-24, Shutdown LOCA, only
. D. Go to OP AP
-1, Excessive RCS Leakage, only. Proposed Answer:
C. Go to OP AP
-24, Shutdown LOCA, only.
Explanation:
A. Incorrect. The RNO for the step in AP
-18 states: If not on RHR, GO TO OP AP-1, RCS LEAKAGE OR AP-24, SHUTDOWN LOCA.
Because RHR is in service and plant is in MODE 4. AP
-1 is not applicable in this MODE.
Plausible
- both are listed in the RNO, if plant conditions B. Incorrect. If RHR is in service, the RNO states, GO TO Either OP AP SD-0, LOSS OF/OR INADEQUATE DECAY HEAT REMOVAL, OR OP AP-24, SHUTDOWN LOCA. However, SD
-0 is only applicable in MODES 5 and 6
. Plausible as this is a choice but not for the given plant conditions.
C. Correct. If RHR is in service, the action is to GO TO Either OP AP SD-0, LOSS OF/OR INADEQUATE DECAY HEAT REMOVAL, OR OP AP-24, SHUTDOWN LOCA established. Because the plant in in MODE 4, SD
-0 does not apply, AP
-24 applies.
D. Incorrect. If its interpreted that the alignment only allows for seals only, a plant shutdown is required
.AP-1 does not apply in MODE 4 on RHR.
Plausible
- this could be correct if not on RHR.
Technical References
- OP AP-1, AP SD-0, AP-24 References to be provided to applicants during exam:
none Learning Objective
- Given an abnormal condition, summarize the major actions of OP AP
-18 to mitigate an event in progress. (3477)
Question Source:
Bank # DCPP Bank P-31764 X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam No Last Two NRC Exam s No DCPP L 1 61 Exam Rev 0 Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 The SRO must be able to determine (interpret) the procedure that addresses the continuing RCS leak for the current mode. The procedure gives the SRO 2 choices based on RHR status.
However while the procedure lists 2, only one of the two will apply for a given set of plant conditions (interpet and execute). The entry for AP's must be known. These abnormal procedures are not "major" abnormal procedures. The applicable modes for SD
-0 and AP-24 is not RO required knowledge.
Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO 013 A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Loss of instrument bus Tier # 2 Group # 1 K/A # 013 A2.04 Rating 4.2 Question 87 Unit 1 is at full power An earthquake causes the following events to occur:
Loss of PY
-11 Bus differential fault on 4 kV Vital Bus G A large LOCA Containment pressure rises to 24 psig E-0, Appendix E, ESF Auto Actions, Secondary And Auxiliaries Status, is complete Which of the following actions will be taken by the Shift Foreman as a transition from E
-0, Reactor Trip or Safety Injection is made?
NOTE: E Loss of Reactor or Secondary Coolant FR-Z.1 - Response to High Containment Pressure A. Prior to entering E
-1, address the Containment Integrity Red path. B. Go to E-1; the one available Containment Spray pump did not start automatically but was started by the operator while performing E
-0, Appendix E
. C. Prior to entering E
-1, address the Containment Integrity Magenta path.
D. Go to E-1; one Containment Spray pump automatically started when Containment pressure exceeded the High
-High setpoint and the second Containment Spray pump started by the operator while performing E
-0, Appendix E
. Proposed Answer:
B. Go to E-1; the one available Containment Spray pump did not start automatically but was started by the operator while performing E-0, Appendix E
. Explanation:
A. Incorrect.
The only Red path for Containment Integrity is Containment pressure greater than 47 psig. Loss of Bus G will render Containment Spray pump 1
-1 unavailable. Loss of PY-11 results in the failure of the 1
-2 Containment Spray pump to automatically start, due to loss of power to Train A of SSPS. However, it will be manually started while the operator is performing Appendix E
. B. Correct. The 1-2 Containment Spray pump will not start automatically but will be started by the operator while performing Appendix E. As a result, at the transition from E
-0, one spray train will be operating and the status tree will be Yellow
. C. Incorrect. With one pump started by the operator, the status tree will be Yellow and not require attention. D. Incorrect. The loss of Bus G renders the 1
-1 Containment Spray (Train B) pump DCPP L 1 61 Exam Rev 0 unavailable. If the candidate does not realize the bus differential prevents the auto transfer, then they will not recognize the loss of the Containment Spray pump Technical References
- OIM J-1, F-0, E-0 appendix E References to be provided to applicants during exam:
none Learning Objective
- 6764 - Describe what procedure or procedure set would be used in an emergency event, based on plant mode/conditions Question Sour ce: Bank #89 DCPP NRC Exam L091 X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam Yes Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 The SRO must know what the appropriate action to take with a loss of a 4 kV bus and determine the intial action taken and what occurs when PY
-1 1 is lost. This allows the SRO to make the proper procedure transition
.
