ML062890364
| ML062890364 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 10/07/2006 |
| From: | - No Known Affiliation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML062890364 (214) | |
Text
Name: Total Points: 100 Emp. ID: Points Received:
Date: 10/07/2006 Grade:
Exam / Quiz
Title:
OCT2006NRC : October 2006 NRC Exam Open Reference [ ] Yes [X] No Reference Material that may be used to support this open reference exam:
- 2. 2.1.10 Att. 1, 5.3GRID, 2.0.5, 5.7.1 Att. 1
- 3. T.S 3.9.3, 3.3.3.1, Fig 3.1.7.1/2, 3.5.1, 3.8.3, 5.5.8, T3.11.2, T3.4.1-1, DLCO3.1.4 GUIDELINES
- 1. Allotted time to complete the exam / quiz is 6/8 Hrs.
- 2. ALL questions shall be directed to the proctor. Students shall not discuss the questions among themselves until all examinees have completed the exam / quiz.
- 3. Restroom trips are limited and only one examinee at a time may leave the room.
- 4. Verify that all questions have been answered prior to turning in your exam/quiz.
- 5. To pass the exam/quiz, you must achieve an overall grade of 80/70% or greater.
- 6. After you have completed the exam/quiz, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the exam / quiz.
This must be done after you have finished.
I have neither given nor received assistance during the administration of this examination/quiz. (Proctor assistance excluded) All work on this exam is my own.
Examinee Signature:
Prepared by: First Grader:
Approved by: Second Grader:
(If required)
Approved by:
(If required)
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 1 21364 00 06/25/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Abnormal/Emergency INT0320104, Ability to interpret the power to flow map Procedures following a partial loss of RR Flow Related Lessons INT0320104 CNS Administrative Procedures General Operating Procedures (Startup and Shutdown) Procedures (Formal Classroom/Pre-OJT Training)
Related Objectives INT032010400E0200 Discuss Precautions and Limitations associated with Procedure 2.1.10, Station Power Changes.
INT032010400E030A Discuss the following as described in Procedure 2.1.10, Station Power Changes: General Guidelines for Station Power Changes.
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295001.AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
(CFR: 41.10 / 43.5 / 45.13) Power/flow map (3.5/3.8) 1
QUESTION: 1 21364 ( point(s))
The plant is operating at 100% power when a lightning strike causes the following:
% Reactor recirc pump runback
% DEH shifts to mode 3.
The crew locked the scoop tubes for both RR pumps shortly after the transient began. When conditions stabilize the operator notes the following indications:
% Core flow is 48 MLBH.
% Reactor power is 95%.
% Bypass valve position is 50%.
What action is appropriate?
- a. Raise reactor recirc flow.
- b. Scram the reactor and enter 2.1.5.
- c. Reduce reactor power by inserting control rods.
- d. Reduce reactor power by reducing reactor recirc flow.
ANSWER: 1 21364
- c. Reduce reactor power by inserting control rods.
Provide the Candidate with the Power to Flow Map (2.1.10).
Explanation:
Procedure 2.1.10 requires that if rod line exceeds 120.8%, take action to reduce rod line to 120.8% or below.
The only action that would reduce the operating rod line is answer c.
Procedure 2.1.10 states that this condition may be corrected by:
Reduce power using recirculation flow per Section 6 and/or control rods per Procedure 10.13.
Notify Reactor Engineering as soon as possible following power reduction.
- a. is incorrect because this action would not reduce the rod line.
- b. is incorrect because there is no requirement to scram the reactor.
- d. is incorrect because this action would not reduce rod line.
Source is Modified 5421 2
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 2 21365 00 06/05/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0010202, Who is required to be notified for a loss of CW pump gland water.
Related Lessons COR0010202 OPS Circulating Water Related Objectives COR0010202001030D Describe the interrelationship between the Circulating Water system and the following: Service Water COR0010202001060C Predict the consequences a malfunction of the following would have on the Circulating Water system: AC Power Related References 5.3AC480 480 VAC Failures Related Skills (K/A) 2.1.14 Knowledge of system status criteria which require the notification of plant personnel (such as Reactivity Management Events). (CFR: 43.5 / 45.12)
(2.5/3.3) 3
QUESTION: 2 21365 (1 point(s))
A plant startup is in progress with power at 25% when a loss of 480V BUS 1E occurs.
What plant notification is required?
- a. Notify Security that numerous outside lights are out.
- c. Notify Maintenance to shut down sensitive Weld Shop equipment.
- d. Notify CW system engineer that gland water is lost to the CW pumps.
ANSWER: 2 21365 Answer:
- d. Notify CW system engineer that gland water is lost to the CW pump.
Explanation:
With power at 25% at least 1 CW pump is running. The loss of 480V Bus 1E results in a loss of gland water to the operating CW pumps. The Circ Water System Engineer is also notified to provide guidance on circ pump operation without gland water.
Distractors:
- a. Is incorrect because the loss of 480V 1E does not cause the loss of outside lighting. This notification is required for the loss of 480V 1A or 480V 1B.
- b. is incorrect because normal UPS is not lost.
- c. Is incorrect because 480V this notification would only be required if Bus 1B is lost.
Source: New 4
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 3 21366 00 06/05/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Abnormal/Emergency COR0020702, Determine the extent of the loss of DC power.
Procedures Related Lessons COR0020702 OPS DC ELECTRICAL DISTRIBUTION Related Objectives COR0020702001060B Describe the interrelationship between the DC Electrical Distribution System and the following: AC Electrical Distribution COR0020702001060I Describe the interrelationship between the DC Electrical Distribution System and the following: Core Spray COR0020702001060J Describe the interrelationship between the DC Electrical Distribution System and the following: RCIC COR0020702001060M Describe the interrelationship between the DC Electrical Distribution System and the following: Reactor Feedwater System COR0020702001080B Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following: Components using DC control power (i.e., breakers)
COR0020702001110A Predict the consequences of the following events on the DC Electrical Distribution System: Loss of AC Electrical Distribution COR0020702001100E Briefly describe the following concepts as they apply to DC Electrical Distribution System: Loss of breaker protection due to loss of DC Power Related References 5.3DC125 Loss of 125 VDC (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
5
Related Skills (K/A) 295004.AA2.02 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: (CFR: 41.10 / 43.5 / 45.13)
Extent of partial or complete loss of D.C. power (3.5/3.9) 6
QUESTION: 3 21366 (1 point(s))
The plant was operating at 50% during power ascension when a DC failure occurs. The following significant conditions were noted by the crew:
% Both RFPTs transfer to MDEM
% RRMGs runback.
% Both Startup FCVs fail full open
% 9-3-1/B-7, CORE SPRAY A LOGIC POWER FAILURE annunciates.
% 9-4-1/A-3, RCIC LOGIC POWER FAILURE annunciates.
% The BOP notes that control power to 4160VAC buses A, C, and E is ON.
What is the extent of the DC failure?
- a. Only AA1 is lost.
- b. Only AA2 is lost.
- c. Only AA1 and AA2 are lost.
- d. 125 VDC Distribution Panel A is lost.
ANSWER: 3 21366
- b. Only AA2 is lost.
Explanation:
If AA2 is lost RFPT trip logic relay 30TTS will de-energize (RFPT trip input to RRMG runback circuit) and B narrow range level instrument will fail downscale. If B narrow range is the selected level control, RRMGs runback, and RFPTs transfer to MDEM at current speed. Both Startup FCVs will fail full open due to loss of RPV level signal input. RCIC will not operate.
High Level trip on HPCI will not function. Core Spray Pump A, RHR Pump A, and RHR Pump B will not start automatically, but can be started from Control Room. DG-1 will not auto start on High Drywell Pressure or Low Vessel Level. The low pressure permissive opening logic for CS-MO-11A, CS-MO-12A, and RHR-MO-27A will not function ; these valves may be driven closed but may not be opened from the Control Room. RFP A has no trip protection other than mechanical overspeed, can not be tripped from Control Room, and must be tripped locally.
Distractors:
- c. is incorrect because no indication of a loss of AA1 is present and if AA1 were lost control power to 4160 A,C and E would be lost.
- d. is incorrect because of the absence of the loss of TIP indication and the continued power to AA1.
Source: New 7
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 4 21367 00 06/05/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Administrative Turbine trip effect on feedwater temperature and determine what scrams the reactor.
Related Lessons INT0060114 ANTICIPATED OPERATIONAL TRANSIENTS INT0060119 Anticipated Operational Transients and Special Events Related Objectives INT00601140010200 Given an anticipated operational transient that is regularly analyzed, select an action or actions that will terminate the transient.
Related References (B)(5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for Related Skills (K/A) 295005.AK2.02 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: (CFR: 41.7 / 45.8) Feedwater temperature. (2.9/3.0) 8
QUESTION: 4 21367 (1 point(s))
The plant is operating at 22% reactor power with DEH in mode IV when a turbine trip occurs.
How is reactor power affected over the next several minutes? (Assume no operator intervention occurs.)
Reactor power
- a. increases due to increased reactor pressure.
- b. increases due to decreased feedwater temperature.
- c. decreases due to decreased reactor pressure.
- d. decreases due to increased feedwater temperature.
ANSWER: 4 21367
- b. increases due to decreased feedwater temperature.
Explanation:
Reactor power is low enough that turbine stop valve closure and turbine control valve fast closure scrams are bypassed. Following the turbine trip feedwater temperature lowers due to the loss of extraction steam. As the colder feedwater enters the reactor power rises and is initially controlled by the bypass valves. As power rises the bypass valves open in order to control pass the increased steam production from the reactor. The increased reactor power and steam flow will cause a slight increase in reactor pressure due to the controller bias on DEH. But the cause of the power increase is due to feedwater temperature reduction.
Distractors:
- a. is incorrect because the cause of the increase in reactor power is feedwater temperature reduction.
- c. is incorrect because power increases.
- d. is incorrect because power increases.
Source: New 9
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 5 21058 00 08/03/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Systems COR0022002, RPIS Indication Following Scram Related Lessons COR0022002 OPS REACTOR MANUAL CONTROL SYSTEM Related Objectives COR0022002001050L Predict the consequences the following would have on the RMCS and/or RPIS: Reactor scram Related References 2039 Control Rod Drive Hydraulics (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 295006.AA1.06 Ability to operate and/or monitor the following as they apply to SCRAM:
(CFR: 41.7 / 45.6) CRD hydraulic system (3.5/3.6) 10
QUESTION: 5 21058 (1 point(s))
The plant is operating at power when a reactor scram occurs. The following indications are noted on the full core display:
% All Rod Drift Lights are lit.
% All control Rod full in lights are lit.
The operator selects control Rod 26-27 and notes that the four rod display is blank. The following time line of conditions/events then occur:
% 12:00 - Reactor pressure and scram discharge volume pressure equalize.
% 12:02 - Half scram reset is obtained.
% 12:04 - Full scram reset is obtained.
% 12:06 - Scram Discharge Volume Vent and Drain valves are opened.
When does the four rod display FIRST indicate a rod position?
- a. 12:00
- b. 12:02
- c. 12:04
- d. 12:06 ANSWER: 5 21058
- a. 12:00 When pressure is equalized between the reactor and the scram discharge volume the D/P across the drive pistons is zero and at that point the control rods would no longer be inserted past 00 position reed switch and the display would go to 00.
Distractors:
- b. is incorrect because indication is first regained when pressure between the reactor and the discharge volume equalize.
- c. is incorrect because indication is first regained when pressure between the reactor and the discharge volume equalize.
- d. is incorrect because indication is first regained when pressure between the reactor and the discharge volume equalize.
Source: Direct 11
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 6 21372 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Emergency Operating COR0023402, Effect that Remote Shutdown actions have on Procedures LLS.
Related Lessons COR0023402 Alternate Shutdown (LO)
Related Objectives COR0023402001020A Describe the interrelationship between ASD and the following:
Nuclear Pressure Relief (NPR) system Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 295016.AK2.01 Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: (CFR: 41.7 / 45.8) Remote shutdown panel: Plant-Specific (4.4*/4.5*)
12
QUESTION: 6 21372 (1 point(s))
Following a toxic gas event requiring the control room to be abandoned, the following conditions existed:
% The MSIVs are closed
% BOTH Low-Low Set valves are cycling The ADS ISOLATION switch in the Alternate Shutdown Room is now placed in ISOLATE.(NO other actions have been taken outside the control room)
What is the effect on the Low-Low Set valves?
- a. Both Low-Low Set valves stop cycling.
- b. Both Low-Low Set valves continue to cycle.
- c. The Low-Low Set valve (71F), which can be controlled from the ASD room continues to cycle.
- d. The Low-Low Set valve (71D) which cannot be controlled from the ASD room continues to cycle.
ANSWER: 6 21372
- d. The Low-Low Set valve (71D) which cannot be controlled from the ASD room continues to cycle.
Reference:
COR0023402 ASD Room EXPLANATION: Placing the Isolation switch in ISOLATE will prevent operation of the Low-Low set valve operated from the ASD room but does not affect the other Low-Low set valve.
Distractors:
- a. is incorrect because Low-Low set valve 71D can continue to cycle.
- b. is incorrect because Low-Low set valve 71D can continue to cycle.
- c. is incorrect because Low-Low set valve controlled from the ASD room is not be able to cycle on Low-Low set.
Source: Direct 13
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 7 21373 00 06/26/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Abnormal/Emergency COR0021902, Loss of REC Effect on Continued Operation of Procedures Components/Systems Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001060I Given a specific REC malfunction, determine the effect on any of the following: RHR pumps COR0021902001060J Given a specific REC malfunction, determine the effect on any of the following: RWCU system COR0021902001060G Given a specific REC malfunction, determine the effect on any of the following: CRDH system Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295018.AK1.01 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.8 to 41.10) Effects on component/system operations (3.5/3.6) 14
QUESTION: 7 21373 (1 point(s))
The plant is at 100% power when a complete loss of REC occurs. The crew scrams the reactor and all control rods insert. Reactor water level is normal and reactor pressure is being controlled with SRVs.
What system/component can continue to operate indefinitely without REC?
- a. One CRD pump.
- b. RWCU in the blowdown mode.
- c. One reactor recirc pump at 20% speed.
- d. One RHR pump in suppression pool cooling.
ANSWER: 7 21373
- d. One RHR pump in suppression pool cooling.
Explanation:
Following a loss of REC CRD, and RR are secured due to the lack of cooling. The loss of REC prevents the use of RWCU, due to the loss of cooling to the NRHX. A single RHR pump can be operated indefinitely without REC.
Distractors:
Source: New 15
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 8 21374 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: Y Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Abnormal/Emergency COR0011702, Reason for Service Air Isolation Procedures Related Lessons COR0011702 Plant Air Related Objectives COR0011702001110A Given plant conditions, determine if any of the following should occur: Service Air isolation Related References (B)(4) Secondary coolant and auxiliary systems that affect the facility.
Related Skills (K/A) 295019.AK3.03 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: (CFR: 41.5 /
45.6) Service air isolations: Plant-Specific. (3.2/3.2) 16
QUESTION: 8 21374 (1 point(s))
The plant is operating at power when a loss of plant air occurs. Instrument air pressure falls to 75 psig. What automatic action occurs and why is this automatic action required?
- a. PCV-609 closes to isolate the service air header from instrument air to ensure that service air header pressure is preserved during a loss of instrument air.
- b. IA-80 MV closes to separate the reactor building critical and non critical headers in order to preserve pressure in the header without the leak.
- c. IA-80 MV closes to isolate the non-critical air loads to ensure instrumentation and controls remain operable to safely shutdown and cooldown the reactor.
- d. PCV-609 closes to isolate the Service Air header from instrument air to ensure instrumentation and controls remain operable to safely shutdown and cooldown the reactor.
ANSWER: 8 21374 Answer:
- d. PCV-609 closes to isolate the Service Air header from instrument air to ensure instrumentation and controls remain operable to safely shutdown and cooldown the reactor.
Explanation:
PCV-609 isolates the Service Air distribution header from the air compressors and receivers on low system pressure. Automatic isolation of the Service Air header, along with the storage capacity of the air receivers, ensures that the Instrument Air header will be available for a safe shutdown and cool down of the reactor.
Distractors:
- a. is incorrect because separating the headers in order to preserve the side without a leak would only occur if the leak were in the service air header. When 609 closes the service air header is isolated from the compressors and its pressure is not preserved no matter the circumstance.
- b. is incorrect even though the closure of MV-80 does isolate the non critical load it does not do it automatically and the purpose of closing the valve is the preservation of load to safely shutdown and cooldown.
- c. Is incorrect because this valve is not automatically closed as specified in the stem..
Source: New 17
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 9 16796 02 06/27/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description Abnormal/Emergency INT032-01-26, CNS Abnormal Procedures (RO) Cooling Procedures Water Related Lessons COR0022202 REACTOR RECIRCULATION INT0320126 CNS Abnormal Procedures (RO) Cooling Water INT0231002 Pre-Outage Industry Events Related Objectives COR0022202001060G Given a specific Reactor Recirculation system or the Recirculation Flow Control system malfunction, determine the effect on any of the following: Reactor Vessel Internals (jet pumps, stratification, bottom head drain temperature, pump starts)
INT0320126Q0Q0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
INT02310020010500 Identify the actions to be taken in response to a loss of shutdown cooling.
Related References 2.4SDC Shutdown Cooling Abnormal (B)(14) Principles of heat transfer thermodynamics and fluid mechanics.
(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295021.AK1.02 Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING: (CFR: 41.8 to 41.10) Thermal stratification (3.3/3.4) 18
QUESTION: 9 16796 (1 point(s))
With the plant shutdown in Mode 4 and RHR loop "B" operating in Shutdown Cooling, the following conditions exist:
% Reactor pressure is 0 psig
% Recirc suction temperature is 170°F
% Reactor water level is 58" (NR)
% RHR pumps "A" and "C" have both motors disconnected from their pumps "B" RHR Loop develops a leak and a SDC isolation results. Following the isolation the following parameters were noted:
% Reactor pressure is 0 psig
% Recirc suction temperature is 170°F
% Reactor water level is 1" (NR)
What action is required, and why?
- a. RPV water level must be raised to > 48" to aid in natural circulation flow and ensure bulk reactor coolant temperature is known.
- b. RPV water level must be raised to > 48" in order to maximize reactor coolant contact with RPV metal for enhanced heat transfer to the Drywell atmosphere.
- c. A Reactor Recirculation pump must be started to reduce the possibility of excessive thermal stresses on the CRD stub tubes.
- d. A Reactor Recirculation pump must be started in order to reduce the possibility of thermal binding of the RHR-MO-25A/B valves caused by the expected coolant heatup.
ANSWER: 9 16796
- a. RPV water level must be raised to > 48" to aid in natural circulation flow.
Explanation:
If the conditions following the SDC isolation persist, the vessel will become thermally stratified.
If this occurs bulk coolant temperature may not be known. Which could allow temperature to increase to the point that mode changes without the knowledge of the operators.
Answer source: 2.4SDC p. 10, Attachment 2, step 1.1 Distractors:
19
- b. Water level is raised to enhance natural circulation flow, not heat transfer.
- c. A Recirculation pump is started, but not to prevent thermal stresses on the stub tubes.
Starting a recirc pump causes thermal stress on the stub tubes.
- d. A Recirculation pump is started, but not to prevent thermal binding of the gate valves.
These valves are cycled if closed during a cooldown.
Source: Direct 20
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 10 21376 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0012102 , Refueling interlock and reason for the interlock.
Related Lessons COR0012102 Refueling Related Objectives COR0012102001100A Given conditions associated with refueling activities, determine if the following should occur: Refueling platform (bridge) movement restrictions COR0012102001100B Given conditions associated with refueling activities, determine if the following should occur: Refueling mast restrictions Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
(B)(4) Secondary coolant and auxiliary systems that affect the facility.
Related Skills (K/A) 295023.AK3.02 Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS: (CFR: 41.5 / 45.6) Interlocks associated with fuel handling equipment (3.4/3.8) 21
QUESTION: 10 21376 (1 point(s))
Fuel Handling operations are in progress with the Reactor Mode Switch is in REFUEL.
The following sequence of events occurs:
00:00 - Drive the refueling bridge from the fuel pool to over core location 26-27.
00:05 - Lower the Mast from the full up position to the bundle at location 26-27.
00:10 - Grapple the fuel bundle.
00:15 - Raise the mast and fuel bundle.
When is a rod withdrawal block first received and what is the reason this rod block is required?
- a. 00:05 to prevent inadvertent criticality.
- b. 00:15 to prevent inadvertent criticality.
- c. 00:05 to prevent unloading a cell with a withdrawn control rod.
- d. 00:15 to prevent unloading a cell with a withdrawn control rod.
ANSWER: 10 21376 Answer:
- a. 00:05 to prevent inadvertent criticality.
Explanation:
The purpose of refueling interlocks is to restrict control rod movement, and refueling equipment operation, to reinforce operational procedures that prevent making the reactor critical during refueling. A rod block results whenever any of the following groups of conditions are satisfied.
- a. If the is in START-UP and the refueling platform is near or over the core.
- b. If the Mode switch is in REFUEL and;
- 1) A second rod is selected for withdrawal when all rods are not full in, or
- 2) The refueling platform is near or over the core and one or more of the following exists:
- monorail mounted hoist loaded.
- frame mounted hoist loaded.
- fuel grapple loaded.