Difficulty: 3.7
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO 059 A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Overfeeding event Tier # 2 Group # 1 K/A # 059 A2.03 Rating 3.1 Question 88 GIVEN: Main Feed Reg Valve,1
-FCV-520, failed open and the reactor tripped from full power 1-2 Steam Generator is isolated 1-2 Steam Generator narrow range level is 94% and rising slowly 1-2 Steam Generator pressure is 1130 psig Pressure in steam generators 1
-1, 1-3 and 1-4 is approximately 900 psig The crew is performing the actions of EOP FR-H.3, Response To Steam Generator High Level
. Which of the following actions should be taken by the Shift Foreman?
Direct the operator to:
A. leave the steam generator isolated and do not dump steam until it is has been determined there is no water in the steamline.
B. leave the steam generator isolated and do not dump steam until sampling is complete to confirm the steam generator is not ruptured.
C. lower level by fully opening the 10% Steam Dump and dumping steam at the maximum rate until pressure is approximately 900 psig or narrow range level is less than 65%.
D. lower level by opening the 10% Steam Dump and dumping steam at a rate of less than 12 0 psig/minute until pressure is approximately 900 psig or narrow range level is less than 65%.
Proposed Answer:
A. leave the steam generator isolated and do not dump steam until it is has been determined there is no water in the steamline.
Explanation:
EOP FR-H.3 Rev. 16, Step 1
CAUTION: If S/G NR Level has increased to GREATER THAN 92% an evaluation should be made for S/G overfill considerations. Steam should not be released from any S/G with level GREATER THAN 92% prior to overfill evaluation.
A. Correct. Per FR-H.3 Background Document
- If affected SG level has increased above the narrow range, the operator cannot be sure if the SG is filled to the steamline. The objective of the status evaluation is to determine if water is in the steamline. Just lowering affected SG level into the narrow range does not ensure that water does not remain in the affected SG steamline. An evaluation of the steamline conditions should occur prior to releasing steam from any SG with level above 92% to prevent potential damage to piping, valves, or turbines. B. Incorrect. Plausible
- There is a check in the procedure (for radiation) to determine if the DCPP L 1 61 Exam Rev 0 steam generator is ruptured and it is plausible to think that checking to see if the water is radioactive prior to a steam release
. C. Incorrect. Plausible
- Dumping steam would lower level. The max rate is used in EOP procedures such as E
-3 if the condenser is not available
. D. Incorrect. Plausible
- Dumping steam would lower level. Dumping steam would lower level. The 125 psig rate is used in EOP procedures such as E
-3 if the condenser is available
. Technical References
- FR-H.3 and background, E
-3 step 9 and 10 References to be provided to applicants during exam:
none Learning Objective
- Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysi s X 10CFR Part 55 Content:
55.43.5 The SRO must assess the plant conditions and determine that overfill is a greater concern than the high level or the steam generator pressure. FR
-H.3 is an Yellow path CSF and not a major plant procedure that an RO is required to know. The assessment of plant status is an SRO function.
Difficulty: 3.8
DCPP L 1 61 Exa m Rev 0 Examination Outline Cross
-Reference Level S RO 061 G2.1.25
- AFW: Ability to interpret reference materials, such as graphs, curves, tables, etc.
Tier # 2 Group # 1 K/A # 061 G2.1.25 Rating 4.2 Question 89 GIVEN: Unit 2 has tripped following a loss condenser vacuum At 1300, the crew entered E
-0.2, Natural Circulation Cooldown Initial CST level is 48% At 1700, o CST level is 24%
o required boron concentration is established and the crew is ready to initiate an RCS cooldown from 547°F to MODE 4 Assuming the cooldown could be performed at the maximum rate allowed by E
-0.2, which of the following describes the action that should be taken by the Shift Foreman?