- fuel grapple not full up Distractors:
22
- b. is incorrect because a rod block first occurs at 00:05.
- c. is incorrect because the reason for the interlock is to prevent inadvertent criticality.
- d. is incorrect because the reason is to prevent inadvertent criticality and the block first occurs at 00:05.
Source: New 23
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 11 3302 01 04/16/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems COR0020302, Effect of High Drywell Pressure on Drywell Ventilation Related Lessons COR0020302 CONTAINMENT Related Objectives COR0020302001130D Describe the PCIS design features and/or interlocks that provide for the following: Bypassing of selected isolations COR0020302001130E Describe the PCIS design features and/or interlocks that provide for the following: Operator action to defeat/reset isolations COR0020302001170A Predict the consequences of the following items on Primary containment: LOCA COR0020302001210C Given plant conditions, determine if the following should have occurred: Drywell cooling fan trip.
Related References 5.8.10 Average Drywell Temperature Calculation (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 295024.EK2.18 Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: (CFR: 41.7 / 45.8) Ventilation. (3.3/3.4) 24
QUESTION: 11 3302 (1 point(s))
A small break LOCA has occurred with the following conditions:
% Reactor water level is +45" (NR).
% Reactor pressure is 560 psig.
% Drywell pressure is 3.1 psig.
% Drywell temperature is 195EF.
% Drywell FCU control switches are in RUN.
What action(s) (if any) will operate ALL available drywell FCUs?
- a. NO actions are required, all FCUs are running.
- b. Start all the FCUs by placing their control switches in OVERRIDE.
- c. Start all the FCUs by placing their control switches to OFF and then to RUN.
- d. Start all the FCUs by placing their control switches in OVERRIDE and then RUN.
ANSWER: 11 3302
- b. Start all the FCUs by placing their control switches in OVERRIDE.
FOILS: a. FCUs have tripped. c. With the switches in RUN the FCU do not operate with a high drywell signal present. d. If the control switches are placed in RUN the FCUs will trip.
REFERENCE:
EOP-3A, PR 5.8.10, Containment Text Source: Direct 25
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 12 21377 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Abnormal/Emergency Monitor RCIC During RCIC Pressure Control with high reactor Procedures pressure.
Related Lessons COR0021802 OPS Reactor Core Isolation Cooling INT0320105 SYSTEM OPERATING PROCEDURES Related Objectives INT03201050000500 Given a specific procedure and situation, discuss any associated cautions or notes stated in the procedure COR0021802001080D Describe the RCIC system design features and/or interlocks that provide for the following: Prevention of turbine damage INT0320105000040B Given a specific procedure, state the associated precautions concerned with the following items: Temperature, Pressure, Power, Flow, Level Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295025.EA1.05 Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: (CFR: 41.7 / 45.6) RCIC: Plant-Specific. (3.7/3.7) 26
QUESTION: 12 21377 (1 point(s))
The plant is operating at 100% power when a group 1 isolation occurred. The crew entered EOP1A recovered level and placed RCIC in pressure control. The following RCIC and plant parameters were noted by the operator:
% RCIC Steam Inlet pressure (RCIC-PI-94) indicates 1045 psig and is slowly rising.
% RCIC Pump Discharge pressure (RCIC-PI-93) indicates 1050 psig.
% RCIC Flow (RCIC-FIC-91) indicates 275 gpm and is fluctuating.
% TURB SPEED (RCIC-SI-3067) indicates 5650 rpm and is fluctuating.
% Reactor water level is 35"and slowly lowering.
Ask what action is required?
- a. Trip the RCIC turbine.
- b. Raise RCIC-FIC-91 setpoint.
- c. Place RCIC-FIC-91 to manual.
- d. Open RCIC-MO-21, PUMP DISCH TO RX VLV ANSWER: 12 21377
- a. Trip the RCIC turbine.
Explanation:
RCIC Turbine speed is above that which requires an automatic turbine trip. RCIC should be manually tripped to accomplish the action that failed to occur automatically.
Distractors:
- b. Is incorrect because a turbine trip is required. This too is an appropriate action if a trip is not required.
- c. is incorrect because a turbine trip is required. If turbine speed were lower then this would be a correct answer. Fluctuating flow when speed is less than 75% requires manual operation of the controller.
- d. Is incorrect because a turbine trip is required. This action would be selected by the candidate that believes the lower reactor level requires immediate injection.
Source: New 27
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 13 19285 03 10/15/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080618, DW Spray with NPSH exceeded Procedures Related Lessons INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806180010200 For each graph used in the flowcharts, identify the action(s) required if the parameters associated indicate operation in the restricted or prohibited area.
Related References 5.8 Emergency Operating Procedures (EOPs)
(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
(B)(8) Components, capacity, and functions of emergency systems.
Related Skills (K/A) 295026.EK1.01 Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE : (CFR: 41.8 to 41.10)
Pump NPSH / 5 28
QUESTION: 13 19285 (1 point(s))
Drywell Sprays, Torus Sprays and Torus Cooling are in service following a LOCA. Several minutes later, the following conditions exist:
% Drywell pressure is 5 psig.
% Drywell temperature is 250°F.
% Torus pressure is 3 psig.
% Torus average water temperature is 180°F.
% Primary containment level is 10 feet.
% RHR loop A system flow is 8000 gpm (ONLY RHR pump A is operating).
% The Control Room Supervisor has directed that operation of RHR and CS pump remain within NPSH and Vortex limits.
How are Drywell sprays affected (if at all) by this direction and why?
Drywell sprays . . .
- a. may continue at current values as no limit is being exceeded.
- c. must be reduced since RHR Pump vortex limit is being exceeded.
- d. must be secured because the Drywell Spray Initiation Limit has been exceeded.
ANSWER: 13 19285
Provide EOP Graphs to the candidate.
The NPSH curve is being exceeded requiring flow to be reduced.
Distractors:
- c. The vortex limit is not be exceeded.
- d. Exceeding the DWSIL after initiation does not require securing the drywell sprays.
29
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 14 21379 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Emergency Operating INT0080618, Ability to Monitor Drywell Pressure to Procedures Determine When Sprays are allowed.
Related Lessons INT0080613 OPS FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806180010200 For each graph used in the flowcharts, identify the action(s) required if the parameters associated indicate operation in the restricted or prohibited area.
INT00806180010300 Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295028.EA1.04 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.7 / 45.6) Drywell Pressure (3.9/4.0) 30
QUESTION: 14 21379 (1 point(s))
The plant experienced a loss of drywell cooling and a small unisolable steam leak into containment. The following parameters were noted:
% Drywell Temperature is 275°F and slowly rising.
% Drywell pressure is 5 psig and slowly rising.
% Torus spray is in operation.
At what point may drywell sprays first be used?
- a. immediately.
- b. When torus pressure reaches 8 psig.
- c. When torus pressure reaches 9 psig.
- d. When torus pressure exceeds 10 psig.
ANSWER: 14 21379 Answer:
- b. When torus pressure reaches 8 psig.
Provide EOP graphs to the candidate.
Explanation:
Currently conditions are in the unsafe region of the DWSIL graph. At a drywell temperature of 270°F the drywell may be sprayed when DW pressure exceeds 7 psig. So of the pressures listed 8 psig is when DW sprays may be initiated.
Distractors:
- a. is incorrect because conditions are currently in the unsafe region of the DWSIL graph.
The candidate that only executes the DW temperature leg of EOP-3A would choose this answer.
- c. Is incorrect even though these conditions are in the safe region of the graph it is not the first as asked in the stem of the question.
- d. Is incorrect and would be chosen by the candidate that only executed the DW pressure leg of EOP-3A.
Source: New 31
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 15 21380 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 1 1 3 Multiple Choice Topic Area Description Emergency Operating INT0080613, Reason a Scram is Required on Low SP Water Procedures Level.
Related Lessons INT0080613 OPS FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related Objectives INT00806130011200 Given plant conditions and EOP flowchart 3A, PRIMARY CONTAINMENT CONTROL, state the reasons for the actions contained in the steps.
Related References (B)(8) Components, capacity, and functions of emergency systems.
Related Skills (K/A) 295030.EK3.06 Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: (CFR: 41.5 / 45.6) Reactor SCRAM. (3.6/3.8) 32
QUESTION: 15 21380 (1 point(s))
The plant is operating at power when a large suppression pool leak occurs. All available sources are making up to the suppression pool. 20 minutes later the following parameters were noted:
% Suppression pool level is 9.8 ft and lowering at 1" every 10 minutes.
% Suppression pool temperature is 88°F.
Why is a reactor scram required?
- a. Insufficient volume of water exists to dampen the dynamic loads on the suppression pool during LOCA.
- b. Insufficient volume of water exists in the suppression pool to absorb the energy from an ADS blowdown.
- c. Level is approaching the opening of the downcomers and steam suppression during a LOCA can no longer be assured.
- d. Level is approaching the opening of the T quenchers and steam suppression during an ADS blowdown can no longer assured.
ANSWER: 15 21380
- c. Level is approaching the opening of the downcomers and steam suppression during a LOCA can no longer be assured.
Explanation:
The RPV is not permitted to remain at pressure or at power if suppression of steam discharged from the RPV cannot be assured. A primary containment elevation of 9.6 feet is the opening of the downcomers, the point at which steam suppression can longer be assured.
Distractors:
- a. is incorrect because even though the bottom of the HCTL graph is at 9.6 ft the basis for this is uncovery of the downcomers.
- b. Is incorrect because at this level and temperature there is sufficient heat capacity to absorb an ADS blowdown.
- d. Is incorrect because the T-quenchers still are covered.
Source: New 33
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 16 21381 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Technical Specifications, INT0070507, Plant condition that indicate non-compliance ODAM, TRM with Technical Specifications.
Related Lessons INT0070507 CNS Tech. Spec. 3.6, Containment Systems Related Objectives INT00705070010100 Given a set of plant conditions, recognize non-compliance with a Chapter 3.6 LCO.
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications. (CFR: 43.2 / 43.3 / 45.3)
(3.4/4.0) **EXAM USE ONLY**
34
QUESTION: 16 21381 (1 point(s))
The plant was operating at 100% with the RCIC Test Surveillance in progress:
% Average Drywell Air Temperature is 145°F.
% Drywell Pressure is 0.70 psig
% Suppression Pool Level is +5.0"
% Suppression Pool Temperature is 99°F.
% Reactor Water Level is 35"
% Reactor Pressure is 1007 psig.
Entry into the Conditions and Action statements of which Technical Specification LCO is required?
- a. 3.6.1.4 Drywell Pressure
- b. 3.6.1.5 Drywell Air Temperature
- c. 3.6.2.1 Suppression Pool Average Temperature
- d. 3.6.2.2 Suppression Pool Water Level ANSWER: 16 21381
- d. 3.6.2.2 Suppression Pool Water Level Explanation:
The indicated suppression pool level is above the LCO level of 12'11". Therefore entry into 3.6.2.2 is required. The indicated level is the equivalent of 13'2".
Distractors:
- a. is incorrect because drywell pressure is below the .75 required for entry into 3.6.1.4.
- c. is incorrect because average drywell temperature is below that which requires entry into 3.6.1.5..
- d. is incorrect because entry into the LCO is not required during testing that adds heat to the SP.
Source: New 35
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 17 21382 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 5 Multiple Choice Topic Area Description Abnormal/Emergency INT0320130, Recognize the Plant Conditions that Require Procedures Entry Into 5.2FUEL.
Related Lessons INT0320130 CNS Abnormal Procedures (RO) High Radiation Related Objectives INT0320130E0E0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) (4.0/4.3)
- LINK ONLY TO EOP/AOP LESSONS/QUESTIONS**
36
QUESTION: 17 21382 (1 point(s))
A plant accident occurred that resulted in a loss of all reactor feedwater. The crew entered EOP1A and the following events occurred:
% 1000 Reactor water level goes below TAF
% 1005 MSL Radiation Monitors rise from near background to their Hi Setpoint
% 1010 Drywell Radiation Monitors rises above 250 Rem/hr.
% 1015 Refuel Floor CAM alarms When is entry into 5.2FUEL first required?
- a. 1000
- b. 1005
- c. 1010
- d. 1015 ANSWER: 17 21382 Answer:
- b. 1005 Explanation:
The rise in MSL Radiation monitors is unexplained for reasons other than fuel failure. This is an entry condition for 5.2Fuel.
Distractors:
- a. is incorrect because although reactor water level less than TAF could precipitate into fuel failure in an of itself it is not an entry condition. And minus other indications of fuel failure entry would be inappropriate.
- c. is incorrect because although this is an entry condition the stem asked for FIRST entry condition and the entry is FIRST required when the MSL Rad monitors rise.
- d. is incorrect because although this is an entry condition the stem asked for FIRST entry condition and the entry is FIRST required when the MSL Rad monitors rise.
Source: New 37
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 18 14475 01 07/29/2003 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Emergency Operating INT0080610, Actions required during ATWS with Group 1 Procedures failure Related Lessons INT0080610 OPS EOP FLOWCHART 7A - RPV LEVEL (FAILURE-TO-SCRAM)
Related Objectives INT00806100010800 Given plant conditions and EOP flowchart 7A, RPV LEVEL (FAILURE TO SCRAM), determine required actions.
Related References EOP-6A EOP Flow Chart 6A EOP-7A EOP Flow Chart 7A (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295037.EA2.07 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: (CFR: 41.10 / 43.5 / 45.13) Containment conditions/isolations. (4.0/4.2*)
38
QUESTION: 18 14475 (1 point(s))
The plant is operating at rated power when a main turbine bypass valve fails full open.
MSL pressure lowered to 700 psig with reactor power at 65%. The Reactor Operator depressed the Manual Scram pushbuttons, but little rod motion occurred. No other operator actions have been taken. Plant conditions are:
% Reactor pressure is 680 psig and lowering.
% Reactor power is 26% and steady.
% Reactor water level is +48" (NR) and steady.
% Steam Flow is 2.4 MLBH.
% MSIVs are OPEN.
% Torus Temperature is 85EF.
What actions are required at this time?
Place the mode switch to shutdown
- c. leave MSIVs open and trip recirc pumps.
ANSWER: 18 14475
Provide to the candidate; EOP 7A with entry conditions, notes and cautions deleted.
EOP 7A requires the operator to ensure that PCIS isolations 1-7 initiated as required. With main steam line pressure below the setpoint with the mode switch still in RUN, a group 1 isolation should be accomplished. Placing the mode switch to shutdown, and Inhibiting ADS are all appropriate initial additional actions. Clear indications are also provided to the operators by the continued lowering reactor pressure that without MSIV closure, reactor pressure cannot be stabilized.
- b. SLC initiation is not required until RMS is in SD, ARI has been initiated, Recirc pumps have been tripped.
39
Source: Direct 40
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 19 21511 00 08/08/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Abnormal/Emergency COR0011602, Interpret Radiation levels and determine when Procedures the MVP trips Related Lessons COR0011602 Off Gas Related Objectives COR0011602001060B Describe the interrelationships between the Off gas system and the following: Process radiation monitoring COR0011602001130C Given plant conditions, determine if the following should occur:
Mechanical Vacuum Pump trip.
Related References (B)(4) Secondary coolant and auxiliary systems that affect the facility.
Related Skills (K/A) 295038.EA2.03 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.10 / 43.5 / 45.13) ?Radiation levels (3.5*/4.3*)
41
QUESTION: 19 21511 (1 point(s))
A hot plant startup plant startup is in progress with the mechanical vacuum pumps in service. As reactor power increased to the POAH MSL and Off-gas Radiation levels increased.
The following then occurs:
16:00 MSL radiation levels reach their High setpoint.
16:05 Off-Gas Radiation Monitors reach their High setpoint.
16:10 MSL radiation levels reach their High-High setpoint.
16:15 Off-Gas Radiation Monitors reach their High-High setpoint.
When do the mechanical vacuum pumps trip?
- a. 16:00
- b. 16:05
- c. 16:10
- d. 16:15 ANSWER: 19 21511
- c. 16:10 Explanation:
The MVPs trip on a high Main Steam line radiation signal of greater than 3 times the normal full power background radiation level. This occurred at 16:10 Source: New 42
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 20 21149 00 08/07/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 3 1 7 Multiple Choice Topic Area Description Systems COR0010502, Will DG CO2 System Actuate Related Lessons COR0010502 FIRE PROTECTION SYSTEM Related Objectives COR0010502001110F Given plant conditions, determine if the following should occur:
Initiation of Total Flooding High Pressure CO2 in associated Diesel Generator Room.
Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 600000.AK2.01 Knowledge of the interrelations between PLANT FIRE ON SITE and the following: Sensors / detectors and valves (2.6/2.7) 43
QUESTION: 20 21149 (1 point(s))
All four DG-2 NORMAL-ISOLATE switches are in ISOLATE following a fire induced shutdown from outside the Control Room. The DG-1 and DG-2 Total Flooding High Pressure CO2 system controls are aligned as follows:
% DG-1 System ABORT SWITCH is in NORMAL.
% DG-2 System MAIN-RESERVE switch is in RESERVE.
A small fire breaks out in the day tank room. The DG-2 day tank room thermal detector actuates and one DG-2 room smoke detector actuates.
How does the system respond?
ANSWER: 20 21149
With DG-2 in the RESERVE position the DG-2 system uses the DG-1 bottles as long as the DG-1 switch is in NORMAL.
Distractors:
- b. is incorrect #1 cylinders discharge into the #2 DG room.
- d. is incorrect because CO2 discharges.
Source: Direct 44
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 21 21385 00 06/21/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0022202, Implications of Low Reactor Level on RR Pump NPSH Related Lessons COR0022202 REACTOR RECIRCULATION Related Objectives COR0022202001050E Briefly describe the following concepts as they apply to the Reactor Recirculation system or to the Recirculation Flow Control system during normal and reduced forced flow conditions: Indications of Pump Cavitation (including low reactor water level affects on carry under, jet pump NPSH and Recirc pump NPSH)
COR0022202001070E Predict the consequences a malfunction of the following would have on the Reactor Recirculation system or the Recirculation Flow Control system: Feedwater Flow/Feedwater Flow Inputs (including core inlet subcooling and recirc pump NPSH)
COR0022202001100A Describe the Reactor Recirculation system and/or Recirculation Flow Control system design features and/or interlocks that provide for the following: Adequate Recirculation Pump NPSH Related References (B)(14) Principles of heat transfer thermodynamics and fluid mechanics.
Related Skills (K/A) 295009.AK1.02 Knowledge of the operational implications of the following concepts as they apply to LOW REACTOR WATER LEVEL: (CFR: 41.8 to 41.10)
Recirculation pump net positive suction head: Plant- Specific (3.0/3.1) 45
QUESTION: 21 21385 (1 point(s))
What event would result in the lowest value of available NPSH to the reactor recirculation pumps? (For all options assume the RR pumps are operating).
- a. Loss of all feedwater at 20% power.
- b. Stuck open relief valve at rated power.
- c. Inadvertent RCIC initiation at 75% power.
- d. Reactor recirc pump runback to 45% from high power.
ANSWER: 21 21385
- a. Loss of all feedwater at 20% power.
Explanation:
NPSH is as a measure of the difference between the saturation pressure and the total pressure felt at the inlet of the pump. The total pressure is made up of two elements; the height of the column of water above the pump, and the amount of subcooling at the pump inlet. If the total pressure at the suction of the pump drops below the required NPSH for the pump, cavitation will occur.
Cavitation causes excessive noise, pump vibration and reduction in pumping capacity. This leads to reduced pump efficiency and possible pump damage.
During low power operations the significant factor affecting the Recirc pump NPSH is the height of the column of water above the pump, about 57 ft (feedwater subcooling effects though present, are minimal). This provides adequate NPSH to the pumps as long as water level is maintained in the normal operating band. Lowering water level will reduce the NPSH for both the Recirc pump and the jet pumps. Therefore a loss of all feedwater at low power would provide a situation where both the low feedwater flow and the low level contribute to a low value of NPSH to the RR pumps.
Distractors:
- b. is incorrect because a stuck open relief valve would in actuality raise the value of available NPSH. Some the steam going through the relief valve would not go through the turbine reducing the amount of extraction steam and feedwater heating.
- c. is incorrect because RCIC initiation would provide relatively cool water to the downcomer and therefore would increase NPSH available.
- d. is incorrect as the this event would not reduce NPSH as much as the loss of feedwater at low power.
Source: New 46
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 22 19739 02 02/26/2003 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080613, Interpret Drywell Pressure to Determine Procedures Required Actions.
Related Lessons INT0080613 OPS FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related Objectives INT00806130011100 Given plant conditions and EOP Flowchart 3A, PRIMARY CONTAINMENT CONTROL, determine required actions.
INT00806130011000 Identify any EOP support procedures referenced in Flowchart 3A and apply any associated special operating instructions or cautions.
Related References 5.8 Emergency Operating Procedures (EOPs)
(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295012.AA2.02 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.10 / 43.5 / 45.13) Drywell pressure (3.9/4.1) 47
QUESTION: 22 19739 (1 point(s))
Given the following conditions:
% A small coolant leak has occurred in the Drywell.
% Drywell temperature is 185°F and rising slowly.
% Drywell pressure is 4.0 psig and rising slowly.
What action is required to control Drywell conditions?
- a. Initiate drywell sprays.
- b. Operate all available drywell cooling.
- c. Vent primary containment with torus vent line.
- d. Vent primary containment with drywell vent line.
ANSWER: 22 19739
- b. Operate all available Drywell Cooling.
This action is specified in the drywell temperature leg of EOP-3A when drywell temperature cannot be maintained below 150EF.
Answer source: EOP flowchart 3A step DW/T-3 Distractors:
- a. Drywell sprays are not permitted with torus pressure at the current 4.0 psig.
- c. Venting the torus is not allowed with the LOCA signal present and pressure well below PCPL-A.
- d. Venting the drywell is not allowed with the LOCA signal present and pressure well below PCPL-A.
Source: Direct 48
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 23 21386 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Abnormal/Emergency INT0060119, Reason for a Scram During an Inadvertent Procedures Reactivity Addition Related Lessons INT0060114 ANTICIPATED OPERATIONAL TRANSIENTS INT0060119 Anticipated Operational Transients and Special Events Related Objectives INT00601140010300 Given a list of transients, select the transient that would most limiting with respect to MCPR considerations.
INT00601190010300 Given a list of Anticipated Operational Transients, select the most limiting transient with respect to MCPR (Minimum Critical Power Ratio).
INT00601140010400 Given a transient and list of reasons, choose the reason the given transient would have MCPR limitations.
INT00601190010400 Given an Anticipated Operational Transient and a list of reasons, select the correct response why the given transient would have MCPR limitations.
Related References (B)(1) Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.
Related Skills (K/A) 295014.AK3.01 Knowledge of the reasons for the following responses as they apply to INADVERTENT REACTIVITY ADDITION: (CFR: 41.5 / 45.6) Reactor SCRAM. (4.1*/4.1) 49
QUESTION: 23 21386 (1 point(s))
The plant is operating at 100% power when a generator load reject occurs with a failure of the turbine bypass valves to open.
What reactor scram signal terminates the transient and why is a scram required?
- b. Emergency trip header pressure signal scrams the reactor to prevent violating the MCPR safety limit.
- d. Emergency trip header pressure signal scrams the reactor to prevent violation of the RPV pressure safety limit.
ANSWER: 23 21386
- b. Emergency trip header pressure signal scrams the reactor to prevent violating the MCPR safety limit.
Explanation:
This transient directly threatens MCPR and the transient is terminated by the emergency trip header pressure scram signal. This is a large positive reactivity addition and the reason for the scram is to prevent exceeding MCPR.
Distractors:
- a. is incorrect because the expected scram that terminates this transient is the emergency trip header pressure signal. Since this transient causes a large pressure increase it is expected that the candidate that doesn't fully understand this transient would choose this answer because a large pressure transient is expected.
- c. is incorrect because the expected scram that terminates this transient is the emergency trip header pressure signal. Since this transient causes a large pressure increase it is expected that the candidate that doesn't fully understand this transient would choose this answer because a large pressure transient is expected. This transient doesn't threaten the pressure safety limit.
- d. is incorrect because the transient doesn't threaten the pressure safety limit.
Source: New 50
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 24 21387 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0020401, Locate local controls and indications during a loss of CRD pumps.
Related Lessons COR0020401 Control Rod Drive Hydraulics Related Objectives COR0020402001050C Briefly describe the following concepts as they apply to the CRDH system: Pressure indication COR0020402001110I Predict the consequences a malfunction of the following would have on the CRDH systems: CRDH pump trip.
COR00204010010300 State the location of the major system components of the Control Rod Drive System. Component locations and the location of local indications/alarms may not be stated in this text. The ability of the individual to trace system flowpaths and state locations is implied.
Specific instances may be covered in the lecture, plant tours and/or OJT.
Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 2.1.30 Ability to locate and operate components / including local controls. (CFR:
41.7 / 45.7) (3.9/3.4) **NRC EXAM ONLY**
51
QUESTION: 24 21387 (1 point(s))
During power operation the operating CRD pump trips and the standby CRD pump cannot be started. Isolation of the Charging Water Header is required.
How is the charging water header isolated?
Where is this valve operated?
- a. CRD-29 Charging Water Header Root valve is closed.
Southeast Reactor Building 903 level.
- b. CRD-29 Charging Water Header Root valve is closed.
Southeast Reactor Building 881 level.
- c. CRD-MO-20, DRIVE PRESSURE CONT VALVE, is closed.
Control Room Panel 9-5.
- d. CRD-MO-20, DRIVE PRESSURE CONT VALVE, is closed.
Southeast Reactor Building 903 level.
ANSWER: 24 21387
- a. CRD-29 Charging Water Header Root valve is closed.
Southeast Reactor Building 903 level.
On the loss of both CRD pumps CRD-29 the CHARGING WATER HEADER ROOT VALVE Located in the reactor building 903 South East is closed. Closing this valve isolates the pressure instrument that sends a signal to the Control Room.
Distractors:
- b. is incorrect because the valve is located on the 903 level.
- c. is incorrect because closing this valve will not preserve the charging water header pressure.
- d. is incorrect because closing this valve will not preserve the charging water header pressure.
Source: New 52
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 25 21388 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Emergency Operating INT0080617, Relationship Between Area Temperature Alarms Procedures and Secondary Containment Related Lessons INT0080617 OPS FLOWCHART 5A - SECONDARY CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL Related Objectives INT00806170010200 State the basis for secondary containment parameter maximum normal operating values (MNO).
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295032.EK2.06 Knowledge of the interrelations between HIGH SECONDARY CONTAINMENT AREA TEMPERATURE and the following: (CFR: 41.7 /
45.8) Area temperature monitoring system (3.3/3.4) 53
QUESTION: 25 21388 (1 point(s))
What is the significance of the setpoint for the Secondary Containment area high temperature alarms?
- a. Secondary Containment equipment required for safe shutdown fails.
- b. Indication that a steam leak may be occurring in secondary containment.
- c. This is the maximum allowed setting for the leak detection instrumentation.
- d. At this temperature personnel access is to secondary containment is precluded.
ANSWER: 25 21388
- b. Indication that a steam leak may be occurring in secondary containment.
Explanation:
An area temperature above its maximum normal operating (MNO) value (Table 9) is an indication that steam from a primary system may be discharging into secondary containment.
The secondary containment temperature MNO values are based on the alarm set points of selected leak detection temperature instrumentation.
Distractors:
- a. is incorrect because this would not occur until the Maximum Safe Operating (MSO) temperature is exceeded.
- c. is incorrect because the maximum allowed setting for the leak detection instrumentation coincides with the MSO temperature.
- d. is incorrect because personnel access is not necessarily precluded at this temperature.
Source: New 54
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 26 21316 00 07/30/2005 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0022802, Monitor SGT System following Auto initiation from a group 6 isolation.
Related Lessons COR0022802 OPS STANDBY GAS TREATMENT Related Objectives COR0022802001080A Describe the Standby Gas Treatment design features and/or interlocks that provide for the following: Automatic system initiation Related References COR0022802 Standby Gas Treatment System (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 295035.EA1.02 Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:
(CFR: 41.7 / 45.6) SBGT/FRVS. (3.8/3.8) 55
QUESTION: 26 21316 (1 point(s))
A Group 6 isolation occurs due to high radiation in the reactor building ventilation exhaust plenum. Following the automatic start of SGT the following parameters and control positions were noted:
% Reactor building pressure is -0.40" H2O.
% SGT train A Filter train pressure drop is 11" H2O D/P.
% SGT train B Filter train pressure drop is 11" H2O D/P.
% SGT-DPCV-546A or SGT-DPCV-546B control switches are in AUTO.
What is the expected position/response of SGT differential pressure control valves (SGT-DPCV-546A or SGT-DPCV-546B) and the SGT fan inlet vortex dampers?
SGT differential pressure control valves...
- a. are full open and remain full open and the vortex dampers are full open and remain full open.
- b. are full open and remain full open and the vortex dampers close to obtain 10" H2O D/P across their filter trains.
- c. modulate to maintain -0.25" H2O in the Reactor Building and the vortex dampers are full open and remain full open.
- d. modulate to maintain -0.25" H2O vacuum in the Reactor Building and the vortex dampers close to obtain 10" H2O D/P across their filter trains.
ANSWER: 26 21316
- b. are full open and remain full open and the vortex dampers close to obtain 10" H2O D/P across their filter trains.
A group 6 isolation signal (the reactor building ventilation exhaust plenum radiation) signal causes the differential pressure control valves (SGT-DPCV-546A or SGT-DPCV-546B) to open and remain open until the signal is reset providing their respective control switches are in AUTO. The fan vortex control system limits air stream flow through each filter train so that total pressure drop across the train remains less than 10" of water D/P.
- a. is incorrect because the vortex dampers would not be full open.
- c. is incorrect because the D/P control valves are open and remain open until the isolation is reset.
56
- d. is incorrect because the D/P control valves are open and remain open until the isolation is reset.
Source: Direct.
Note: Used on OCT2005NRC EXAM 57
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 27 21390 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Emergency Operating INT0080617, Interpret the cause of the high secondary Procedures containment water level.
Related Lessons INT0080617 OPS FLOWCHART 5A - SECONDARY CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL Related Objectives INT00806170010200 State the basis for secondary containment parameter maximum normal operating values (MNO).
INT00806170010300 Explain why the reactor must be shutdown and depressurized if a secondary containment parameter exceeds its maximum safe operating value in 2 or more areas and the primary system is discharging into secondary containment.
INT00806170010400 State the basis for the limits of the maximum safe operating values (MSO) as they apply to personnel protection and equipment operability.
INT00806170010700 Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 295036.EA2.03 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:
(CFR: 41.10 / 43.5 / 45.13) Cause of the high water level. (3.4/3.8) 58
QUESTION: 27 21390 (1 point(s))
During Operation a large suppression pool leak occurs. Water levels in the NW and SW quads eventually exceed their maximum safe operating levels. (Assume that the crew is able to maintain normal level in the SP).
What action is required by EOP-5A?
- a. Shutdown per 2.1.4.
- b. Shutdown per 2.1.5.
ANSWER: 27 21390
- b. Shutdown per 2.1.5.
Provide EOP-5A with the entry conditions and cautions removed.
Explanation:
With two areas greater than Maximum Safe but no primary system discharging into secondary containment a shutdown per 2.1.5 is required by SC-15.
Distractors:
- a. is incorrect because a shutdown per 2.1.5 is required. This action would have been appropriate prior to two areas exceeding their maximum safe value.
- c. is incorrect because entry into EOP-1A from EOP-5A only occurs when a primary system is discharging into containment.
- d. is incorrect because entry into EOP-1A from EOP-5A only occurs when a primary system is discharging into containment and an ED is only required if a primary system is discharging into containment.
Source: New 59
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 28 21391 00 06/27/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0022302, Effect that a malfunction of pressure maintenance will have on the RHR/LPCI.
Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001080D Predict the consequences a malfunction of the following will have on the RHR system: Pressure maintenance system Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 203000.K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC): (CFR: 41.7 /
45.7) Keep fill system (3.3/3.5) 60
QUESTION: 28 21391 (1 point(s))
The RHR system is in normal LPCI standby lineup with condensate secured to the Reactor Building Auxiliary Condensate System.
What potential effect would the loss of the Reactor Building Auxiliary Condensate Pump have on the RHR system?
- b. Low shutoff head on LPCI initiation.
- c. Water hammer on LPCI system initiation.
- d. Low system peak flow rate on LPCI initiation.
ANSWER: 28 21391
- c. Water hammer on LPCI system initiation.
Explanation:
The Condensate system normally provides pressure maintenance in order to maintain pressure in the pump discharge piping to prevent water hammer on system startup.
Distractors:
- a. is incorrect because the loss of pressure maintenance has no effect on NPSH.
- b. is incorrect because once the pump suction is flooded the LPCI shutoff head will be identical. The pump suction will remain flooded irrespective of the status of the keep fill system.
- d. is incorrect because system flow rate will not be effected. In fact flow may be higher for a short period of time while the voids in the system fill.
Source: New 61
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 29 21392 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0022302, Effect that a lowering reactor water level has on SDC Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001090A Explain the significance of the following as they apply to a loss of Shutdown Cooling: Reactor water (level, pressure, temperature)
Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 205000.K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): (CFR: 41.7 / 45.7) Reactor water level (3.6/3.6) 62
QUESTION: 29 21392 (1 point(s))
The plant is shutdown with B RHR pump in shutdown cooling with reactor water level at +60" when a loss of coolant accident occurs. Reactor water level drops steadily from its current level to less than -150"(WR).
What is the status of RHR-MO-25B (Inboard Injection Valve) and RHR pump B?
- a. RHR-MO-25B is open and RHR pump B is ON.
- b. RHR-MO-25B is open and RHR pump B is OFF.
- c. RHR-MO-25B is closed and RHR pump B is ON.
- d. RHR-MO-25B is closed and RHR pump B is OFF.
ANSWER: 29 21392
- d. RHR-MO-25B is closed and RHR pump B is OFF.
Explanation:
With B Loop of RHR in Shutdown cooling the SDC isolation is enabled. That is with RHR-MO-17 and 18 open a group 2 isolation results in the closure of 17 and 18 and RHR-MO-25.
When 17 and 18 go closed RHR pump B trips on no suction path. If water level continues to drop to the LPCI initiation setpoint then the RHR pump gets a start signal but the pump trips with the breaker anti pump feature due to the continued presence of the no suction path trip signal. RHR-MO-25 remains closed until the 25 reset pushbutton is depressed.
Distractors:
- a. is incorrect because RHR-MO-25 is closed and the pump is OFF. The candidate that believes that the LPCI initiation signal realigns MO-25 and successfully starts MO-25 would choose this answer.
- b. is incorrect because RHR-MO-25 is closed. The candidate that is knowledgeable about the anti -pump circuit but not the MO-25 logic would choose this answer.
- c. is incorrect the RHR pump is OFF. The candidate that is knowledgeable about the MO-25 logic but not the anti pump logic would choose this answer.
Source: New 63
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 30 21505 00 08/02/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Technical Specifications, INT0070506, Knowledge of HPCI Technical Specification ODAM, TRM LCO Related Lessons INT0070506 OPS Tech. Spec. 3.5, Emergency Core Cooling (ECCS) and Reactor Core Isolation Cooling (RCIC) System Related Objectives INT00705060010100 Given a set of plant conditions, recognize non-compliance with a Section 3.5 LCO.
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR:
43.2 / 45.2) (3.4/4.1) **EXAM USE ONLY**
64
QUESTION: 30 21505 (1 point(s))
What plant conditions would require entry into the conditions and actions of Technical Specification LCO 3.5.1 ECCS Operating?
- a. HPCI inoperable with reactor pressure at 450 psig.
- b. RCIC inoperable with reactor pressure at 450 psig.
- c. One Low Low Set (LLS) is found to inoperable in Mode 1.
- d. RHR loop A containment spray valve found stuck closed in Mode 1.
ANSWER: 30 21505
- a. HPCI inoperable with reactor pressure at 450 psig.
Explanation:
Since reactor pressure is greater than 150 psig LCO 3.5.1 requires HPCI to be operable.
Distractors:
- b. is incorrect because RCIC inoperable required entry into a different LCO 3.5.3.
- c. is incorrect because this LLS inoperability requires entry into 3.6.1.6.
- d. is incorrect because this requires entry into TRM 3.6.1.
65
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 31 21394 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0020602, Ability to control reactor water level with Core Spray from the Control Room Related Lessons COR0020602 CORE SPRAY Related Objectives COR0020602001050H Describe the Core Spray system design features and/or interlocks that provide for the following: Automatic system initiation COR0020602001080H Given a Core Spray component manipulation, predict and explain the changes in the following: System lineup COR0020602001090A Predict the consequences of the following items on the Core Spray System: Valve closures COR0020602001090B Predict the consequences of the following items on the Core Spray System: Pump trips COR0020602001120B Given plant conditions, determine if any of the following Core Spray Actions should occur: Pump starts.
COR0020602001120D Given plant conditions, determine if any of the following Core Spray Actions should occur: Valve reposition.
Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 209001.A2.02 Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those...: (CFR: 41.5 /
45.6) Valve closures (3.2/3.2) 66
QUESTION: 31 21394 (1 point(s))
A LOCA occurs that causes high drywell pressure and low reactor water level. The Core Spray subsystem automatically initiates and level is restored to +60"(NR).
% Both CS-MO-12A and 12B are closed.
% Both Core Spray pump control switches are taken to stop and then released.
How is the CS system impacted, if at all, should reactor water level fall to -150(WR)? (Assume that the High Drywell Pressure initiation signal for Core Spray remains for the entire from the initial LOCA to present.)
If water level falls to -150(WR), what action(s), if any, is/are required to inject with Core Spray?
- a. Both Core Spray systems remain idle.
Start the Cores Spray pumps only.
- b. Both Core Spray systems remain idle.
Start the Core Spray pumps and open CS-MO-12A and 12B.
- c. Both Core Spray pumps automatically start.
Open CS-MO-12A and 12B.
- d. Both Core Spray pumps automatically start.
No actions are required to inject CS-MO-12A and 12B automatically open.
ANSWER: 31 21394
- b. Both Core Spray systems remain idle.
Start the Core Spray pumps and open CS-MO-12A and 12B.
Explanation:
Since the Core Spray initiation signal remains present the system does not restart just because reactor water level drops to less than the initiation setpoint. Therefore in order to inject with core spray the operator must restart the CS pumps and reopen CS-MO-12A and 12B per 2.2.9.
Distractors:
- a. is incorrect because just restarting the pumps does not restore injection 12A and 12B must be reopened.
- c. is incorrect because the CS pumps do no auto start.
- d. is incorrect because the CS pumps do no auto start and 12A and 12B have to be reopened.
Source: New 67
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 32 21506 00 08/02/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0022902, Automatic actions that occur following SLC initiation.
Related Lessons COR0022902 STANDBY LIQUID CONTROL Related Objectives COR0022902001050F Describe the SLC design features and/or interlocks that provide for the below: RWCU isolation Related References (B)(9) Shielding, isolation, and containment design features, including access limitations.
Related Skills (K/A) 211000.A3.06 Ability to monitor automatic operations of the STANDBY LIQUID CONTROL SYSTEM including: (CFR: 41.7 / 45.7) RWCU system isolation: Plant-Specific (4.0*/4.1*)
68
QUESTION: 32 21506 (1 point(s))
What automatic actions occur when the Standby Liquid Control injection pump "A" control switch is placed in "START"?
SLC pump A starts, Squib valve 14A fires
- a. and squib valve 14B fires.
- b. and RWCU-MOV-MO15, Inboard Isolation Valve closes.
- c. and RWCU-MOV-MO18, Outboard Isolation Valve closes.
- d. RWCU-MO-MO15 and RWCU-MO-18 Inboard and Outboard Isolation Valves Close.
ANSWER: 32 21506
- b. and RWCU-MOV-MO15, Inboard Isolation Valve closes.
Explanation:
RWCU-MO-15 automatically closes with the start of SLC pump 1A.
Distractors:
- c. is incorrect because the RWCU-MO-18 valve does not automatically close upon actuation of the SLC pump 1A switch.
- d. is incorrect because the RWCU-MO-18 valve does not automatically close upon actuation of the SLC pump 1A switch.
Source: New 69
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 33 19035 01 07/03/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems COR0022902, 4160 Bus 1F failure, what is the preferred SLC injection method?
Related Lessons COR0022902 STANDBY LIQUID CONTROL Related Objectives COR0022902001100C Predict the consequences a malfunction of the following would have on the SLC system: A.C.Power Related References NONE Related Skills (K/A) 211000.K2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7)
SBLC pumps (2.9*/3.1*)
70
QUESTION: 33 19035 (1 point(s))
An ATWS condition exists when a loss of power to Bus 1F occurs. The decision to inject SLC has been made.
Based on this power failure, what is the preferred SLC injection method?
- a. RCIC
- b. RWCU
- c. "A" SLC Pump
- d. "B" SLC Pump ANSWER: 33 19035
- d. "B" SLC Pump Explanation: SLC pump B is still available since it is powered from MCC-S which is powered by 4160 Bus 1G.
Distractors:
- b. is incorrect. RWCU is not utilized unless the SLC system is unavailable (EOP 5.8.8) and power is lost to one of its isolation valves.
- c. is incorrect. A SLC pump has lost power and will not start it is powered by MCC K which is powered by 4160 Bus 1F.
Source: Direct 71
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 34 570 3 08/06/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Systems COR0022102 Reactor Protection System Related Lessons COR0022102 REACTOR PROTECTION SYSTEM Related Objectives COR0022102001040A Describe the RPS design features and/or interlocks that provide for the following: System redundancy and reliability COR0022102001050A Briefly describe the following concepts as they apply to RPS: Logic arrangements Related References NONE Related Skills (K/A) 212000.K5.02 Knowledge of the operational implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM: (CFR: 41.5 / 45.3) Specific logic arrangements (3.3/3.4) 72
QUESTION: 34 570 (1 point(s))
The low water level scram is arranged to provide __________ to ensure reliability and minimize inadvertent trips.
- a. one-out-of-two-taken-once logic
- b. one-out-of-two-taken-twice logic
- c. two-out-of-three-taken-once logic
- d. two-out-of-four-taken-twice logic ANSWER: 34 570
- b. one-out-of-two-taken-twice logic The automatic scram channels contain the redundant contacts and switches which are operated by the separate reactor plant parameter sensors (i.e., each channel contains one contact that is operated by a single pressure switch on high reactor vessel pressure). The automatic scram channels, with their respective trip channels, are aligned to provide the "one-out-of-two taken twice" trip logic of the Reactor Protection System
.Source: Direct 73
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 35 21510 00 08/06/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0021202, Ability to monitor the effect of operating back panel switches.