A. Establish approximately a 25°F/hour cooldown and remain in E
-0.2. B. The cooldown rate will have to be greater than the maximum allowed rate of 25°F/hour, transition to E
-0.3, Natural Circulation Cooldown With Steam Void In Vessel (With RVLIS). C. Establish approximately a 50°F/hour cooldown and remain in E
-0.2. D. The cooldown rate will have to be greater than the maximum allowed rate of 50°F/hour, transition to E
-0.3, Natural Circulation Cooldown With Steam Void In Vessel (With RVLIS). Proposed Answer:
D. The cooldown rate will have to be greater than the maximum allowed rate of 50°F/hour, transition to E
-0.3, Natural Circulation Cooldown With Steam Void In Vessel (With RVLIS)
. Explanation:
A. Incorrect. Plausible
- This is the Unit 1 rate
. If the curves are not used or misread and wrong unit number applied, this would be correct.
B. Incorrect. Plausible
- If the unit 1 rate is used, and the proper curve, this would be correct
. C. Incorrect. Plausible
- Using the wrong curve could make it appear there is adequate volume. D. Correct. At the unit 2 rate, it will take approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, there is not adequate CST level (requires greater than 50%)
. Technical References
- IF-2, IF-4 and IF-5 , E-0.2 Unit 1 and E
-0.2 Unit 2 References to be provided to applicants during exam:
IF-2, IF-4 and IF-5 Learning Objective
- Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event. (3552)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No DCPP L 1 61 Exa m Rev 0 Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 The SRO must interpret the curves and apply the proper cooldown rate then determine the proper procedural guidance.
Difficulty: 3.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO 006 A2.13
- Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Inadvertent SIS actuation Tier # 2 Group # 1 K/A # 006 A2.13 Rating 4.2 Question 90 The reactor trips due to a spurious Safety Injection actuation.
Procedurally , when will the crew terminate charging injection and establish normal charging?
A. E-1, Loss of Reactor or Secondary Coolant to sequentially secure SI pumps while checking for indications of a RCS break.
B. E-1.1, Safety Injection Termination to allow for a check of the Critical Safety Functions prior to stopping ECCS pumps.
C. E-0, Reactor Trip or Safety Injection to minimize the possibility of Pressurizer overfill
. D. 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
. Proposed Answer:
C. E-0, Reactor Trip or Safety Injection to minimize the possibility of Pressurizer overfill
. Explanation:
A. Incorrect
- A transition to E
-1 is not appropriate. Plausible as other EOPs, such as E
-2 transition to E
-1 and not E
-1.1. E-1 has a step for checking if SI should be terminated. Also, if its not known that E
-0 also checks, then going to E
-1 would seem appropriate.
B. Incorrect
- Plausible
- The status trees have not been checked until the transition, however this is not a reason to wait until E
-1.1 to realign charging.
C. 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.
D. Incorrect
- Plausible
- While normal charging will be placed in service i n E-0, E-0.1 is only entered if SI did not actuate.
Technical References
- E-0, Reactor Trip or Safety Injection, 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 # DCPP NRC 11/2012, #86 X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam Yes Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 Difficulty: 2.8
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO 029 G2.2.25 - Containment Purge: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
Tier # 2 Group # 2 K/A # 029 G2.2.25 Rating 4.2 Question 91 Per ECG 23.3, Containment Ventilation System, which of the following states the amount of time the containment purge supply and/or exhaust isolation valves may be open and the reason for the limit?
________________hours to ________________.
A. 200; prevent exceeding the NPDES permit B. 200; minimize the probability of a LOCA occurring while the valves are open C. 90; prevent exceeding the NPDES permit D. 90; minimize the probability of a LOCA occurring while the valves are open Proposed Answer:
B. 200; minimize the probability of a LOCA occurring while the valves are open Explanation:
A. Incorrect. Plausible because the purge could be considered a discharge (and impact the NPDES), it does not and is not the reason stated in the bases.
B. Correct. The ECG states the valves can be open for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per calendar year. The bases states the purging time restriction is meant to minimize the probability of a LOCA while conducting purging operations and thereby limit offsite boundary doses.