Related Lessons COR0021202 INTERMEDIATE RANGE MONITOR Related Objectives NONE Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 215003.A4.04 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /
45.5 to 45.8) IRM back panel switches, meters, and indicating lights (3.1/3.3) 74
QUESTION: 35 21510 (1 point(s))
In order to support troubleshooting on IRM D, maintenance requires that IRM D mode switch be taken out of operate. None of the planned troubleshooting activities will generate an INOP signal.
What sequence of operation would prevent the generation of an IRM INOP signal?
- a. Depress and hold the INOP inhibit pushbutton then take the mode switch out of operate.
- b. Take the IRM mode switch out of operate then depress and hold the inop inhibit pushbutton.
- c. Momentarily depress the INOP inhibit pushbutton and then take the mode switch out of operate.
- d. Take the IRM mode switch out of operate then momentarily depress the inop inhibit pushbutton.
ANSWER: 35 21510
- a. Depress and hold the INOP inhibit pushbutton then take the mode switch out of operate.
Explanation:
Taking the mode switch out of operate generates an INOP signal unless the INOP inhibit pushbutton is depressed and held for the entire time the mode switch is out of operate.
Source: New 75
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 36 1994 01 07/09/2006 10/07/2006 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Systems COR0023002, SRM detector physical relationship with the RPV.
Related Lessons COR0023002 SOURCE RANGE MONITOR SUBSYSTEM Related Objectives COR0023002001050A Given an SRM control manipulation, predict changes in the following:
Detector position Related References (B)(2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
Related Skills (K/A) 215004.K1.06 Knowledge of the physical connections and/or cause- effect relationships between SOURCE RANGE MONITOR (SRM) SYSTEM and the following:
(CFR: 41.2 to 41.9 / 45.7 to 45.8) Reactor vessel (2.8/2.8) 76
QUESTION: 36 1994 (1 point(s))
Normal operation of the Source Range Monitor detector insert and retract mechanism positions the in-core detector over which of the following ranges?
- a. Bottom of active fuel to top of active fuel.
- b. Centerline of the core to the bottom of active fuel.
- c. 18 inches above the core centerline to 24 inches below the fuel region.
- d. 24 inches above core centerline to 24 inches outside the reactor pressure vessel.
ANSWER: 36 1994
- c. 18 inches above the core centerline to 24 inches below the fuel region.
The detector has a 10 ft. maximum travel between mechanical limits. The upper mechanical stop is 24 in. above the core midplane. The lower mechanical stop is 24 in. below the active fuel.
The detector has approximately a 9.5 ft. travel between the electrical limits (limit switch actuation). The upper electrical stop is 18 in. above the core midplane. This puts the "Full In" position at the same axial position as the centerline of the neutron sources. Therefore, the "Full In" position should be at the point of peak flux during shutdown conditions and while pulling toward criticality. The lower electrical stop is set just above the mechanical stop.
Distractors:
- a. is incorrect because the lower limit is 24" below the bottom of the core.
- b. is incorrect because the detector upper range is 24" above core midplane
- d. is incorrect because the detector is electrically limited to 18 inches above midplane.
REFERENCE:
Source Range Monitor Text Source: Direct 77
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 37 21398 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description Systems COR0020102, Ability to predict the effect of status of rod blocks from APRM Related Lessons COR0020102 AVERAGE POWER RANGE MONITOR Related Objectives COR0020102001070A Given a specific APRM malfunction, determine the effect on any of the following: Reactor Protection System (RPS)
COR0020102001070B Given a specific APRM malfunction, determine the effect on any of the following: Reactor Manual Control System (RMCS)
COR0020102001080A Describe the APRM design feature(s) and/or interlock(s) that provide for the following: Rod withdrawal blocks COR0020102001080B Describe the APRM design feature(s) and/or interlock(s) that provide for the following: Reactor SCRAM signals Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 215005.A1.03 Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including: (CFR: 41.5 / 45.5)
Control rod block status (3.6/3.6) 78
QUESTION: 37 21398 (1 point(s))
The reactor mode switch is in RUN, and reactor power is 15%. IRMs "B" and "H" failed upscale. Before any operator action is taken APRM "B" and "E" both fail downscale.
What is/are the minimum actions to clear all rod blocks and/or scrams?
- a. Bypass APRM "B".
- b. Bypass APRM "E".
ANSWER: 37 21398
Explanation: The upscale IRM B and the downscale APRM B generates a RPS trip on the "B" RPS. Both the downscale APRM "B" and "E" cause a rod block. To clear the RPS trip APRM "B" has to be bypassed. Which still leaves us with a rod block from the downscale APRM "E".
Therefore bypassing APRM E bypasses the rod block.
Distractors:
- a. is incorrect because a half scram also occurs and to clear all blocks and scrams requires that APRM "E" also be bypassed.
- b. is incorrect because the half scram is on RPS "B" and to clear all blocks and scrams requires that APRM "E" also be bypassed.
- c. is incorrect because APRM E would still be generating a rod block..
Source: Direct 79
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 38 21399 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0021802, Actions required by RCIC Annunciator Related Lessons COR0021802 OPS Reactor Core Isolation Cooling Related Objectives NONE Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.4.31 Knowledge of annunciators alarms and indications / and use of the response instructions. (CFR: 41.10 / 45.3) (3.3/3.4) 80
QUESTION: 38 21399 (1 point(s))
The plant is operating at power with HPCI out of service when a loss of feedwater occurred.
RCIC automatically initiated when reactor water level lowered to less than -42". Several minutes later the following indications were present:
% Reactor Water level is -45".
% RCIC flow is 250 gpm and steady.
% RCIC controller is in automatic.
% Annunciator 9-4-1/A-2 RCIC STEAM LINE HIGH D/P is alarming.
What operator action is required?
- a. Depress and Hold the RCIC TRIP pushbutton.
- c. Place RCIC TEST SWITCH to TEST and raise RCIC flow with the TEST POTENTIOMETER.
ANSWER: 38 21399
Explanation:
Annunciator 9-4-1/A-2 RCIC STEAM LINE HIGH D/P is alarming which indicates that RCIC steam line flow is at or above the high flow isolation setpoint. AP 9-4-1/a-2 lists the automatic action of closing RCIC-MO-15 and 16. Since these actions have not occurred and did not they should be taken by the operator.
Distractors:
- a. is incorrect because an isolation is required..
- b. is incorrect because indications of a steam leak are present and an isolation is required.
- c. is is incorrect because indications of a steam leak are present and an isolation is required.
Source: New 81
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 39 21494 00 07/29/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0021602, Knowledge of the ADS Design Feature that ensures adequate pneumatics to operate the ADS valves.
Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001080D Predict the consequences a malfunction of the following would have on the NPR system: Air/Nitrogen to ADS valves Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 218000.K4.04 Knowledge of AUTOMATIC DEPRESSURIZATION SYSTEM design feature (s) and/or interlocks which provide for the following: (CFR: 41.7)
Insures adequate air supply to ADS valves: Plant-Specific (3.5/3.6) 82
QUESTION: 39 21494 (1 point(s))
Immediately following a simultaneous loss of Nitrogen and a loss of instrument air how many times, if any, can an ADS valve be actuated from the pneumatics stored in its accumulator if drywell pressure is at atmospheric pressure? Assume that reactor pressure is and remains near normal operating pressure).
- a. None
- b. 2 times
- c. 5 times
- d. 14 times ANSWER: 39 21494
- c. 5 times\
Explanation:
In the event of a failure of the Instrument Air/Nitrogen Supply, the six relief valves associated with the ADS have accumulators with sufficient capacity to operate their respective valves two times at 70% design Drywell pressure. This equates to five operations at atmospheric pressure. The two relief valves associated with the LLS have larger accumulators with sufficient capacity to operate their respective valves fourteen times.
Distractors:
- a. is incorrect because the accumulators are protected by check valves and the loss of pneumatic supply will no result in the immediate depressurization of the accumulators.
- b. is incorrect because at atmospheric pressure in the drywell the valves can be cycled 5 times.
- c. is incorrect because this is the value for the LLS accumulators.
Source: New 83
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 40 5608 01 03/21/2003 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0021602, Knowledge of the power supply to ADS logic.
Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001020A State the electrical power supply to the following NPR components:
ADS logic COR0021602001080F Predict the consequences a malfunction of the following would have on the NPR system: D.C. power Related References 2.2.1 Nuclear Pressure Relief System (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
791E253 Automatic Blowdown System Related Skills (K/A) 218000.K2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) ADS logic (3.1*/3.3*)
84
QUESTION: 40 5608 (1 point(s))
An accident has occurred, resulting in the following conditions:
% Reactor pressure is 720 psig and lowering.
% RPV water level is -120" (WR) and stable.
% Drywell pressure is 6.2 psig and rising.
% 125 VDC panel AA2 is de-energized.
If present conditions continue, how will ADS respond?
ADS valves . . .
- a. fail to open due to loss of logic power.
- b. fail to open due to RPV water level conditions not met.
- c. are opened by the B logic circuit powered from its normal power source.
- d. are opened by both logic circuits powered from their alternate power sources.
ANSWER: 40 5608
- c. are opened by the B logic circuit powered from its normal power source.
Answer source: COR002-16-02, p. 21, & p. 22, section 3, p. 41 COR002-16-02 Figures 4 & 5 Distractors:
- a. ADS will initiate powered from BB2.
- b. ADS will initiate.
- d. ADS "A" has no alternate source and is de-energized.
Source: Direct 85
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 41 21495 00 07/28/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0020302, Ability to reset system isolation.
Related Lessons COR0020302 CONTAINMENT Related Objectives COR0020302001050A Describe the interrelationship between the Primary Containment system and the following: PCIS COR0020302001060O Describe the interrelationship between PCIS and the following:
Containment nitrogen inerting COR0020302001080C State the electrical power supplies to the following: PCIS logic power Related References (B)(9) Shielding, isolation, and containment design features, including access limitations.
Related Skills (K/A) 223002.A4.03 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /
45.5 to 45.8) Reset system isolations (3.6/3.5) 86
QUESTION: 41 21495 (1 point(s))
Given the following conditions:
% The plant is operating at 100% power with a Nitrogen Makeup to the Primary Containment in progress
% A loss of the "B" Reactor Protection System (RPS) Motor Generator Set occurs
% Torus N2 Supply Isolation Valve MO-1301 AND Drywell N2 Supply Isolation Valve MO-1312 closed
% RPS power has been restored Which of the following actions restore the Nitrogen makeup flowpath?
- a. Reset the group isolation signal with MO-1301 and MO-1312 to switches in OPEN.
- b. Positions MO-1301 and MO-1312 to switches to OPEN then reset the group isolation signal.
- c. Positions MO-1301 and MO-1312 to switches to CLOSE then to OPEN and then reset the group isolation signal.
- d. Reset the group isolation signal then place the MO-1301 and MO-1312 to switches to CLOSE then to OPEN.
ANSWER: 41 21495
- d. Reset the group isolation signal then place the MO-1301 and MO-1312 to switches to CLOSE then to OPEN.
EXPLANATION OF ANSWER:
After the isolation is reset, the control switches are positioned to close to reset the valve circuit and then to open to reposition the valve.
REFERENCE:
COR0020302 Distractors:
- a. is incorrect because the logic to open the valve is not reset until the switches are placed to close after the isolation is reset.
- b. is incorrect because the logic to open the valve is not reset until the switches are placed to close after the isolation is reset.
- c. is incorrect the valves will not open unless the switch is first positioned to close and then to open after the isolation is reset.
Bank Cognitive Level 1 87
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 42 18247 01 09/09/2003 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0021902, REC valve response with loss of NSST, SSST and no pumps in STANDBY Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001110C Given plant conditions, determine if any of the following should occur: Any REC valve automatic reposition Related References (B)(9) Shielding, isolation, and containment design features, including access limitations.
Related Skills (K/A) 223002.K1.19 Knowledge of the physical connections and/or cause- effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Component cooling water systems (2.7/2.9) 88
QUESTION: 42 18247 (1 point(s))
The plant is operating normally at 100% power with the following conditions:
% REC pumps 1A, 1B, and 1D are running.
% To support I&C repair efforts, NO REC pumps are in standby.
% REC heat exchanger A is in service.
A steam leak occurs in the drywell causing drywell pressure to rises and stabilize at +3.5 psig.
The reactor scrams, and immediately following the trip of the main turbine, a startup station service transformer lockout occurs.
Which REC valve is expected to be OPEN 2 minutes after the turbine trip?
- a. REC-MO-713, Heat Exchanger 1B Outlet.
- b. REC-MO-712, Heat Exchanger 1A Outlet.
- c. REC-MO-714, South Critical Loop Supply.
- d. REC-MO-1329, Augmented Radwaste Supply.
ANSWER: 42 18247
- c. REC-MO-714, South Critical Loop Supply.
Following the loss of all offsite power, 4160V bus 1F and1G are de-energized, resulting in a loss of power to MCC-S and MCC-K and the subsequent trip of all running REC pumps. As system pressure degrades following the trip of all REC pumps, REC-MO-712 and 713 are signaled to close on REC Heat exchanger outlet low pressure (61.5 and 59.5 psig) following a 40 second time delay, and REC-MO-1329 is signaled to close when the REC system supply header experience a low pressure of 60.5 psig following a 40 second time delay. Undervoltage on 4160V bus 1F and 1G will start DG-1 and DG-2, and signal breakers 1FS and 1GS to close to power bus 1F and 1G from the Emergency Station Service Transformer. After power is restored from the Emergency Station Service Transformer, all valves signaled to close on low pressure will do so, and the drywell high pressure condition (PCIS Group 6 channel B signal) will open REC-MO-714 after a 30 second time delay and the auto opening of the REC heat exchanger B Service Water outlet valve (SW-MO-651) to its minimum flow position.,
A. is incorrect. Because REC-MO-713 will close on low pressure.
B. is incorrect. Because REC-MO-712 will close on low pressure.
D. is incorrect. Because REC-MO-1329 will close on low pressure.
Source: Direct 89
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 43 21401 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0021602, Effect that depressing LLS logic pushbuttons has on LLS operation..
Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001030J Describe the interrelationships between the Nuclear Pressure Relief system and the following: RPS (low-low set initiation)
COR0021602001080E Predict the consequences a malfunction of the following would have on the NPR system: A.C. power Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 239002.A1.04 Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including: (CFR: 41.5 /
45.5) Reactor pressure (3.8/3.8) 90
QUESTION: 43 21401 (1 point(s))
The plant is operating at power when an inadvertent group 1 isolation occurs. The reactor scrams and RV-71D and RV-71F initially lift to control pressure. RV-71D then cycles lifting at 1015 psig and reseating at 875 psig.
If both LLS reset pushbuttons are depressed now, how does reactor pressure respond? (Assume the group 1 isolation is still present and no manual pressure control actions have occurred.)
Reactor pressure
- a. continues to cycle between 1015 psig and 875 psig.
- b. rises to approximately 1080 psig then cycles between 1015 and 875.
- c. rises to 1080 psig and then cycles between 1080 psig and 1030 psig.
- d. rises to 1100 psig and then cycles between 1100 psig and 1050 psig.
ANSWER: 43 21401
- b. rises to approximately 1080 psig then cycles between 1015 and 875.
Explanation:
Following the initial transient LLS armed and commenced controlling reactor pressure. The loss of offsite power would have deenergized both RPS systems. The loss of RPS prevents the resetting of the LLS logic.
Distractors:
- a. is incorrect because the logic would reset and would not arm again until pressure reached 1080 psig and the lower set relief opened and rearmed LLS.
- c. is incorrect because when RV-71D lifts LLS is rearmed and pressure cycles between 1015 an 875 psig.
- d. is incorrect because RV-71D will lift before pressure reaches 1100 psig.
Source: New 91
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 44 21402 00 06/22/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 2 1 3 Multiple Choice Topic Area Description Systems INT0320135, Ability to perform actions immediately required without reference to procedures.
Related Lessons INT0320135 CNS Abnormal Procedures (RO) - Condensate/Feedwater Related Objectives INT0320135H0H0100 Given plant condition(s), determine from memory any automatic actions listed in the applicable Abnormal/Emergency Procedure(s) which will occur due to the event(s).
INT0320135I0I0100 Given plant condition(s), determine from memory all immediate operator actions required to mitigate the event(s).
Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10 / 43.2
/ 45.6) (4.0/4.0) 92
QUESTION: 44 21402 (1 point(s))
The plant is operating at rated power when an inadvertent scram occurs. The following conditions are now present:
% RFC-LC-83, MASTER LEVEL CONTROLLER is set at 15 inches.
% Reactor MODE SWITCH is in SHUTDOWN.
% Pressure set is at 926 psig and BPVs are maintaining pressure.
% RFPT control is in AUTOMATIC.
% Reactor water level is at 59 inches (NR) and slowly rising.
% Both RFPs are running.
What action is immediately required?
- a. Trip one RFPT.
- b. Trip both RFPTs.
- c. Place RFPT Controllers RFC-MA-84A/B to MAN.
- d. Lower RFC-CS-SUMAST, STARTUP MASTER CONTROL LEVEL SETPOINT to 15".
ANSWER: 44 21402
- b. Trip both RFPTs.
Reactor water level is greater than the high trip point for the RFPTs and therefore both should be tripped.
Distractors:
a,c and d. are all incorrect as any action other than a turbine trip is inappropriate.
Source Direct Production TM 21268 93
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 45 21403 00 06/23/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0022802, Effect a malfunction of SGT has on Secondary Containment Temperature.
Related Lessons COR0022802 OPS STANDBY GAS TREATMENT Related Objectives NONE Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 261000.K3.01 Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: (CFR: 41.7 /45.6)
Secondary containment and environment differential pressure (3.3/3.6) 94
QUESTION: 45 21403 (1 point(s))
An accident occurs that involves a steam leak in the reactor building. The following conditions exist:
% Both SGT trains started on an automatic initiation signal.
% SGT fan 1F was placed to OFF then placed in STANDBY.
% Reactor building differential pressure is -0.5" H2O.
% SGT train 1E flow is 1300 cfm.
A failure of the vortex dampers for SGT 1E Fan causes them to slowly close. No operator action occurs.
How is reactor building differential pressure affected by this failure?
Reactor building Differential pressure falls
- a. to and remains at approximately 0" H2O.
- b. to 0" H2O then increases and stabilizes with a positive D/P.
- c. until reactor building differential pressure is 0.0" H2O then recovers to approximately -0.5" H2O.
- d. until train flow drops to less than 800 cfm then reactor building d/p recovers to approximately -0.5" H2O.
ANSWER: 45 21403
- d. until train flow drops to less than 800 cfm then reactor building d/p recovers to approximately -0.5" H2O.
Explanation:
At 800 cfm the SGT in Standby starts and recovers pressure to approximately the value before the failure of the vortex damper.
Distractors:
- a. is incorrect because the standby train starts at 800 cfm and recovers pressure. The candidate that believes a trip of the running fan causes the start of the standby train would choose this answer.
95
- b. is incorrect because the standby train starts at 800 cfm and recovers pressure. The candidate that believes reactor building pressure would naturally be positive and that believes that the standby train would fail to start would choose this answer.
- c. is incorrect because the standby train starts at 800 cfm and restores reactor building D/P.
Source: Direct PTM 3817 96
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 46 16489 01 04/14/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0022802, Knowledge of the cause effect relationship between SGT and the process radiation monitoring system.
Related Lessons COR0022802 OPS STANDBY GAS TREATMENT Related Objectives COR0022802001080A Describe the Standby Gas Treatment design features and/or interlocks that provide for the following: Automatic system initiation COR0022802001130A Given plant conditions, determine if any of the following should occur: SGT automatic initiation Related References (B)(11) Purpose and operation of radiation monitoring systems, including alarms and survey equipment.
Related Skills (K/A) 261000.K1.08 Knowledge of the physical connections and/or cause- effect relationships between STANDBY GAS TREATMENT SYSTEM and the following:
(CFR: 41.2 to 41.9 / 45.7 to 45.8) Process radiation monitoring system (2.8/3.1) 97
QUESTION: 46 16489 (1 point(s))
The plant is at full power when, annunciator 9-4-1/E-4, RX BLDG VENT HI-HI RAD alarms.
Radiation Monitor readings are:
% RMP-RM-452A: 14 mrem/hr
% RMP-RM-452B: 12 mrem/hr
% RMP-RM-452C: 8 mrem/hr
% RMP-RM-452D: 7 mrem/hr Which one of the following is the effect on the Secondary Containment and why?
(Note: Use actual setpoints.)
- a. NOT affected because only the DIVISION I logic has actuated.
- b. NOT affected because only the DIVISION II logic has actuated.
- c. Isolates and both SGT systems initiate. There is a start signal from both Divisions.
- d. Isolates but only "A" SGT initiates. There is NO start signal from one Division.
ANSWER: 46 16489
- c. Isolates and both SGT systems initiate. There is a start signal from both Divisions.
Justification: If RMP-RM-452A or C AND RMP-RM-452B or D exceed 10 mrem/hr, Reactor Building isolates, and the SGT system starts.
REFERENCE:
2.2.73; 1.3.1.2 (logic) 2.3_9-4-1; Set Points Distracter a: Both Divisions are actuated. The reactor building will isolate and SGT starts.
Distracter b: Both Divisions are actuated. The reactor building will isolate and SGT starts.
Distracter d: Both Divisions are actuated. The reactor building will isolate and both trains of SGT will start.
Source: Direct 98
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 47 21404 00 06/23/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 2 1 3 Multiple Choice Topic Area Description Systems COR0010102, Principle involved in paralleling AC sources.
Related Lessons COR0010102 AC Electrical Distribution Related Objectives COR0010102001100A Briefly describe the following concepts as they apply to AC Electrical Distribution System: Principle involved with paralleling two AC sources Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 262001.K5.01 Knowledge of the operational implications of the following concepts as they apply to A.C. ELECTRICAL DISTRIBUTION: (CFR: 41.5 / 45.3) Principle involved with paralleling two A.C. sources (3.1/3.4) 99
QUESTION: 47 21404 (1 point(s))
DG2 is running unloaded. The crew intends to parallel DG2 with 4160 1G.
% DG voltage is 4050 VAC.
% 4160 1G voltage is 4160 VAC.
SYNCH SWITCH is placed to EG2 and the synchroscope is rotating slowly in the COUNTER -CLOCKWISE direction.
What DG2 speed and voltage adjustments if any are required before closing EG2?
- a. Raise speed and lower voltage.
- b. Raise speed and raise voltage.
- c. No speed adjustment required and lower voltage.
- d. No speed adjustment required and raise voltage.
ANSWER: 47 21404
- b. Raise speed and raise voltage.
Explanation:
DG speed is less than synchronous speed so speed should be raised. DG voltage is lower than running so voltage should be raised.
Distractors:
- a. is incorrect because voltage should be raised.
- c. is incorrect because speed should be raised and voltage lowered.
- d. is incorrect because speed should be raised.
Source: Direct Production TM 21322 100
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 48 17798 01 06/21/2002 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0010102, Knowledge of the design features that transfer NBPP from preferred power to alternate power.
Related Lessons COR0010102 AC Electrical Distribution Related Objectives COR0010102001060C Describe the interrelationship between the AC Electrical Distribution System and the following: No Break Power Supply Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 262002.K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.): (CFR: 41.7 /
45.7) A.C. electrical power (2.7/2.9) 101
QUESTION: 48 17798 (1 point(s))
Plant operating normally at 100% power when annunciator "C-1/A-1, 250 VDC Bus 1A Blown Fuse" alarmed. The CRT alarm message indicates "(3705) Static Inverter 1A feeder".
Then, while investigating the above alarm, annunciators "C-2/C-1, 4160V BUS 1A BKR 1AN lockout" and "C-1/B-6, 4160V BUS 1F BKR 1FA TRIP" alarm.
Which of the following describes the expected impact on the NBPP?
The NBPP is . . .
- a. deenergized.
- b. energized by DG-1.
- c. energized from the inverter.
- d. energized by the Emergency Transformer.
ANSWER: 48 17798
- d. energized by the Emergency Transformer.
References:
5.3NBPP Justification: Static Inverter 1A loses power from 250 VDC Switchgear 1A due to the blow fuse.
However, when 4160V BUS 1A is deenergized undervoltage on 4160V bus 1F signals breaker 1FS to close, powering bus 1F from the Emergency Station Service Transformer; thus, the alternate AC source to the NBPP from MCC-R is available from 480V switchgear 1F and MCC-K which are powered from 4160V BUS 1F.
Foils: a. is incorrect. Because the NBPP is energized from the alternate AC source. b. is incorrect. Because DG-1 will start on 4160V bus 1F undervoltage; however, breaker EG1 will NOT close automatically unless breaker 1FS fails to close. c. is incorrect. Because Static Inverter 1A has lost power from 250 VDC Switchgear 1A due to the blow fuse.
Source: Direct 102
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 49 12518 00 03/07/2001 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0020702, Ability to monitor alarms and determine system status.
Related Lessons COR0020702 OPS DC ELECTRICAL DISTRIBUTION Related Objectives COR0020702001080D Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following: Battery chargers Related References 3058 DC One Line Diagram 2.2.25.1 125 VDC Electrical System (Div 1)
(B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 263000.A3.01 Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including: (CFR: 41.7 / 45.7) Meters, dials, recorders, alarms, and indicating lights (3.2/3.3) 103
QUESTION: 49 12518 (1 point(s))
The plant is at 100% power with the 125 VDC electrical distribution system aligned for normal lineup. The following occur:
% Annunciator C-4/C-7, 125 VDC BATT CHARGER 1B TROUBLE alarms
% CRT alarm message indicates:
(3765) 125V DC BATTERY CHARGER 1B DC VOLTAGE HIGH (in and reset)
(3762) 125V DC BATTERY CHARGER 1B AC VOLTAGE FAILURE.
(3764) 125V DC BATTERY CHARGER 1B DC VOLTAGE LOW What is the position (open or closed) of the 125V charger 1B AC input and DC output breakers?
The AC input breaker is
- a. open and the DC output breaker is closed.
- b. open and the DC output breaker is open.
- c. closed and the DC output breaker is closed.
- d. closed and the DC output breaker is open.
ANSWER: 49 12518
- a. open and the DC output breaker is closed.
The AC input breaker has tripped open due to battery charger 1B DC voltage high. DC output over voltage causes the AC input breaker on a 125V CHARGER to trip. The DC output breaker does NOT automatically trip open. Neither breaker automatically trips open due to loss of AC power to the chargers.
Distractors:
- b. is incorrect because the DC breaker is closed.
- c. is incorrect as the AC breaker opens.
REFERENCE:
ALARM PROCEDURE 2.3_C-4, 2.2.25.2 Source: Direct 104
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 50 21405 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0020702, Knowledge of DC breaker Interlocks Related Lessons COR0020702 OPS DC ELECTRICAL DISTRIBUTION Related Objectives COR0020702001090B Describe the DC Electrical Distribution System design feature(s) and/or interlock(s) that provide for the following: Breaker interlocks, permissives, bypasses and crossties Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 263000.K4.01 Knowledge of D.C. ELECTRICAL DISTRIBUTION design feature (s) and/or interlocks which provide for the following: (CFR: 41.7) Manual/
automatic transfers of control: Plant- Specific (3.1/3.4) 105
QUESTION: 50 21405 (1 point(s))
What interlocks/physical controls exist for the HPCI Starter Rack that prevent the HPCI Starter Rack from being supplied simultaneously from its normal and emergency source?
- a. Transfer switch design prevents simultaneous supply from both sources.
- b. Electrical interlock prevents the simultaneous closure of both supply breakers.
- c. Mechanical interlock prevents the simultaneous closure of both supply breakers.
- d. Administrative control of a padlock on the breakers prevents simultaneous closure of both supply breakers.
ANSWER: 50 21405
- a. Transfer switch design prevents simultaneous supply from both sources.
Explanation:
The transfer switch prevents the simultaneous supply of the HPCI SR from both sources. The transfer switch can only be aligned to one source at a time.
- b. is incorrect because no electrical interlock on the breakers prevents their simultaneous closure. During normal transfer both breakers are simultaneously closed
- c. is incorrect because no mechanical interlock on the breakers prevents their simultaneous closure. During normal transfer both breakers are simultaneously closed.
- d. is incorrect because the padlock is used to prevent the inadvertent transfer to the alternate source. During transfer the padlock is removed and both breakers are closed.
Source: New 106
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 51 21145 00 09/02/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Systems COR0020802, Voltage Regulator Failure Effect on ECCS Pumps.
Related Lessons COR0020802 DIESEL GENERATORS Related Objectives COR0020802001080A Given a specific Diesel Generator malfunction, determine the effect on any of the following: Emergency Core Cooling Systems Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 264000.K3.01 Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: (CFR: 41.7 / 45.4)
Emergency core cooling systems (4.2*/4.4*)
107
QUESTION: 51 21145 (1 point(s))
The plant was operating when a loss of offsite power and a LOCA occurred.
% Both diesel generators initially start and tie to their respective busses as designed.
% Immediately after EG1 automatically closed with DG voltage at 4160 Volts, DG1 voltage regulator fails causing the generator voltage to lower continuously at 50 volts/second.
If this trend continues at its current rate, how are the RHR A, RHR B and CS A pumps affected?
RHR A, RHR B and CS A pump breakers...
- a. open 41 seconds after EG1 closed.
- b. open as soon as voltage drops to 3880 volts.
- c. remain closed until DG lockout occurs due to loss of field.
- d. pump breakers remain closed irrespective of generator status and bus voltage.
ANSWER: 51 21145
- a. open 41 seconds after EG1 closed.
Lockout of load-shedding (blocking the trip function of motor breakers due to undervoltage) on the critical bus occurs if the off-site power source is unavailable and the bus is energized from its Diesel Generator. This will preclude spurious undervoltage trips for 41 seconds if diesel generator output voltage drops during sequential loading or due to energizing a loaded bus. The load shedding feature will be reinstated 41 seconds after EG1(EG2) close (to allow completion of the load-sequencing action), but will only recur on first level (2300 VAC) undervoltage conditions.
Source: Direct 108
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 52 21406 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems COR0011701, Ability to monitor control room pressure indication and determine system alignment.
Related Lessons COR0011701 OPS Plant Air COR0011702 Plant Air Related Objectives COR0011701001060A Describe the operation of the interlocks associated with the following components in the Plant Air System: Station Air Compressors COR0011702001060D Predict the consequences the following would have on the Plant Air System: Leak in system Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 300000.A4.01 Ability to manually operate and / or monitor in the control room: (CFR: 41.7
/ 45.5 to 45.8) Pressure gauges (2.6/2.7) 109
QUESTION: 52 21406 (1 point(s))
The Plant is operating at power with the Plant Air compressors operating in the CCM local mode with only one compressor initially running. The following indication is observed by the operator:
PI-606 (Instrument Air Header Supply Pressure) is observed to initially lower to 65 psig, then rise to 85 psig and stabilizes.
- a. Only the lead compressor is loaded and running.
- b. The lead and the next compressor are loaded and running.
- c. The lead, next and lag compressors are running and loaded.
ANSWER: 52 21406
- c. The lead, next and lag compressors are running and loaded.
Explanation:
If operating in CCM local mode the lead compressor is set at 100 - 110 psi. The next is set at 95
-105 psi, and last at 90-100 psi. Since the air leak resulted in decreased system pressure the next and lag compressors started and loaded. Since pressure never rose to greater than the setpoint for the lag compressor all remain on and loaded.
Distractors:
- a. is incorrect because pressure fell to less than that required to start and load the other compressors.
- b. is incorrect because the lag compressor is also running.
- d. is incorrect because all are loaded and running.
Source: New 110
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 53 21407 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 3 1 5 Multiple Choice Topic Area Description Systems COR0021902, Loss of REC surge tank level and system automatic actions.
Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001040A Describe the REC design features and/or interlocks that provide for the following: Repositioning of REC Supply Valves to components COR0021902001050A Briefly describe the following concepts as they apply to REC: Leak or lowering system pressure during accident and transient conditions COR0021902001040D Describe the REC design features and/or interlocks that provide for the following: Isolation of Non-Critical Cooling loops COR0021902001110A Given plant conditions, determine if any of the following should occur: Non-Critical loop isolation COR0021902001110B Given plant conditions, determine if any of the following should occur: Standby pumps automatic start COR0021902001110C Given plant conditions, determine if any of the following should occur: Any REC valve automatic reposition Related References (B)(4) Secondary coolant and auxiliary systems that affect the facility.
Related Skills (K/A) 400000.A2.02 Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: (CFR: 41.5 / 45.6) High/low surge tank level (2.8/3.0) 111
QUESTION: 53 21407 (1 point(s))
The plant is shutdown with B RHR in shutdown cooling, when the following indications/reports are received:
% M-1/A-1, REC SYSTEM LOW PRESSURE
% M-1/A-3, REC SURGE TANK LOW LEVEL
% Report from the Reactor building Station Operator that a large rupture exists on the REC Supply Header at the Tee to REC-MO-1329, AUGMENTED RADWASTE SUPPLY.
If these conditions persist for the next few minutes, what loads, if any, remain lined up to REC?
(Assume no group 6 isolation signal is present and no operator actions are taken.)
How is the REC system aligned (either automatically or manually) for these conditions?
- a. No loads are currently lined up.
Supply critical loops with REC.
- b. Only South Critical loop is currently lined up.
Supply critical loops with Service Water.
- c. No loads are currently lined up.
Supply critical loops with Service Water.
- d. Only South Critical loop is currently lined up.
Supply critical loops with REC.
ANSWER: 53 21407
- d. Only South Critical loop is currently lined up.
Supply REC critical loops with a single REC pump.
Explanation:
The loss of REC pressure will cause REC-MO-700, REC-MO-702, REC-MO-712, REC-MO-713, and REC-MO-1329 to close this isolates all non-critical REC loads. Since RHR B is in service the South critical loop will remain lined up to REC since there is no auto closure of the critical loop valves.
Distractors:
- a. is incorrect because the south critical loop remains aligned to REC.
112
- b. is incorrect because the leak is/will be isolated by the automatic action of REC-MO-700, REC-MO-702, REC-MO-712, REC-MO-713, and REC-MO-1329 to closure and now a single REC pump is started to supply the intact critical loops.
- c. is incorrect because the south loop remains aligned to REC and SW would not be used in this condition to supply the critical loops.
Source: New 113
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 54 21408 00 06/27/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0020402, Ability to monitor automatic alarms in the CRD hydraulic system.
Related Lessons COR0020402 CONTROL ROD DRIVE HYDRAULICS Related Objectives COR0020402001040F Describe the CRDH system design features and/or interlocks that provide for the following: Isolation of scram discharge volumes during scram conditions Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 201001.A3.10 Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including: (CFR: 41.7/45/7) Lights and alarms (3.0/2.9) 114
QUESTION: 54 21408 (1 point(s))
The plant is operating when the south scram discharge volume drain valve fails closed. Leakage past the seat for the scram outlet valve that discharges to the south volume cause level in the volume to increase.
As the level in the scram discharge volume rises to 60 " what alarms/automatic actions occur as level rises to 60"?
- a. Rod Block ONLY.
- b. South SDIV NOT DRAINED Alarm ONLY.
- c. South SDIV NOT DRAINED Alarm and rod block ONLY.
ANSWER: 54 21408
- c. South SDIV NOT DRAINED Alarm and rod block ONLY.
Explanation:
The SDIV not drained alarm comes in at 11.5" and the rod block occurs at 46". Therefore by the time level reaches 60" the alarm and the rod block are in.
Distractors:
- a. is incorrect because the SDIV NOT Drained alarm is also in.
- b. is incorrect because a rod block is also present.
- d. is incorrect because a scram signal is not present.
Source: New 115
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 55 2109 00 08/12/1999 10/07/2006 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0022002001060B Reactor Manual Control System and Rod Position Indication System Related Lessons COR0022002 OPS REACTOR MANUAL CONTROL SYSTEM Related Objectives COR0022002001060B Given a RMCS control manipulation, predict and explain the response of the following: Rod movement sequence lights Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 201002.A1.03 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR MANUAL CONTROL SYSTEM controls including: (CFR: 41.5 / 45.5) Rod movement sequence lights (3.0/2.9) 116
QUESTION: 55 2109 (1 point(s))
Select the sequence in which the following Reactor Manual Control System lamps are expected to be illuminated as a control rod is withdrawn one notch.
Rod
- a. Out Permit Out Settle
- b. Out Permit In Out Settle
- c. In Out Permit Out Settle
- d. Out Permit In Settle Out Settle ANSWER: 55 2109
- b. Out Permit In Out Settle
REFERENCE:
Reactor Manual Control System Text
\If there are no RWM or other rod withdraw blocks, withdrawal motion is selected and sealed in for one timer cycle, allowing Unlatch (drive in light), Drive Out(out light) and Settle (Settle light).
Distractors:
- a. is incorrect because the drive in is first in sequence to unlatch the rod.
117
- c. is incorrect because the out permit is expected to be energized from the outset.
- d. is incorrect because there is no settle between the in and out.
Source: Direct 118
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 56 3987 00 11/04/1999 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 2 1 4 Multiple Choice Topic Area Description Systems COR0020502001120C Control Rod Drive Mechanisms Related Lessons COR0020502 CONTROL ROD DRIVE MECHANISM Related Objectives COR0020502001120C Determine the interrelationships between the CRDMs and the following: Rod Position Indicating System Related References 2.4CRD CRD Trouble Related Skills (K/A) 201003.A4.02 Ability to manually operate and/or monitor in the control room: (CFR: 41.7
/ 45.5 to 45.8) CRD mechanism position: Plant-Specific (3.5/3.5) 119
QUESTION: 56 3987 (1 point(s))
Rod 38-27 is being continuously withdrawn from position 38 to position 48.
Which statement below describes the condition that will indicate rod 38-27 is uncoupled?
- a. Rod position indication does NOT change when the rod is withdrawn.
- b. The Red Full Out light on the Full Core Display is extinguished AND indicated position on the Four-Rod display is 48.
- c. The Green Over-travel light on the Full Core Display is illuminated AND indicated position on the Four-Rod display is 48.
- d. Position indication is lost (goes blank) on the Four-Rod display AND the PMIS Computer indicates 99.
ANSWER: 56 3987
- d. Position indication is lost (goes blank) on the Four-Rod display AND the PMIS Computer indicates 99.
EXPLANATION OF ANSWER: d. Correct. Since the reed switch at position 52 (Over-travel Out) only provides input to the Over-travel alarm, indicated position goes blank. a. Reed switches are actuated by magnets in the drive piston which operates properly. b. Since the reed switch at position 52 (Over-travel Out) only provides input to the Over-travel alarm, indicated position goes blank. c. The Green light on the Full Core Display is for Full In. Since the reed switch at position 52 (Over-travel Out) only provides input to the Over-travel alarm, indicated position goes blank.
REFERENCE:
STCOR0020502 Control Rod Drive Mechanism Rev 7; PR 2.4.1.1.2 Uncoupled Control Rod Page 1 Section 1 Rev 8 Source: Direct 120
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 57 21410 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0022202, Predict the effect of a RR pump signal failure and determine the corrective actions.
Related Lessons COR0022202 REACTOR RECIRCULATION Related Objectives NONE Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 202001.A2.06 Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate...: (CFR: 41.5 / 45.6) Inadvertent recirculation flow decrease (3.6/3.8) 121
QUESTION: 57 21410 (1 point(s))
The plant is operating at 96% power with the following conditions:
% BOTH Recirculation Pump MG sets are operating at 90% speed
% A failure in the "B" Recirculation Pump Flow control circuit results in a demanded pump speed signal to the scoop tube positioner of 10%.
- 1) If no operator actions are taken what is the impact on B RRMG set speed?
- 2) What immediate operator action is required to mitigate the impact of this failure?
- a. 20%
Lock the B RRMG set scoop tube.
- b. 45%
Lock the B RRMG set scoop tube.
- c. 20%
Attempt to stabilize flow with RRFC-SIC-16B.
- d. 45%
Attempt to stabilize flow with RRFC-SIC-16B.
ANSWER: 57 21410
- a. 20%
Lock the B RRMG set scoop tube.
Explanation:
The speed control signal failure would reduce RRMG set speed to minimum of 20% where electrical and mechanical stops would prevent further decrease. 2.4RR immediate operator action requires the scoop tube Lock out.
Distractors:
- b. is incorrect because with no operator action speed would decrease to 45%.
- c. is incorrect because the only appropriate operator response is a scoop tube lockup..
- d. is incorrect because with no operator action speed would decrease to 20% and the only appropriate operator response is a scoop tube lockup.
Source: New 122
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 58 7759 00 05/29/2000 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Systems COR0022202, Knowledge of the interlocks associated with RR pump Speed.
Related Lessons COR0022202 REACTOR RECIRCULATION Related Objectives COR0022202001100L Describe the Reactor Recirculation system and/or Recirculation Flow Control system design features and/or interlocks that provide for the following: Recirculation Pump Runback COR0022202001130B Given plant conditions, determine if any of the following should occur: Recirculation pump runback to the dual speed limiter.
Related References (B)(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Related Skills (K/A) 202002.K4.02 Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature (s) and/or interlocks which provide for the following: (CFR: 41.7)
Recirculation pump speed control: Plant-Specific (3.0/3.0) 123
QUESTION: 58 7759 (1 point(s))
The plant was operating at 80% power when 'A' RFP tripped. RPV water level lowered to 25"(NR) prior to recovering.
What is the automatic response of the Reactor Recirculation System?
- a. RR pump speed lowers to 20%.
- b. RR pump speed lowers to 45%.
- c. RR pump speed remains constant.
ANSWER: 58 7759
- b. RR pump speed lowers to 45%.
RR pumps run back to 45% with one feed pump tripped and level < 27.5 inches.
- a. is incorrect. This limiter requires < 20% total feedwater flow.
- c. is incorrect since the 45% limiter is activated.
- d. is incorrect this would occur if level reached -42 inches.
Source: Direct 124
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 59 21411 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0012002, Action to prevent overpressure of the RWCU system.