C. Incorrect. Plausible 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> is stated in the Bases as the time the NRC had wanted as a time limit
. D. Incorrect. Plausible 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> is stated in the Bases as the time the NRC had wanted as a time limit
. Technical References
- ECG 23.3 References to be provided to applicants during exam:
none Learning Objective
- 66064 -Apply the requirements of System 23 ECGs.
Question Source:
Bank # 98 DCPP NRC Exam L091, 07/2011 X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam Yes Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.2 The SRO must know the bases for the limitation of the ECG (which is a license document)
.
Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO 041 A2.02
- Ability to (a) predict the impacts of the following malfunctions or operations on the SDS system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Steam valve stuck open Tier # 2 Group # 2 K/A # 04 1 A2.02 Rating 3.9 Question 92 GIVEN: E-3, Steam Generator Tube Rupture is being performed 13 Steam Generator is ruptured RCS pressure is 1300 psig and stable Pressurizer level is 35%
and stable The crew has just completed terminating SI and placed letdown in service A 10% steam dump valve on 12 Steam Generator fails open causing pressure in that steam generator to begin decreasing uncontrollably.
What action will be taken by the Shift Foreman?
A. Go to E-2, Faulted Steam Generator Isolation.
B. Go to ECA-3.1, SGTR With Loss Of Reactor Coolant, Subcooled Recovery Desired.
C. Direct the operator to reinitiate SI and return to E
-0, Reactor Trip or Safety Injection.
D. Continue to perform the actions of E
-3 unless pressurizer level or subcooling cannot be maintained, then reinitiate ECCS flow and go to ECA
-3.1, SGTR With Loss Of Reactor Coolant, Subcooled Recovery Desired.
Proposed Answer:
A. Go to E-2, Faulted Steam Generator Isolation.
Explanation:
A. Correct. Per the foldout page, a transition is made to E
-2 whenever there is indication of a faulted steam generator, even if SI has been terminated.
B. Incorrect. Plausible
- this would be correct if the faulted steam generator was the ruptured steam generator.
C. Incorrect. SI would not be reinitiated. Plausible
- because SI is terminated, possible to think that the appropriate action is to go back and perform the diagnosis steps of E
-0. D. Incorrect. Despite SI not injecting, it is not correct to wait for a loss of subcooling or pressurizer level to reinitiate flow. Plausible
- this is the action if either condition occurs after SI is terminated.
Technical References
- EOP E-3 References to be provided to applicants during exam:
none Learning Objective
- Question Source:
Bank #93 DCPP NRC 02/2009 X (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam Yes Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental
DCPP L 1 61 Exam Rev 0 Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 SRO must apply the knowledge of proper use of EOP transitions.
Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO 075 A2.02
- Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Loss of circulating water pumps Tier # 2 Group # 2 K/A # 075 A2.02 Rating 2.7 Question 93 A turbine load increase is in progress with all systems operating normally in automatic control when the following occurs:
Turbine Load is 800 MWE Condenser Quadrant DP's:
North East quadrant
- 9.5 psid and rising South East quadrant
- 9 psid and rising North West quadrant
- 1 2.6 psid and rising rapidly South West quadrant
- 1 2.8 psid and rising rapidly Condenser Pressure PI
-44 reads 5.5" Hg Abs. Condenser Pressure Recorder PR
-11A and B both show condenser pressure rising slowly PK10-11 COND PRESS/LEVEL is in ALARM PK13-04, CONDENSER DELTA P HI PPC is in ALARM Which of the following describes the action that should be taken by the Shift Foreman?
Per OP AP-7, Degraded Condenser ___________ direct the operator to _____________.
A. section A (Loss of Condenser Vacuum); reduce load as necessary to restore condenser pressure to within operating limits B. section A (Loss of Condenser Vacuum), trip the reactor C. section B (Condenser Fouling); stop the 1-1 Circ Water Pump (west quadrant
) D. section B (Condenser Fouling); reduce load to less than 25% and then shutdown both circ water pumps Proposed Answer:
C. section B (Condenser Fouling); stop the 1
-1 Circ Water Pump (west quadrant) Explanation:
A. Incorrect. Plausible
- condenser pressure is rising. However, the trip setpoint for 75% is about 8.5. With pressure at 6.5 and rising, the immediate concern is the rising pressure in the west quadrants. Currently vacuum is in the operating range.