Related Lessons COR0012002 OPS Reactor Water Cleanup Related Objectives COR0012002001090D Describe the RWCU design features and/or interlocks that provide for the following: Piping over-pressurization protection COR0012002001130G Given a RWCU component manipulation, predict and explain the changes in the following parameters: RWCU system pressure Related References (B)(9) Shielding, isolation, and containment design features, including access limitations.
(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10 / 43.2
/ 45.6) (4.0/4.0) 125
QUESTION: 59 21411 (1 point(s))
The plant was operating at power when a RWCU isolation (Group 3) occurred.
Which of the following actions would the operator take to prevent overpressurization of the RWCU system?
- a. Return Isolation Valve, MO-68 is cracked open.
- b. Blowdown Flow Control Valve PCV-55 is closed.
- c. Demin Suction Bypass Valve MO-74 is cracked open.
- d. Drain Valve to Radwaste System MO-57 and Drain Valve to the Condenser MO-56 are both cracked open.
ANSWER: 59 21411
- c. Demin Suction Bypass Valve MO-74 is cracked open.
Following a RWCU isolation Procedure 2.1.22 requires that MO-74 be cracked open to prevent over pressurization by mini-purge. CRD purge of RWCU Pump seals can over pressurize the pump and piping following closure of MO-15 or MO-18. Opening MO-74 provides a path for CRD flow around the demins to the Reactor Vessel.
Distractors:
- a. MO-68 should already be open and this valve alone would not provide overpressure protection from mini-purge following isolation because a path around the now out of service demineralizers is required.
- b. This valve should already be closed, in addition its closure would do nothing to prevent over pressurization of the RWCU piping. FCV-55 closes to protect downstream piping from high pressure or upstream piping from low pressure.
- d. These valves should not be opened simultaneously as this could result in a loss of vacuum.
Source: Production TM 19675 126
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 60 16471 00 07/25/2001 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 3 1 4 Multiple Choice Topic Area Description Systems COR0021502, Knowledge of the effect that a loss of NBI instruments have on the MT.
Related Lessons COR0021502 NUCLEAR BOILER INSTRUMENTATION Related Objectives COR0021502001020F Describe the interrelationships between NBI and the following: Main Turbine/Feedwater COR0021502001040D Briefly describe the following concepts as they apply to NBI: Vessel DP measurement COR0021502001050A Predict the consequences of the following on the NBI: Detector equalizing valve leaks Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 216000.K3.16 Knowledge of the effect that a loss or malfunction of the NUCLEAR BOILER Instrumentation will have on following: (CFR: 41.7 / 45.4) Main turbine (3.0/3.1) 127
QUESTION: 60 16471 (1 point(s))
The plant is at 100% power when the equalizing valve for NBI-LT-52A (Narrow Range Reactor Water level instrument) is accidentally fully opened by I&C. The instrument was NOT isolated prior to opening the equalizing valve.
Assume NO operator actions are taken.
What effect will this have on plant operation?
- a. The RFPs and the Main Turbine will trip.
- b. Only a low reactor water level alarm is received.
- c. Only a high reactor water level alarm is received.
ANSWER: 60 16471
- a. The RFPs and the Main Turbine will trip.
Two of three instruments are required to satisfy the logic. When the "A" level instrument is equalized, it drains the reference leg. "A" and "C" narrow range instruments share a common reference leg, so both instruments would be affected and read higher than actual level. The two high level trip signals are actuated causing the RFPs and Main Turbine to trip.
REFERENCE:
4.6.1; Attachment 2 - 1.2.1, 1.2.4.3, and 2.1 Distracter b: A full scram is received.
Distracter c: A high level trip occurs.
Distracter d: A full scram is received.
Source: Direct 128
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 61 21412 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: Y Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0010602, Knowledge of the relationship between handling the fuel shipping cask and fuel pool ventilation.
Related Lessons COR0010602 FUEL POOL COOLING AND DEMINERALIZING SYSTEM Related Objectives COR0010602001050F Describe the interrelationship between the FPC system and the following: Reactor Building Ventilation COR0010602001120D Given a Fuel Pool Cooling component manipulation, predict and explain the changes in the following parameters: Fuel Pool level COR0012102001020E Describe the interrelationship between reactor refueling and servicing equipment and the following: Spent fuel cask COR0012102001030A Given a Reactor Refueling and Servicing Equipment manipulation, predict and explain the changes in the following parameters: Spent fuel pool level Related References (B)(13) Procedures and equipment available for handling and disposal of radioactive materials and effluents.
Related Skills (K/A) 234000.K1.09 Knowledge of the physical connections and/or cause- effect relationships between FUEL HANDLING EQUIPMENT and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Fuel pool ventilation: Plant-Specific (2.8/2.9) 129
QUESTION: 61 21412 (1 point(s))
Why is the fuel shipping cask required to be lowered into the fuel pool at such a slow rate?
- a. Protect reactor building integrity by limiting velocity of the cask.
- b. Protect fuel storage racks by limiting the initial velocity in the event the fuel shipping cask is dropped.
- c. Prevents flooding the fuel pool ventilation ducts by limiting the rate of displacement of fuel pool water.
- d. Prevent loss of visibility and limits refuel floor radiation levels by reducing the disturbance of particulate in the fuel pool.
ANSWER: 61 21412
- c. Prevents flooding the fuel pool ventilation ducts by limiting the rate of displacement of fuel pool water.
Explanation:
The empty shipping cask is lowered slowly into the fuel pool, to assure that fuel pool ventilation ducts are not flooded. This ensures that the water can be processed fast enough to keep from flooding the ducts.
At Cooper Nuclear Station a Reactor Operator can be in charge of this evolution on the refueling floor. This is not a refueling evolution and does not involve the movement of Special Nuclear Materials. It does involve interaction with plant systems, ie fuel pool cooling and the ventilation system. Therefore no 10CFR55.43(b) topic can be linked to this question. However this valid for an RO and can be related to 10CFR55.41(b)13.
Distractors:
- a. is incorrect as this is actually the reason for the restricted path mode of operation. Which limits crane movement to 18 fpm and to a path over structural members, but doesn't restrict the rate the cask is lowered into the pool.
- b. is incorrect this is why the cask cannot be lifted over the fuel racks.
- d. is incorrect although this occurrence is possible it is unlikely and the rational behind the reduced lowering rate is just to limit the rate of displacement to prevent flooding the ventilation ducts.
Source: New 130
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 62 1651 00 02/28/2003 10/07/2006 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems COR0010802, HEATING, VENTILATION AND AIR CONDITIONING Related Lessons COR0010802 OPS HEATING, VENTILATION AND AIR CONDITIONING Related Objectives COR0010802001070E Describe the interrelationship between the Control Room HVAC and the following: Fire protection COR0010802001120A Describe the control Room HVAC design features and interlocks that provide for the following: Control room HVAC reconfigurations COR0010802001140A Briefly describe the following concepts as they apply to Control Room HVAC: Airborne contamination (e.g., radiological,toxic gas, smoke) control COR0010802001160D Predict the consequences a malfunction of the following would have on the Control Room HVAC system: Fire protection COR0010802001200D Predict the consequences of the following items on the Control Room HVAC: Initiation/failure of fire protection system Related References COR0010802 Heating, Ventilation, and Air Conditioning 2.2.84 HVAC Main Control Room and Cable Spreading Room (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 290003.K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROOM HVAC: (CFR: 41.7 / 45.7) Fire protection: Plant-Specific (2.6/2.8) 131
QUESTION: 62 1651 (1 point(s))
What automatic actions occur when Smoke Detector SD-1001 associated with the Main Control Room ventilation trips?
- a. Running Supply Fan trips, and all Fire/Smoke dampers close.
- b. Exhaust Fans BF-C-1B and EF-C-1B trip, and all Fire/Smoke dampers close.
- c. Running Supply, Exhaust, and Booster Fans trip, and all Fire/Smoke dampers close.
- d. Emergency Booster Fan BF-C-1A starts, damper HV-270-AV closes, and HV-271-AV opens.
ANSWER: 62 1651
- a. Running Supply Fan trips, and all Fire/Smoke dampers close.
When actuated, a smoke detector (SD-1001) located in the cable spreading room exhaust duct will trip off the supply fans and close fire smoke dampers AD-1544, AD-1545, AD-1546, AD-1547, AD-1581, and AD-1582.
- b. is incorrect because the exhaust fans do not trip.
- d. is incorrect because this is not an emergence booster fan start signal.
- c. is incorrect because only the supply fans trip.
Source: Direct 132
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 63 21414 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 2 1 3 Multiple Choice Topic Area Description Systems COR0020202, Ability to monitor feedwater system valves in the CR.
Related Lessons COR0020202 OPS CONDENSATE AND FEED Related Objectives COR0020202001120C Given plant conditions, determine if: Minimum Flow Valves should have repositioned Related References (B)(4) Secondary coolant and auxiliary systems that affect the facility.
Related Skills (K/A) 259001.A4.04 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /
45.5 to 45.8) System valves (3.1/2.9) 133
QUESTION: 63 21414 (1 point(s))
RFP 1A is being placed in service with flow at 3000 gpm. The minimum flow valve is closed.
A failure in the control system for the feed pump occurs. As a result of the failure flow first drops to 1000 gpm for approximately 1 minute then the rises and stabilizes at 2500 gpm.
How did the RFP-1A minimum flow valve respond to this transient?
- a. Opened as flow decreased below 2000 gpm and it remains open.
- b. Opened as flow decreased below 2000 gpm and re-closed as flow rose above 2000 gpm.
- c. Opened as flow decreased below 1250 gpm and it remains open.
- d. Opened as flow decreased below 1250 gpm and re-closed as flow rose above 1250 gpm.
ANSWER: 63 21414
- a. Opened as flow decreased below 2000 gpm and it remains open.
Each Reactor Feed pump (RFP) is equipped with an air operated minimum flow valve which will automatically open if pump flow decreases to approximately 2000 gpm. These valves have no automatic closing feature and must be closed, using the minimum flow c/s on Panel A, when pump discharge flow is greater than 2000 gpm.
Source New 134
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 64 21415 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Systems COR0011602, Implications of adding additional charcoal filters in series.
Related Lessons COR0011602 Off Gas Related Objectives COR0011602001080E Describe the Off Gas system design feature(s) and/or interlock(s) that provide for the following: Maximizing charcoal bed efficiency COR0011602001090D Explain the following Off Gas system related concepts: Charcoal absorption of fission product gases COR0011602001010O State the purpose of the following items related to Off Gas system:
Charcoal Absorber Beds Related References (B)(5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for Related Skills (K/A) 271000.K5.08 Knowledge of the operational implications of the following concepts as they apply to OFFGAS SYSTEM: (CFR: 41.7 / 45.4) Charcoal absorption of fission product gases (2.5/2.6) 135
QUESTION: 64 21415 (1 point(s))
The plant is operating at 100% power with Augmented Off-gas charcoal beds A, B and C in service and D, E and F isolated to install special test equipment.
Following the maintenance charcoal beds D, E and F are place in service in parallel with Charcoal beds A, B and C.
What makes the largest contribution to the decrease in radiation levels at the ERP following this lineup change?
The increased hold up time
- a. of tritium allows for more decay before reaching the ERP.
- b. of radioactive noble gases allows for more decay before reaching the ERP.
- d. of activated corrosion products allows for more decay before reaching the ERP.
ANSWER: 64 21415
- b. of radioactive noble gases allows for more decay before reaching the ERP.
Explanation:
Placing the charcoal beds in service in parallel effectively reduces to half the flow rates through the charcoal filters. This would increase the hold up times for noble gases and reduce the ERP release rate.
Distractors:
- a. is incorrect because tritium is not a measurable component of the ERP effluent.
- d. is incorrect because activation products are primarily particulate and are already effectively filtered by the beds already in service.
- c. is incorrect because essentially all the N-16 has decayed before it reaches the charcoal filters. And therefore adding filters would not effect the release rates.
Source: New 136
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 65 21416 00 06/28/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 1 1 3 Multiple Choice Topic Area Description Systems COR0010502 , Knowledge of the power supplies to Fire Detection.
Related Lessons COR0010501 FIRE PROTECTION SYSTEM COR0010502 FIRE PROTECTION SYSTEM Related Objectives COR0010502001050J Describe the interrelationships between the Fire Protection system and the following: A.C. power Related References (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Related Skills (K/A) 286000.K2.03 Knowledge of electrical power supplies to the following: (CFR: 41.7) Fire detection system: Plant-Specific (2.5*/2.7*)
137
QUESTION: 65 21416 (1 point(s))
What provides power to Diesel Generator room thermal and smoke detectors?
- a. 120VAC from NBPP
- b. 120VAC from CPP-1
- c. 125VDC from DG1 and DG2 d 250VDC Starter Rack (Turbine Building)
ANSWER: 65 21416
- c. 125VDC from DG1 and DG2 Explanation:
DG Cardox System is powered from DG1 and DG2.
Distractors:
- a. is incorrect because NBPP does not supply the Cardox detectors.
- b. is incorrect because CPP-1 does not supply the Cardox system or detectors.
- d. is incorrect because 250 VDC does not supply Cardox. the Cardox system does not supply Source: New 138
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 66 21417 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Administrative INT0320105, Explain and apply system limits and precautions.
Related Lessons INT0320105 SYSTEM OPERATING PROCEDURES Related Objectives INT0320105000030B Given a specific system operating procedure, state the administrative limits concerned with the following items: Temperature, Pressure, Power, Flow, Level INT0320105000040B Given a specific procedure, state the associated precautions concerned with the following items: Temperature, Pressure, Power, Flow, Level INT0320105000040C Given a specific procedure, state the associated precautions concerned with the following items: Valve operations INT03201050000500 Given a specific procedure and situation, discuss any associated cautions or notes stated in the procedure Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.1.32 Ability to explain and apply system limits and precautions. (CFR: 41.10 /
43.2 / 45.12) (3.4/3.8)
- NRC EXAM ONLY**
139
QUESTION: 66 21417 (1 point(s))
The plant is shutdown with the following conditions:
% B RHR loop is in shutdown cooling.
% RHR loop flow is 7100 gpm.
% Both RR pumps are shutdown.
% Reactor Coolant temperature is 125°F.
% The fuel from one entire core quadrant has been removed to perform a shroud inspection.
What action is required and why?
- b. Raise RHR system flow to prevent excessive pump vibration.
- c. Raise RHR system flow to ensure proper reactor coolant mixing.
- d. Reduce RHR system flow to prevent excessive vibration of in-core instruments.
ANSWER: 66 21417
- d. Reduce RHR system flow to prevent excessive vibration of in-core instruments.
Explanation:
During extended periods when fuel or blade guides are removed from around dry tubes, flow should not exceed 7000 gpm recirculating or shut down cooling system drive flow. This measure is required to limit excessive flow induced vibration of in-core instrumentation.
Distractors:
- a. is incorrect because flow does need to be reduced but flow is below the 7700 gpm where runout is a concern.
- b. is incorrect because flow should be reduced this would be a problem with much lower flows. (NRC IN 89-08)
- c. is incorrect because flow should be reduced.
Source: Modified 19901 140
Question Question Revision Revision Last Used Exam Bank Applicability Number ID Number Date Date 67 19324 00 12/12/2005 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 4 Multiple Choice N Topic Area Description Systems COR0022202,Reactor Recirculation; Local Scoop Tube Operation Related Lessons COR0022201 REACTOR RECIRCULATION COR0022202 REACTOR RECIRCULATION Related Objectives COR00222010010800 Demonstrate the ability to locate, in the plant, all local indications associated with the Recirculation System and state the significance of each. Component locations and the location of local indications/alarms may not be stated in this text. The ability of the individual to trace system flowpaths and state locations is implied.
Specific instances may be covered in the lecture, plant tours and/or OJT.
COR0022202001080A Given a Reactor Recirculation system control manipulation, predict and explain the changes in the following parameters: Flow: Core, Jet Pump Related References PR 2.2.68.1 Related Skills (K/A) ROI SROI 2.1.30. Ability to locate and operate components / including local 3.9 3.4 controls. (CFR: 41.7 / 45.7) (3.9/3.4 )
141
QUESTION: 67 19178 (1 point(s))
The plant is operating at 75% power when a failure requires local scoop tube operation of Reactor Recirculation Pump "A".
Which of the following describes the procedural requirements to lock out the scoop tube to obtain local control?
- a. Scoop Tube System 1 Breaker is opened locally.
- b. Scoop Tube System Test Switch is placed in TEST locally.
- d. Scoop Tube external limit switch LS-6 or LS-7 are momentarily operated.
ANSWER:
- d. Scoop Tube external limit switch LS-6 or LS-7 are momentarily operated.
Explanation:
Local control of the scoop tube is obtained by activating one of the external limit switches.
- a. is incorrect. This would cause a scoop tube lock due to loss of power to the scoop positioner but is not the method used to establish local control.
- b. is incorrect. This interrupts power to the motor run circuitry for testing but is not the method for establishing local control.
- c. is incorrect. This is how a scoop tube is locked from the control room, but is not the method used for establishing local controls.
REFERENCES:
COR0022202 Section IV, B.3; 2.2.68.1 Step 17 Source: Direct 142
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 68 14431 01 06/02/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Administrative COR0010301, Ability to use the plant communications equipment.
Related Lessons COR0010301 Communications Related Objectives COR0010301001020A Identify the function of each of the following major components in the communications system: Plant Telephone system Related References 10CFR55.41(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.1.16 Ability to operate plant phone / paging system / and two-way radio. (CFR:
41.10 / 45.12) (2.9/2.8) 143
QUESTION: 68 14431 (1 point(s))
There has been a complete failure of the PBX telephone system at CNS.
Which of the following describes the operation of the control room bypass telephone?
- a. Incoming and outgoing calls can be made on-site only.
- b. Incoming and outgoing calls can be made to or from other bypass telephones or off-site numbers.
- c. Outgoing calls can be made to other bypass telephones. Incoming call capability is disabled.
- d. In-coming calls can be taken from other bypass telephones or from off-site. Outgoing calls can only be made to other bypass telephones.
ANSWER: 68 14431
- b. Incoming and outgoing calls can be made to or from other bypass telephones or off-site numbers.
Explanation:
Bypass phones can call each other on-site and can dial numbers off-site as well as receive calls from off-site. In the event of a total system failure of the PBX, power fail transfer relays will provide service automatically to the exchange network for seven designated stations located in key areas (i.e. Access Control, Control Room, CAS, SAS, Plant Managers office, Senior Managers suite, Switchboard). These telephones are marked as a bypass telephone. In order to operate these telephones (during a system failure), merely pick up the handset which directly connects you to the exchange network, omit the 9 access code, and then dial the desired local or long distance number. Incoming calls will also be received on these telephones.
Source: Direct 144
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 69 20692 00 05/01/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description Administrative SKL0080302, Authorization to Manipulate Valves and Hang Tags Related Lessons SKL0080305 Tagout General SKL0080302 OPS Configuration Management - Ops Related Objectives SKL00803020010100 Given a component or system control device, determine who is authorized to manipulate it using Administrative Procedure 0.31 as a guide.
SKL00803020010300 Given a Tagging Order situation, identify any precautions and limitations associated with it.
Related References 0.31 Equipment Status Control 0.9 Tagout (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.2.13 Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)
(3.6/3.8)
- NRC EXAM ONLY**
145
QUESTION: 69 20692 (1 point(s))
IAC personnel are replacing Level Transmitter NBI-LT-91B this shift. The following valves on Instrument Rack 25-52 are required to be repositioned and tagged to support the instrument replacement:
% NBI-515, LT-91B VALVE MANIFOLD HIGH SIDE SHUTOFF
% NBI-516, LT-91B VALVE MANIFOLD LOW SIDE SHUTOFF
% NBI-517, LT-91B VALVE MANIFOLD EQUALIZER What combination of personnel may be used to perform the valve manipulations and valve tagging?
- a. Any qualified Station Operator may reposition the valves and any qualified Station Operator may tag the valves.
- b. An IAC worker qualified to manipulate valves may reposition the valves and any IAC worker may tag the valves.
- c. An IAC personnel qualified to manipulate valves may reposition the valves and any qualified Station Operator may tag the valves.
- d. Any qualified Station Operator may operate the valves if IAC workers identify the valves and specify the order of operation of the valves any qualified Station Operator may tag the valves.
ANSWER: 69 20692
- c. An IAC personnel qualified to manipulate valves may reposition the valves and any qualified Station Operator may tag the valves.
Procedure 0.9 Tagout, step 2.10 requires that "Tagouts affecting instrumentation on Instrument Rack 25-5, 25-5-1, 25-6, 25-6-1, 25-51, or 25-52 require IAC assistance in identification, manipulation, and specifying order in which tags shall be hung and released. IAC personnel shall perform all valve manipulations on these instrument racks." All three of these valves are on instrument rack 25-52.
Personnel designated to hang and pick up Tagouts shall be a Qualified Operator or shall be certified to Non-OPS Tagging Order Performer. Operations, IAC, and Radiological Department personnel may be required to manipulate valves in accordance with Procedure 0.31 during performance of Tagouts. The qualified Station Operator May hang the tags.
Distractors:
146
- a. is incorrect because IAC personnel are required to manipulate the valves.
- b. is incorrect because this distractor specifies ANY IAC person may hang the tags. The IAC person would have to be qualified to Non-OPS Tagging Order Performer in order to hang the tags.
- c. is incorrect because IAC personnel are required to manipulate the valves.
Source: Direct 147
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 70 16466 01 03/19/2003 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Procedures INT0320104, CNS ADMINISTRATIVE PROCEDURES GENERAL OPERATING PROCEDURES Related Lessons SKL0080301 REACTIVITY CONTROL Related Objectives INT032010400A0100 Discuss Precautions and Limitations outlined in General Operating Procedure 2.1.1, Startup Procedure.
SKL00803010000200 Briefly discuss the key points of the principles to be followed while controlling reactivity.
SKL0080301000030E Discuss precautions and requirements during control rod movements associated with: Reactor period SKL00803010000700 Given plant conditions determine the required action for an emergency power reduction (attachment 7 of procedure 10.13).
Related References 2.1.1 Startup Procedure (B)(1) Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.
(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.2.1 Ability to perform pre-startup procedures for the facility / including operating those controls associated with plant equipment that could affect reactivity.
(CFR: 45.1) (3.7/3.6) 148
QUESTION: 70 16466 (1 point(s))
During a xenon free reactor startup and heatup, reactor period was infinity after withdrawing a control rod. The following conditions are present with NO control rod movement for the last two (2) minutes:
% The reactor is on range 5 of the IRMs (rising)
% Reactor period is +120 seconds (shortening)
% Reactor coolant temperature is 180°F (rising)
What action is required?
- a. Insert control rods in reverse order to make the reactor subcritical.
- b. Insert control rods only as necessary to maintain period longer than 50 seconds.
- c. Bypass the RWM and insert emergency power reduction rods (10.13 Att. 7) to position 00.
ANSWER: 70 16466
- a. Insert control rods in reverse order to make the reactor subcritical.
Step 2.20 of procedure 2.1.1 states that conservative action is required whenever an unexpected situation arises with respect to reactivity, criticality, power level, or any other anomalous behavior of reactor core. This conservative action should include rod insertion to reduce power or a reactor scram without hesitation whenever such unanticipated or anomalous behavior is encountered. In this case indicated power is below the POAH yet temperature is rising and even with this negative reactivity feedback power is rising and period is getting shorter. All of which are significant indications of a significant anomaly.
Answer source: 2.1.1 p. 4, step 2.20 Distractors:
- a. While the administrative limit for period is 50 seconds, the reactor is currently exhibiting anomalous behavior.
- b. This action is not conservative; this action would allow the anomalous reactor behavior to continue.
- d. At this point in the startup operation would be below the 80% rod line and the emergency power reduction control rods are not available.
Source: Direct 149
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 71 19169 01 03/20/2003 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080617, Why do you restart TB Vent while in 5A?
Procedures Related Lessons INT0080617 OPS FLOWCHART 5A - SECONDARY CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL Related Objectives INT00806170010700 Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.
Related References (B)(12) Radiological safety principles and procedures.
Related Skills (K/A) 2.3.11 Ability to control radiation releases. (CFR: 45.9 / 45.10) (2.7/3.2) 150
QUESTION: 71 19169 (1 point(s))
What is the basis for restarting building ventilation in the Turbine Building when executing EOP-5A, RADIOACTIVITY RELEASE CONTROL?
Operation of Turbine Building ventilation
- a. maintains equipment availability AND assures that radioactivity releases pass through a monitored release point.
- b. preserves personnel accessibility AND assures that radioactivity releases pass through a monitored release point.
- c. maintains equipment availability AND assures a minimum amount of radioactivity plates out on turbine building surfaces.
- d. preserves personnel accessibility AND assures a minimum amount of radioactivity plates out on turbine building surfaces.
ANSWER: 71 19169
- b. preserves personnel accessibility AND assures that radioactivity releases pass through a monitored release point.
Explanation: Continued personnel access to the turbine building, radwaste and augmented radwaste may be essential for responding to emergencies. These structures are not air tight and radioactivity release inside them would not only limit personnel access, but would eventually lead to an unmonitored ground level release.
Operation of ventilation in these structures preserves accessibility, and assures that radioactivity is discharged through an elevated, monitored release point.
Answer source: INT008-06-17 p. 13, section B.1 Distractors:
- a. The purpose of restarting Turbine Building ventilation is not to preserve equipment availability.
- c. The purpose of restarting Turbine Building ventilation is not to preserve equipment availability nor to minimize deposition of radioactivity in the building.
- d. The purpose of restarting Turbine Building ventilation is not to minimize deposition of radioactivity in the building.
Source: Direct 151
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 72 21418 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: Y Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Administrative INT0320115, Knowledge of radiation exposure limits Related Lessons INT0320115 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures (Formal Classroom/Pre-OJT Training)
Related Objectives INT0320115D0D010I Discuss the following as described in Rad Protection Procedure 9.ALARA.1, Personnel Dosimetry and Occupational Radiation Exposure Program: Lifetime TEDE Guideline Related References (B)(12) Radiological safety principles and procedures.
Related Skills (K/A) 2.3.4 Knowledge of radiation exposure limits and contamination control / including permissible levels in excess of those authorized. (CFR: 43.4 / 45.10)
(2.5/3.1) 152
QUESTION: 72 21418 (1 point(s))
If you have exceeded your Lifetime TEDE Guideline how much exposure are you allowed during the year at CNS? What authority, if any, may grant extension to this allowed exposure?
- a. 1000 mrem Radiological Manager and Site Vice President may authorize an extension.
- b. 0 mrem No extensions are allowed.
- c. 1000 mrem No Extensions are allowed.
- d. 0 mrem Radiological Manager and Site Vice President may authorize an extension.
ANSWER: 72 21418
- c. 1000 mrem No Extensions are allowed.
Explanation: The Lifetime TEDE Guideline states that NPPD shall normally limit an individual's lifetime TEDE in rem to the individual's age in years. In addition an individual exceeding the lifetime TEDE Guideline will be limited to a TEDE of 1000 mrem and will not be granted an extension.
Distractors:
- a. is incorrect even though 1000 mrem are allowed no extension to this dose is allowed.
- b. is incorrect because 1000 mrem TEDE is allowed.
- d. is incorrect because 1000 mrem TEDE is allowed and now extensions are allowed.
Source: Direct 153
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 73 20498 01 03/24/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 2 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080605, Dispatching personnel during an emergency Procedures Related Lessons INT0080605 OPS FLOWCHART 1A - RPV CONTROL/RPV PRESSURE INT0080501 EOP WALKTHROUGHS Related Objectives INT00806050010900 Identify any EOP support procedures addressed in Flowchart 1A and apply any associated special operating instructions or cautions.
INT00805010010100 In the control room/plant/Simulator, demonstrate ability to use the control room systems, instrumentation, controls, and SPDS terminals for operator action statements from the EOP's or EOP Support Procedures.
Related References 5.8.4 Alternate Injection Subsystems (Table 4)
(B)(12) Radiological safety principles and procedures.
Related Skills (K/A) 2.4.39 Knowledge of the RO's responsibilities in emergency plan implementation.
(CFR: 45.11) (3.3/3.1) 154
QUESTION: 73 20498 (1 point(s))
During an emergency, the following conditions exist:
% TSC not yet operational.
% DW Radiation Monitors are reading 4050 Rem/hr.
% Area Radiation Monitor RM-RA-8 (Reactor Building South CRD) is reading 80 mRem/hr (on scale).
% Area Radiation Monitor RM-RA-9 (Reactor building 903 NE CRD equipment area) is reading 70 mRem/hr (on scale).
% A radiation survey shows the CRD Drive Water Filter area is 100 mRem/hour.
% CRD must be lined up as an alternate injection subsystem to maintain adequate core cooling.
In addition to standard RP practices, what additional requirements, if any, must be met to dispatch an operator to perform this task?
The operator can perform this task...
- a. with no additional requirements.
- b. only if extremity dosimetry is worn.
- c. only after the TSC is declared operational.
- d. if a survey instrument that monitors radiation dose rates is taken.
ANSWER: 73 20498
- d. if a survey instrument that monitors radiation dose rates is taken.
Explanation:
If Station Area Radiation Monitors in travel path and work location of dispatched personnel are alarming, but on-scale, dispatched personnel shall carry a survey instrument capable of monitoring radiation dose rates in travel path and work areas. Dispatched personnel accompanied by a Radiological Protection Technician or Chemistry/Radiological Protection On-Site Availability Technician also satisfies this criteria.
Distractors:
- b. is incorrect because extremity monitoring is not operator may be dispatched without an RP technician if a survey instrument is used.
- c. is incorrect because the TSC need not be operational to dispatch the operator.
155
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 74 21419 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Abnormal/Emergency INT0320107, Knowledge of RO tasks performed outside the Procedures Control Room during emergencies including implications of performing those tasks.
Related Lessons INT0320107 ABNORMAL CONDITION PROCEDURES Related Objectives INT03201070000500 Given a specific procedure, explain or analyze any NOTES and CAUTIONS addressed in the procedure INT03201070000300 Given a specific procedure title, or adequate information of plant conditions and indications, analyze the immediate actions required INT03201070000400 Given a specific procedure title, appraise the key concepts from the discussion section Related References (B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system implications.
(CFR: 43.5 / 45.13) (3.8/3.6) 156
QUESTION: 74 21419 (1 point(s))
The plant was operating when the following occurred:
% Control Room is evacuated due to toxic gas intrusion.
% The reactor is not scrammed before leaving the Control Room.
How is the reactor scrammed?
What group isolations result when the reactor is scrammed from outside the Control Room?
(Assume no automatic isolations occurred prior to the scram).
- a. The breaker for RPS Alternate power and the RPS MG set AC input breaker are opened in RPS MG set rooms. Only group isolations 2, 3 and 6 result.
- b. The breaker for RPS Alternate power and the RPS MG set AC input breaker are opened in RPS MG set rooms. Group isolations 1,2,3,6 and 7 result.
- c. The breakers for all APRMs in the cable spreading room are tripped. Only group isolations 2, 3 and 6 result.
- d. The breakers for all APRMs in the cable spreading room are tripped. Group isolations 1,2,3,6 and 7 result.
ANSWER: 74 21419
- b. The breaker for RPS Alternate power and the RPS MG set AC input breaker are opened in RPS MG set rooms. Group isolations 1,2,3,6 and 7 result.
Explanation:
5.1ASD has the Control Building operator open the alternate power supply breakers and the MG set supply breakers in order to scram the reactor. This completely deenergizes RPS and results in group isolations 1,2,3,6 and 7..
Distractors:
- a. is incorrect because a group 1 and 7 also occur. The candidate that just believes that the only isolations that occur are due to shrink may pick this answer.
- c. is incorrect because the reactor is scrammed by deengizing RPSPP1A and 1B.
Deenergizing the APRMs would cause a scram but not as directed by procedure.
- d. is incorrect because the reactor is scrammed by deengizing RPSPP1A and 1B.
Deenergizing the APRMs would cause a scram but not as directed by procedure..
Source: Modified 4231 157
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 75 1744 02 07/19/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 2 1 4 Multiple Choice Topic Area Description Systems COR0022302, Residual Heat Removal System Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001030P Describe RHR System design feature(s) and/or interlocks which provide for the following: Spray flow cooling COR0022302001170C Given plant conditions, determine actions necessary to place RHR in the following flowpaths: Drywell Spray Related References 791E261 RHR Elementary Diagram 2.2.69 Residual Heat Removal System 2.2.69.3 RHR Suppression Pool Cooling And Containment Spray (B)(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(B)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Related Skills (K/A) 2.4.48 Ability to interpret control room indications to verify the status and operation of system / and understand how operator actions and directives affect plant and system conditions. (CFR: 43.5 / 45.12) (3.5/3.8) 158
QUESTION: 75 1744 (1 point(s))
Following a LOCA, the following conditions are present:
% Reactor pressure is 700 psig (lowering slowly)
% RPV water level is +50" (Corrected FZ) and stable.
% Torus pressure is 11.0 psig (rising slowly)
% Drywell pressure is 12.0 psig (rising slowly)
What are the actions that are required by the Drywell Spray initiation logic, in order to initiate RHR "A" Drywell Sprays?
- a. Place Drywell Inbd (MO-31A) and Outbd (MO-26A) Spray Valve control switches in OPEN only.
- b. Place Containment Cooling Valve Control Permissive switch in MANUAL, then place Drywell Inbd (MO-31A) and Outbd (MO-26A) Spray Valve control switches in OPEN only.
- c. Place Containment Cooling 2/3 Core Valve Control Permissive switch in OVERRIDE, place the Containment Cooling Valve Control Permissive switch in MANUAL, then place Drywell Inbd (MO-31A) and Outbd (MO-26A) Spray Valve control switches in OPEN only.
- d. Depress Containment Spray Initiation Signal Reset pushbutton, place Containment Cooling 2/3 Core Valve Control Permissive switch in OVERRIDE, place the Containment Cooling Valve Control Permissive switch in MANUAL, then place Drywell Inbd (MO-31A) and Outbd (MO-26A) Spray Valve control switches in OPEN.
ANSWER: 75 1744
- b. Place Containment Cooling Valve Control Permissive switch in MANUAL, then place Drywell Inbd (MO-31A) and Outbd (MO-26A) Spray Valve control switches in OPEN only.
Distractors:
- a. The permissive switch must be placed in MANUAL.
- c. No need to place 2/3 core height in override.
- d. No need to place 2/3 core height in override or reset logic.
Source: Direct 159
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 76 21420 00 02/28/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Abnormal/Emergency INT0320124, Interpret Plant indications including core flow to Procedures determine the plant is in Natural Circulation.
Related Lessons INT0320124 CNS Abnormal Procedure (RO) Reactor Recirculation Related Objectives INT032012400F0F00 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).
Related References 10CFR55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Related Skills (K/A) 295001.AA2.03 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
(CFR: 41.10 / 43.5 / 45.13) Actual core flow (3.3/3.3) 160
QUESTION: 76 21420 (1 point(s))
The plant is operating at 50% power and 40 MLBH Core Flow in single loop operation.
RR Loop 1A is isolated. A transient occurs to the B RR loop and the following indications are noted by the crew:
% Recirc Pump Diff. Pressure, DPI-156B falls to 0 psid.
% Recirc Loop 1B Flow FI-159B falls to 0 gpm.
% Recirc pump speed indicates 77%.
% RRMG set drive motor breaker and field breaker are closed.
% Actual Core flow is determined to be 23 MLBH.
% Reactor power falls to 44%.
What is required?
- a. Enter 2.4RR only
- b. Enter 2.4RXPWR
- c. Enter 5.3AC120 for a loss of CDP-1A
- d. Enter 2.4RR and enter procedure 2.1.5 ANSWER: 76 21420
- d. Enter 2.4RR and enter procedure 2.1.5 Explanation.
The parameters given indicate a sheared shaft for the B RR pump. This plant is now operating at greater than 1% power with both RR pumps essentially tripped. 2.4RR directs that if both pumps are tripped to scram the reactor and enter 2.1.5.
Distractors:
- a. Entry into 2.4RR is required but a scram and entry to 2.1.5 is also required.
- b. is incorrect because entry into 2.4 RXPWR is not required because the cause of the power reduction is known.
- c. is incorrect because a loss of CDP-1A has not occurred.
Source New 161
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 77 19219 02 09/24/2003 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Technical Specifications, INT0070509, Degraded Grid, SSST/ESST operability ODAM, TRM Related Lessons INT0070509 OPS Tech. Spec. 3.8, Electrical Power Systems Related Objectives INT00705090010100 Given a set of plant conditions, recognize non-compliance with a Section 3.8 LCO.
Related References 5.3GRID Degraded Grid Voltage 3.8.1 AC Sources - Operating 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
Related Skills (K/A) 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications. (CFR: 43.2 / 43.3 / 45.3)
(3.4/4.0) **EXAM USE ONLY**
162
QUESTION: 77 19219 (1 point(s))
The plant is operating at rated power with the following conditions:
% A degraded grid condition exists.
% Voltage on CNS 161 kV bus is 168 KV.
% Voltage on Cooper Cornfield 69 kV bus is 71 KV.
% Security Analysis is not in service.
% Doniphan notifies CNS Control Room of "MVAR on Cooper Generator" alarm.
% The main generator is at + 350 MVAR What is the status of off-site source operability?
- a. BOTH off-site sources are OPERABLE.
- b. BOTH off-site sources are INOPERABLE.
- c. The Emergency Transformer is INOPERABLE; the Startup Transformer is OPERABLE.
- d. The Emergency Transformer is OPERABLE; the Startup Transformer is INOPERABLE.
ANSWER: 77 19219
- b. BOTH off-site sources are INOPERABLE.
The 345 KV system is connected to the 161 KV system at CNS and is connected to the 69 KV at Brock. Plant experience has shown that CNS VAR changes directly and immediately affect 69 and 161 KV voltages. 5.3GRID Attachment 1 .a states "If MVAR meter OUT $ 150, and the Security Analysis is not in service or not solving,:
- Declare SSST inoperable and enter appropriate Condition and Required Action of LCO 3.8.1, AC Sources - Operating.
- Declare ESST inoperable and enter appropriate Condition and Required Action of LCO 3.8.1, AC Sources - Operating" Distractors:
a, c, d BOTH off-site sources become INOPERABLE.
Source: Direct 163
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 78 21422 00 02/28/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Abnormal/Emergency INT0320127, Interpret temperatures and select appropriate Procedures procedure.
Related Lessons INT0320127 CNS Abnormal Procedures (RO) Turbine/Generator Related Objectives NONE Related References 10CFR55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Related Skills (K/A) 295018.AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR:
41.10 / 43.5 / 45.13) Component temperatures (3.3/3.4) 164
QUESTION: 78 21422 (1 point(s))
The plant is operating at near rated power when the RO reports rising generator hydrogen temperatures. The RO also reports that bearing and oil temperatures are normal and TEC temperature and pressure are normal.
What is required?
- a. Enter 2.4TURB only.
- b. Enter 2.4GENH2 only.
- c. Enter 2.4TEC and 2.4GENH2.
- d. Enter 2.4TEC and 2.4TURB.
ANSWER: 78 21422
- b. Enter 2.4GENH2 only.
Explanation:
The entry conditions for 2.4GENH2 are:
Abnormal generator H2 gas temperatures or pressures.
Abnormal generator stator temperatures or vibrations.
Abnormal H2 seal oil pressures.
Abnormal exciter amps or volts.
Since generator hydrogen temperatures are rising and approaching their alarm setpoint an entry conditions exists.
Since only one component cooled by TEC is affected there is no entry condition for 2.4TEC.
No entry condition exists for 2.4TURB.
Distractors:
- a. is incorrect because no entry condition exists for 2.4TURB.
- c. is incorrect because only 2.4GENH2 should be entered.
- d. is incorrect because neither of these procedures should be entered.
Source: New 165
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 79 4171 5 08/09/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Administrative COR0099900 QUESTION #502 Related Lessons INT0320203 CONDUCT OF OPERATIONS PROCEDURES (SRO)
INT0320103 CNS Administrative Procedures Conduct of Operations and General Alarm Procedures (Formal Classroom/Pre-OJT Training)
INT0320201 CNS PROCEDURES (SRO)
Related Objectives INT03202030000900 Using procedure 2.0.5, identify conditions requiring one or four hour NRC notification. (2.0.5)
Related References 2.0.5 Reports to NRC Operations Center Related Skills (K/A) 2.4.30 Knowledge of which events related to system operations/status should be reported to outside agencies. (CFR 43.5 / 45.11) (2.2/3.6) 166
QUESTION: 79 4171 (1 point(s))
The plant is operating at 70% power and instructions are to raise to 100% power. A generator fault results in a turbine trip and reactor scram. Groups 2, 3, and 6 low reactor water level isolations are received before level is stabilized. All rods are full in. Reactor level is 35" controlled by RFPs and reactor pressure is 900 psig controlled by turbine bypass valves.
What is the most limiting NRC notification requirement?
- a. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
- b. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- c. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
- d. License Event Report ANSWER: 79 4171
- b. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> EXPLANATION:
RPS actuation and ESF actuations are 4 hr. Reportable events per 10CFR50.72(b)(2)(iv)(B).
Distractors:
- a. is incorrect because a one hour report is not required.
- b. is incorrect because an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report is not required.
- d. is incorrect because this is not the limiting report.
Source: Direct 167
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 80 19939 00 03/07/2003 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 3 1 1 Multiple Choice Topic Area Description Technical Specifications, INT00705070010200, CNS TECH SPEC 3.6, ODAM, TRM CONTAINMENT SYSTEMS Related Lessons INT0070507 CNS Tech. Spec. 3.6, Containment Systems Related Objectives INT00705070010200 Discuss the applicable Safety Analysis in the Bases associated with each Chapter 3.6 specification.
Related References 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
3.6.2.1 Suppression pool average temperature Related Skills (K/A) 295013.AK1.04 Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL TEMPERATURE: (CFR: 41.8 to 41.10) Complete condensation. (2.9/3.2) 168
QUESTION: 80 19939 (1 point(s))
The plant is operating at 100% power with suppression pool temperature at 98°F 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> after HPCI was secured from testing.
According to Technical Specification Bases what potential consequence is there to operation in this condition?
- b. The capacity of the torus-to-drywell vacuum breakers will be exceeded during a LOCA.
- c. The reactor building-to-torus vacuum breakers will operate if drywell sprays are initiated during a LOCA.
- d. Peak primary containment pressure and temperatures do not exceed maximum allowable values during a design basis accident (DBA).