B. Incorrect. A trip due to a loss of vacuum is not called for at this time. Plausible
- if Setpoint is not known, (currently approximately 8.5 inches
). C. Correct. Per Section B, states: Remove affexceeds 13 PSID. The west quadrant (pump 1
-1) is approaching 13 psid and the pump must be stopped.
D. Incorrect. Plausible
- All quadrants are affected and if the west quadrants were not so close to 13 psid, the action would be to reduce power and stop both pumps.
Technical References
- AR PK10-11 , AR PK13-04, OP-AP-7 DCPP L 1 61 Exam Rev 0 References to be provided to applicants during exam:
none Learning Objective
- Given an abnormal condition, summarize the major actions of the abnorma l operating procedure to mitigate an event in progress. (3477G)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
83 DCPP NRC Exam (L 081) 1/2010 New Question History:
Past NRC Exam Yes Last Two NRC Exam s No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.5 The SRO must determine the most important event in progress is the rising pressure in the west quadrant vice the rising condenser pressure. Then determine the appropriate action per the applicable section of OP AP
-7.
Difficulty: 3.7
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO G2.1.40 - Knowledge of refueling administrative requirements Tier # 3 Group # 1 K/A # G2.1.40 Rating 3.9 Question 94 In accordance with OP B
-8DS1, Core Unloading, the Refueling SRO has the responsibili ty for which of the following
? 1. Maintaining the record of fuel movement in accordance with the Fuel Movement Tracking Sheet
- 2. Safe and orderly evacuation of the refueling crew in the event of a high radiation alarm at a refueling station
- 3. Providing technical guidance and trending source range count rates.
A. 1 only B. 1 and 3 C. 2 only D. 2 and 3 Proposed Answer:
C. 2 only Explanation:
Per OP OP B-8DS1 (Core Unload), the following are the responsibility of the refueling SRO:
Direct supervision of CORE ALTERATION activities with no concurrent duties.
All fuel handling operations.
Safe and orderly evacuation of the refueling crew in the event of a high radiation alarm at a refueling station.
Determining the cause of high radiation alarms.
Determining that fuel handling personnel are properly qualified for their duty stations.
A. Incorrect. This is a duty of the operator in the control room. Plausible as the refueling SRO provides direct oversight and this could be thought of as falling under that definition.
B. Incorrect. #1 is a duty of the operator in the control room, #3 is the duty of the reactor engineer. Plausible as the refueling SRO provides direct oversight and these could be thought of as falling under that definition.
C. Correct. This is the only one listed that is the responsibility of the refueling SRO D. Incorrect. 3 is the responsibility of the reactor engineer Technical References
- OP B-8SD1 References to be provided to applicants during exam:
none Learning Objective
- Question Source:
Bank # (note changes; attach parent) Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.2 SRO must know the duties of the refueling SRO, this is solely an SRO task.
DCPP L 1 61 Exam Rev 0 Difficulty: 2.5
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO G2.1.41 - Knowledge of refueling process Tier # 3 Group # 1 K/A # G2.1.41 Rating 3.7 Question 95 Which of the following would be the FIRST core alteration activity requiring direct supervision by the Refueling SRO?
A. Lifting the reactor vessel head B. Moving the first fuel assembly C. Lifting the upper internals D. Unlatching RCCAs Proposed Answer:
D. Unlatching RCCAs Explanation:
Plausible, all actions take place during fuel movement A. Incorrect. Not a core alteration B. Incorrect. This is a core aleration, but not the first one listed C. Incorrect. Not a core alteration D. Correct. This is a core alteration and done prior to moving any fuel assemblies Technical References
- OP L-6, OP B-8DS1 References to be provided to applicants during exam:
none Learning Objective
- 6497 - State the responsibilities and duties of Refueling SRO Question Source:
Bank #96 DCPP NRC Exam 02/2005 (note changes; attach parent)
Modified Bank #
New Question History:
Past NRC Exam Yes Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.6 SRO is responsible for supervision of refueling activities.
Difficulty: 2.5
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level S RO G2.2.23 - Ability to track Technical Specification limiting conditions for operations Tier # 3 Group # 2 K/A # G2.2.23 Rating 4.6 Question 96 GIVEN: At 1000, an LCO ACTION has been entered that has a the following REQUIRED ACTIONACTIONSCONDITION REQUIRED ACTION COMPLETION TIME A. One channel inoperable. A.1 Verify affected subsystem isolated.
AND A.2 Restore subsystem to OPERABLE status. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> The initial verification to meet the initial A.1 action is performed at 1010At 1900 it is discovered that the subsequent surveillance has not been performed. The Shift Foreman immediately directs the verification be performed and it is completed successfully at 1945. At 1900: A. remain in only CONDITION A with its original COMPLETION TIME, the subsequent surveillance was done with in the required time limits.
B. exit CONDITION A and enter CONDITION B until 1945 then exit CONDITION B and enter CONDITION A with a new completion time.
C. remain in CONDITION A and enter CONDITION B until 1945 then exit CONDITION B and continue to track only CONDITION A with its original COMPLETION TIME. D. remain in only CONDITION A with its original COMPLETION TIME and invoke SR 3.0.3 for the missed surveillan ce. Proposed Answer:
A.remain in only CONDITION A with its original COMPLETION TIME, th e s ubsequent surveillance w as d one w ith th e r equired time limits. Explanation:
A. Correct. The frequency for the subsequent surveillance is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> plus an additional 25% if needed. Therefore, the crew has until 2010 (time is from time of last successful DCPP L 1 61 Exam Rev 0 completion) before the surveillance would have been missed.
B. Incorrect. Condition B would have applied if the surveillance was not complete prior to 2000, however, once the surveillance was completed (satisfactorily), then Condition B would have been exited and Condition A would have been the only action tracked WITH ITS ORIGINAL completion time.
C. Incorrect. This would have been correct if the frequency had been missed.
D. Incorrect. This applies to missed surveillances.
Technical References
- LCO 3.0, B3.0 and section 1.0 of Tech Spec References to be provided to applicants during exam:
none Learning Objective
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.2 Difficulty: 3.1
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO G2.2.37 - Ability to determine operability and/or availability of safety related equipment.
Tier # 3 Group # 2 K/A # G2.2.37 Rating 4.6 Question 97 A degraded or non
-conforming condition that might have an impact on applicable Systems, Structures, or Components (SSC) operability has just been reported to the Shift Foreman.
The Shift Manager requests that a Prompt Operability Assessment (POA) be performed.
While the POA is performed, the SSC is considered to be ________________.
A. OPERABLE B. available but inoperable C. inoperable and not available D. available but OPERABILITY is indeterminate Proposed Answer:
A. OPERABLE Explanation:
A. Correct. For a POA to be done, the immediate determination must be OPERABLE bu t more detailed investigation, evaluation or analysis is necessary to demonstrate the basis.
B. Incorrect. Per OP1.DC17, Availability
- INOPERABLE equipment is considered available for risk purposes when the equipment is able to perform its intended function but not all licensing or design basis assumptions may be maintained. However, for a POA to be performed, the determination is the expected condition is OPERABLE.
C. Incorrect. The SSC is considered OPERABLE while the POA is performed. Plausible, equipment can be inoperable but available or not available. Could be that its believed it is inoperable until the POA is done and also not available until POA is complete.
D. Incorrect. "indeterminate" is not an allowable outcome. Its either inoperable, operable or operable but more detail is needed.
Plausible that if a POA is requested its due to not knowing the operable/inoperable status.
Technical References
- OM7.ID12 References to be provided to applicants during exam:
none Learning Objective
- Explain the use of Prompt Operability Assessment
- s. (35481) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.3 Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO G2.3.12 - Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high
-radiation areas, aligning filters, etc Tier # 3 Group # 3 K/A # G2.3.12 Rating 3.7 Question 98 Unit 1 is in MODE 1.
Plant and radiological conditions are stable.
Operators are preparing to make multiple containment entries per RCP D-230, Radiological Control for Containment Entry
. Which of the following action(s) listed below is/are the responsibility of the Shift Foreman?
- 1. Authorizing the containment entry
- 2. Maintaining the plant in a stable condition
- 3. Maintaining possession of the MIDS keys A. 1 only B. 1 and 2 C. 2 and 3 D. 1 and 2 a nd 3 Proposed Answer:
B. 1 and 2 Explanation:
Per RCP D-230, SFM authorization must be obtained and it is the responsibility of the SFM to maintain the plant stable for the safety of the workers.
A. Incorrect.
In addition to authorizing the entry, the SFM is also responsible for maintaining stable plant conditions. Plausible because if knowledge of how changing plant conditions could affect the safety of the workers, then it would be possible to think the status of the plant/reactor is not a factor for entr
- y. B. Correct. Both are listed as the responsibility of the SFM in RCP D
-230. C. Incorrect. MIDS keys are kept the RP. Plausible because the SFM maintains possession of most keys.
D. Incorrect. MIDS keys are kept by RP Technical References
- RCP D-230 References to be provided to applicants during exam:
none Learning Objective
- Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundament al X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.4 SRO responsibilities for containment entry.
DCPP L 1 61 Exam Rev 0 Difficulty: 2.5
DCPP L 161 Exam Rev 0 Examination Outline Cross
-Reference Level S RO G2.3.14 -Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. Tier # 3 Group # 3 K/A # G2.3.14 Rating 3.8 Question 99 Which of the following actions that could be taken in t he Emergency Operating P rocedure network, should be evaluated for the possible consequences pr ior to in itiation if th ere is h ig h R C S a ctivity? A. Re-i nitiating RCP seal injection B. Stopping containment spray pumps C. Re-establishing RCP seal return flow D. Initiating cold leg recirc flow from the containment sump Proposed Answer:
C. Re-establishing RCP seal return flow Explanation:
A. Incorrect. The threat is to increased radiation level in the aux building, initiating seal injection, is into containment, not out of containment. Plausible as in EOPs, such as ECA-0.0, there is a caution with seal injection if it has been lost to the RCPs for a period of time.
B. Incorrect. Plausible - If radiation is high, containment spray pumps are not stopped even if containment pressure is below the setpoint. This is an action but it is not based on high RCS activity.
C. Correct. Caution states that if RCS activity is high, D. Incorrect. Flow is initiated regardless of activity level. Plausible as a PA announcement is made prior to initiating flow.
Technical References
- E-1.3, E-1.2, E-1 References to be provided to applicants during exam:
none Learning Objective
- 7920B Explain basis of emergency procedure steps (E
-1) Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X Question History:
Past NRC Exam No Last Two NRC Exams No Question Cognitive Level:
Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:
55.43.4 The SRO must know that of the actions that could be directed to be taken, only an action to bring the activity out of containment requires evaluation.
Difficulty: 3.0
DCPP L 1 61 Exam Rev 0 Examination Outline Cross
-Reference Level S RO G2.4.43 - Knowledge of emergency communications systems and techniques Tier # 3 Group # 4 K/A # G2.4.43 Rating 3.8 Question 100 An Alert Emergency Classification has been declared.
According to the Work Control Shift Foreman checklist in EP
-G-2, Interim Emergency Response Organization
- ________ must be activated within 30 minutes of emergency classification by using a LAN connected computer in ___________.
ERDS - Emergency Response Data System ENS - Emergency Notification System A. ENS; the SM office B. ENS; the affected unit's Control Room C. ERDS; the SM office D. ERDS; the affected unit's Control Room Proposed Answer:
D. ERDS; the affected unit's Control Room Explanation:
A. Incorrect. The ENS is a phone line used to communicate with the NRC during an emergency at Alert or higher. Plausible, it is a method of conveying information to the NRC. B. Incorrect. The ENS is a phone line used to communicate with the NRC during an emergency at Alert or higher. Plausible, it is a method of conveying information to the NRC. C. Incorrect. Plausible
- ERDS is the method used, however, the computer(s) that are used are in the Control Room
. D. Correct.ERDS is the method of sending information to NRC HQ and is done by logging on using a control room computer
. Technical References
- EP G-2, EP EF-1,Activation and Operation of the Technical Support Center References to be provided to applicants during exam:
none Learning Objective
- 3705 As described in EP G
-2, state the responsibilities of the following:
Shift Manager (SM)
Shift Foreman (SFM)
Control Room Communicator #1 Control Room Communicator #2 Work Control Shift Foreman (WCSFM)
Question Source:
Bank # (note changes; attach parent)
Modified Bank #
New X DCPP L 1 61 Exam Rev 0 Question History:
Past NRC Exam No Last Two NRC Exam s No Question Cognitive Level:
Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:
55.43.1 SRO knowledge