ANSWER: 80 19939
- d. Peak primary containment pressure and temperatures do not exceed maximum allowable values during a design basis accident (DBA).
Answer source: Tech Spec bases p. B 3.6.2.1 Distractors:
- b. This would be an issue for initiating drywell spray with evaporative cooling and is positively affected by high torus temp, not negatively affected.
- c. This would be an issue for initiating drywell spray with evaporative cooling and is positively affected by high torus temp, not negatively affected.
55.43 section(s): (2)
Source: Direct 169
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 81 21518 00 08/17/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 1 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080610, ATWS What actions are required?
Procedures Related Lessons INT0080610 OPS EOP FLOWCHART 7A - RPV LEVEL (FAILURE-TO-SCRAM)
Related Objectives INT00806100010700 Identify any EOP support procedure addressed in Flowchart 7A and apply any associated special operating instructions or cautions.
INT00806100010800 Given plant conditions and EOP flowchart 7A, RPV LEVEL (FAILURE TO SCRAM), determine required actions.
Related References EOP FLOWCHART 7A - RPV LEVEL (FAILURE-TO-SCRAM)
Related Skills (K/A) 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) (4.0/4.3)
- LINK ONLY TO EOP/AOP LESSONS/QUESTIONS**
170
QUESTION: 81 21518 (1 point(s))
A group 1 isolation and an ATWS occurred. The crew intentionally lowered level when level power conditions were met and established a level band of -25" to +50" (FZ). An emergency depressurization is required when HCTL is approached. The crew commenced injection when minimum steam cooling pressure was reached.
% All available outside shroud injection systems are injecting at full flow.
% Reactor water level is -30" (Corrected FZ) and lowering.
% Reactor pressure is 55 psig and steady.
% Torus pressure is 10 psig and steady.
% 20% of the SLC tank has been injected
% ESP 5.8.3 Alternate Rod Insertion Methods is in progress.
What is required?
- c. Enter ESP 5.8.2 Alternate Emergency Depressurization Systems.
- d. Continue attempts to raise level per ESP 5.8.13 Outside the Shroud Injection Systems.
ANSWER: 81 21518
Explanation:
After the ED the crew is at step FS/L-17 on EOP-7A. All outside the shroud systems are at maximum and the reactor is depressurized and level continues to lower. Outside the shroud systems are now required. The crew should inject with CS and enter 5.8.15.
Distractors:
- a. is incorrect because level is determinate.
- c. is incorrect because the reactor is depressurized (within 50 psi of the torus).
- d. Is incorrect because with the decreasing level and outside the shroud systems already injecting at maximum additional flow is required.
Source: Direct 171
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 82 21426 00 06/24/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description Emergency Plan GEN0030121, Ability to interpret radiation levels and classify the event.
Related Lessons GEN0030121 EMERGENCY ACTION LEVELS Related Objectives GEN00301210000100 Given a scenario describing plant conditions requiring declaration of an Emergency, the student will identify the correct Emergency Class using the Emergency Planning Implementing Procedures.
Related References 10CFR55.43 (5) Assessment of facility conditions and selestion of appropriate procedures during normal, abnormal, and emergency sitituations.
Related Skills (K/A) 295038.EA2.03 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.10 / 43.5 / 45.13) ?Radiation levels (3.5*/4.3*)
172
QUESTION: 82 21426 (1 point(s))
The plant is operating at 100% power when the following timeline of events occur:
10:00 Large LOCA occurs and on the subsequent scram very few rods insert into the core.
10:05 Attempts to start SLC pumps are unsuccessful.
10:10 Reactor Water Level Drops below TAF and cannot be raised.
10:15 Drywell Radiation Monitors rise above 104 Rem/hr.
10:20 Drywell pressure is 26 psig and slowly rising.
10:25 Projected Dose at the Site Boundary is 50 mrem TEDE and 100 mrem CDE (thyroid).
When is the declaration of a General Emergency FIRST required?
- a. 10:10
- b. 10:15
- c. 10:20
- d. 10:25 ANSWER: 82 21426
- c. 10:20 Provide the candidate with Procedure 5.7.1.
Explanation:
For this scenario a general emergency is first declared when 2 fission product barriers are lost and the potential exists for the loss of the third. The first barrier is lost with the LOCA and the second is lost (fuel clad) when the DW rad monitors read greater than 2500 rem/hour. At 1020 DW pressure rises to greater than 25 psig which is the potential loss of the third barrier and therefore constitutes a GE.
Source: New 173
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 83 21427 00 06/24/2006 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Abnormal/Emergency INT0320107, Determine the cause of the reactivity addition.
Procedures Related Lessons INT0320107 ABNORMAL CONDITION PROCEDURES Related Objectives INT03201070000100 Given a list of symptoms, identify the abnormal condition Related References 10CFR55.43 (6) Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.
Related Skills (K/A) 295014.AA2.03 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: (CFR: 41.10 / 43.5 / 45.13)
Cause of reactivity addition (4.0/4.3*)
174
QUESTION: 83 21427 (1 point(s))
A plant startup and power ascension is in progress with the following conditions:
% Reactor power is 15%
% Main Turbine speed is approaching 1800 rpm.
% Core Plate D/P is 1.4 psid.
% Reactor Pressure is 930 psig.
Shortly after the turbine reached 1800 rpm reactor power quickly rose to and stabilized at 20%
with the following conditions:
% Reactor power is 20%
% Main turbine speed is 1800 rpm and steady.
% Core Plate D/P is 1.4 psid.
% Reactor pressure is 931 psig.
% No annunciation occurred before during or after the transient What is the cause of the reactor power increase?
What procedure entry is required?
- a. Dropped control rod 2.4CRD, CRD Trouble
- b. Dropped control rod 2.4RXPWR, Reactor Power Anomalies
- c. Control Rod Drifting Out 2.4RXPWR, Reactor Power Anomalies
- d. Control Rod Drifting Out 2.4CRD, CRD Trouble ANSWER: 83 21427
- b. Dropped control rod 2.4RXPWR, Reactor Power Anomalies The cause of the power rise is a dropped control rod. The drifting control rod is ruled out because there is no annunciation. An unexplained increase in reactor power required enty into 2.4RXPWR.
Distractors:
175
- a. 2.4CRD would not be entered for a dropped control rod.
- c. Indications are that a rod drop accident occurred. For a rod drop accident 2.4RXPWR is entered.
- d. Indications are that a rod drop accident occurred.
Source: New 176
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 84 21428 00 07/09/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Technical Specifications, INT0070508, Knowledge of the bases in TS for CREFS ODAM, TRM actuation.
Related Lessons INT0070508 CNS Tech. Spec. 3.7, Plant Systems Related Objectives INT00705080010200 Discuss the applicable Safety Analysis in the Bases associated with each Chapter 3.7 Specification.
Related References 10CFR55.43 (2) Facility operating limitations in the technial specifications and their bases.
Related Skills (K/A) 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (CFR: 43.2) (2.5/3.7)
- LINK ONLY TO TECH SPEC LESSONS & QUESTIONS**
177
QUESTION: 84 21428 (1 point(s))
Technical Specifications require that the CREF system be operable during handling of irradiated fuel.
What is the bases for this requirement?
The bases for this requirement is that in the event if a fuel handling accident that releases radioactivity to Secondary containment the dose to control room personnel is limited to
- a. 5 Rem whole body for a 200 man-day continuous occupancy.
- b. 5 Rem whole body for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the accident.
- c. 25 Rem whole body for a 200 man-day continuous occupancy.
- d. 25 Rem whole body for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the accident.
ANSWER: 84 21428
- a. 5 Rem whole body for a 200 man-day continuous occupancy.
Explanation:
The CREF System is designed to maintain the control room environment for a 200 man day continuous occupancy after a DBA without exceeding 5 rem whole body dose or its equivalent to any part of the body.
Distractors:
- b. is incorrect because the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reference is incorrect. The candidate that confuses the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> language from 10CFR100 may choose this answer.
- c. is incorrect because 25 rem is greater than the design value.
- d. is incorrect because 25 rem is greater than the design value and the design time is 200 man-days Source: New 178
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 85 21429 00 06/25/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Abnormal/Emergency INT0320130, Recognize indications that require entry into Procedures emergency/abnormal procedures.
Related Lessons INT0320107 ABNORMAL CONDITION PROCEDURES INT0320130 CNS Abnormal Procedures (RO) High Radiation Related Objectives INT03201070000100 Given a list of symptoms, identify the abnormal condition Related References 10CFR55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Related Skills (K/A) 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) (4.0/4.3)
- LINK ONLY TO EOP/AOP LESSONS/QUESTIONS**
179
QUESTION: 85 21429 (1 point(s))
The plant was operating at power when an accident occurred that resulted in the following conditions:
% Radiation release rate has increased to twice the ODAM limit.
% Annunciator R-2/A-5, REACTOR BLDG PUMP ROOM HIGH TEMP is alarming.
% Reactor Building Exhaust Plenum radiation level is 7 mrem/hr.
% Refuel floor radiation levels are 5 mrem/hr.
% Refuel Floor CAM is alarming.
% Reactor Building D/P is 0"H2O.
What procedure(s) is/are required to be entered?
- a. only.
- b. and 5.1 RAD only.
- c. and 2.4HVAC only.
- d. 2.4HVAC and 5.1RAD.
ANSWER: 85 21429
- d. 2.4HVAC and 5.1RAD.
EOP 5A entry is required by the reactor building low D/P. 5.1RAD is required to be entered due to the CAM alarm and 2.4HVAC entry is required by high area temperature alarm.
Source: New 180
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 86 21262 00 10/16/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description Technical Specifications, INT0070506, HPCI and one Loop of LPCI inop ODAM, TRM Related Lessons INT0070506 OPS Tech. Spec. 3.5, Emergency Core Cooling (ECCS) and Reactor Core Isolation Cooling (RCIC) System Related Objectives INT00705060010100 Given a set of plant conditions, recognize non-compliance with a Section 3.5 LCO.
INT00705060010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.5 LCO, determine the ACTIONS that are required.
Related References 3.5.1 ECCS Operating Related Skills (K/A) 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR:
43.2 / 45.2) (3.4/4.1) **EXAM USE ONLY**
181
QUESTION: 86 21262 (1 point(s))
The plant is at 100% power. The following conditions then occur:
% 1200 on 10/16, HPCI is declared inoperable
% 1200 on 10/17, LPCI loop A is declared inoperable.
When is the plant first required to be in mode 3 by Technical Specifications?
- a. 0100 on 10/18.
- b. 0000 on 10/21.
- c. 0000 on 10/24.
- d. 1200 on 10/30.
ANSWER: 86 21262
- b. 0000 on 10/21.
Since HPCI and one LP ECCS system are inoperable condition D of TS3.5.1 is entered. This starts a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock to restore one of the systems to operable. At the end of the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Condition G is entered which now directs 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to be in mode 3. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> plus 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from 1200 on 10/17 is 0000 on 10/21.
- a. is incorrect because at this time the plant is not yet required to be in mode 3. The examinee that mistakenly enters 3.0.3 would choose this answer.
- c. is incorrect because the plant is required to be in mode 3 prior to this time. The examinee that mistakenly enters only 3.5.1.A and then 3.5.1.B would choose this answer.
- d. is incorrect because the plant is required to be in mode 3 prior to this time. The examinee that only entered Action C would choose this answer.
Source: Direct 182
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 87 21172 01 09/11/2004 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Technical Specifications, INT0070502, Predict the immediate effect of a loss of heaters ODAM, TRM on SLC and determine required actions.
Related Lessons INT0070502 CNS Tech. Spec. 3.1, Reactivity Control Systems Related Objectives INT00705020010100 Given a set of plant conditions, recognize non-compliance with a Section 3.1 LCO.
Related References 2.2.74 Standby Liquid Control System 6.LOG.601 Daily Surveillance Log - Modes 1, 2, and 3 3.1.7 Standby liquid control (SLC) system 3.1.7-1 Technical Specification LCO Figure 3.1.7-2 Technical Specification LCO 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
Related Skills (K/A) 211000.A2.05 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences...: (CFR:
41.5/45..5) Loss of SBLC tank heaters (3.1/3.4) 183
QUESTION: 87 21172 (1 point(s))
With the plant operating at 100% power, the following occur:
08:00 9-5-2/F-7 SLC TANK HEATER GROUND TO SOLUTION alarms.
08:20 The heater is deenergized to support troubleshooting.
09:30 9-5-2/G-8 SLC TANK HI/LOW TEMP alarms.
09:31 SO reports SLC Boron Solution Temp at 85EF and Tank Volume at 78%.
09:31 Chemistry sample indicates SLC Boron Solution Concentration at 15.4 weight percent pentaborate.
What impact does this have on SLC operability?
By procedure, what action is appropriate next?
- a. Both SLC subsystems are inoperable.
Enter 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO.
- b. Both SLC subsystems are inoperable.
Enter a 7 day LCO.
Install heaters in the area to maintain SLC operability.
Partially drain and refilled the SLC tank with demin water to maintain SLC operable.
ANSWER: 87 21172
Install heaters in the area to maintain SLC operability.
Provide the candidate with Figures 3.1.7-1 Per 2.2.74 11.1.2, have maintenance install portable heating in the SLC System area to raise ambient temperatures.
Distracter a, b. Both SLC subsystems remain operable. 78% tank level equates to 3560.8 gallons. Per TS Figure 3.1.7-1, concentration at 15.4% with volume at 3560.8 gallons is in the ACCEPTABLE region. Per TS Figure 3.1.7-2, temp at 85EF with concentration at 15.4% is in the ACCEPTABLE region. If the examinee incorrectly calculates tank volume in gallons, they could determine unacceptable SLC parameters are present when they are not.
Distracter d. Although an acceptable action in 2.2.74, section 11. This action would place the SLC solution closer to unacceptable parameters by diluting the concentration and lowering the SLC solution temp with colder demin water.
Source: Direct 184
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 88 21430 00 07/09/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Technical Specifications, INT0070504, Knowledge of TS Bases ODAM, TRM Related Lessons INT0070504 CNS Tech. Spec. 3.3, Instrumentation Related Objectives INT00705040010200 Discuss the applicable Safety Analysis in the Bases associated with each Section 3.3 Specification.
Related References 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
Related Skills (K/A) 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (CFR: 43.2) (2.5/3.7)
- LINK ONLY TO TECH SPEC LESSONS & QUESTIONS**
185
QUESTION: 88 21430 (1 point(s))
What is the Technical Specification basis for having the Intermediate Range Monitor Neutron Flux-High trip?
- a. Provides protection against the control rod drop accident by limiting fuel enthalpy to less than 280 cal/gm fuel damage limit.
- b. Provides protection against local control rod withdrawal errors to limit peak fuel enthalpy below 170 cal/gm fuel failure threshold criterion.
- c. Provides protection against a single sequence error that results a high local flux to ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
- d. Provides protection against violation of the MCPR Safety Limit and the cladding 1%
plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event.
ANSWER: 88 21430
- b. Provides protection against local control rod withdrawal errors to limit peak fuel enthalpy below 170 cal/gm fuel failure threshold criterion.
The IRM provides mitigation of the neutron flux excursion. To demonstrate the capability of the IRM System to mitigate control rod withdrawal events, generic analyses have been performed (Ref. 3) to evaluate the consequences of control rod withdrawal events during startup that are mitigated only by the IRM. The continuous rod withdrawal during reactor startup analysis (Refs.
2 and 3), which assumes that one IRM channel in each trip system is bypassed, demonstrates that the IRMs provide protection against local control rod withdrawal errors and results in peak fuel enthalpy below the 170 cal/gm fuel failure threshold criterion.
55.43 section(s): (2)
SRO Justification: SRO persons assess plant conditions and determine compliance with Technical Specifications knowledge of Technical Specification bases are required to make these assessments.
Source: Direct PTM21309 186
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 89 21431 00 07/09/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Administrative INT0320211, Predict the impact of a stuck detector and determine required actions Related Lessons INT0320211 INSTRUMENTATION OPERATING PROCEDURES (SRO)
Related Objectives INT03202110000100 Given the procedure, discuss any Notes, Limitations, Precautions, and interlocks that pertain to operator actions INT03202110000200 Given the appropriate procedure discuss how and when control rod coupling verification is to be performed INT03202110000300 Identify the signature and logging requirements for temporary setpoint changes. (4.0.1)
INT03202110000400 Describe when a temporary setpoint change may be required and who must be informed. (4.0.1)
Related References 10CFR55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Related Skills (K/A) 215004.A2.03 Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those...:
(CFR: 41.5 / 45.6) Stuck detector (3.0/3.3) 187
QUESTION: 89 21431 (1 point(s))
A plant startup is in progress with the following conditions:
% IRM C and IRM F are failed downscale and are bypassed.
% SRM B is downscale and has not been bypassed.
% The shorting link switches are in OPEN.
% Power level is on IRM range 3 for all operable IRMs.
% The SRMs counts are approaching 105 cps.
As the crew attempts to withdraw the SRMs, SRM A detector fails to move. The power ascension is stopped and maintenance determines that the detector is stuck and cannot be freed.
1.) If the power ascension is continued in this condition, what effect would the stuck detector have on operation?
2). What actions are required?
- a. Rod block and Full Scram.
Enter procedure 4.1.1, bypass SRM channel A and deenergize SRM A by removing drawer fuses after the mode switch is in RUN.
- b. Rod block only.
Immediately discontinue rod withdrawal, enter procedure 2.1.4 and shutdown the reactor.
- c. Rod block and Full Scram.
Immediately discontinue rod withdrawal, enter procedure 2.1.4 and shutdown the reactor.
- d. Rod block only.
Enter procedure 4.1.1, bypass SRM channel A and deenergize SRM A by removing drawer fuses after the mode switch is in RUN.
ANSWER: 89 21431
- a. 1). Rod block and Full Scram.
2). Enter procedure 4.1.1, bypass SRM channel A and deenergize SRM A by removing drawer fuses after the mode switch is in RUN.
Explanation:
188
If power is raised with SRM A detector stuck and unbypassed with the shorting link switches open a rod block and full scram would occur. 4.1.1 provides directions to bypass the stuck detector. Since IRMs are on range 4 Technical specifications allow the continued power ascent.
When the mode switch is in RUN, 4.1.1 requires that the SRM A fuses in the SRM drawer be pulled.
Distractors:
- b. is incorrect because a full scram would also occur and continued rod withdrawal is allowed.
- c. is incorrect because continued rod withdrawal is allowed.
- d. is incorrect because a full scram would occur if power is raised before any other actions are taken.
Source: New 189
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 90 21432 00 07/10/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 2 1 3 Multiple Choice Topic Area Description Abnormal/Emergency INT0320131, Predict the impact of over current on breakers Procedures and determine the procedures to enter to mitigate the consequences.
Related Lessons INT0320131 CNS Abnormal Procedures (RO) Electrical Related Objectives INT0320131T0T0100 Given plant condition(s), determine from memory any automatic actions listed in the applicable Abnormal/Emergency Procedure(s) which will occur due to the event(s).
INT0320131S0S0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).
Related References 10CFR55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Related Skills (K/A) 262001.A2.10 Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences...: (CFR: 41.5 / 45.6)
Exceeding current limitations (2.9/3.4) 190
QUESTION: 90 21432 (1 point(s))
Given the following conditions:
% Reactor is in Hot Shutdown.
% The emergency transformer is deenergized.
% DG 1 is paralleled to 1F.
% The diesel governor fails to maximum fuel rack position (DG current rises to greater than 180% of normal full load current.)
- 1) What breaker 1FA or 1AF trips in response (either direct or indirect) to the high DG1 current?
- 2) What action would be required following the trip?
- a. ONLY 1AF trips.
Enter 5.3EMPWR
- b. ONLY 1FA trips.
Enter 5.3EMPWR
- c. ONLY 1AF trips.
Enter 5.3AC480
- d. ONLY 1FA trips.
Enter 5.3AC480 ANSWER: 90 21432
- d. ONLY 1FA trips.
Enter 5.3AC480 Explanation:
1FA is tripped by the over current condition. 1AF will not trip because 1AF is in NORMAL AFTER CLOSE and Bus 1A is energized. Once 1AF trips the DG will overspeed and trip Deenergizing 4160 1F. Since 4801F is lost entry into 5.3AC480 is required. The entry conditions for 5.3EMPWR are not met because 4160A, B an E remain energized.
REFERENCES:
STCOR001-01-02 PR 2.2.18, page 13, 15, section 4.6.2, 4.9.2.4 PR 2.2.20, page 16, section 4.9 Distractors 191
- a. is incorrect because 1FA trips on over current and 5.3EMPWR is not entered.
- b. is incorrect because 5.3EMPWR is not entered.
- c. is incorrect because 1FA trips on over current.
Source: Modified 19223 192
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 91 21433 00 07/09/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Technical Specifications, INT0070513, Determine conditions that are entry level TS ODAM, TRM Related Lessons INT0070513 CNS Tech. Spec. 5.0, Administrative Controls Related Objectives INT00705130010100 Given a set of plant conditions, recognize non-compliance with a Chapter 5.0 Requirement.
INT00705130010200 Given a set of conditions that constitutes non-compliance with a Chapter 5.0 Requirement, determine the actions that are required.
Related References 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
Related Skills (K/A) 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications. (CFR: 43.2 / 43.3 / 45.3)
(3.4/4.0) **EXAM USE ONLY**
193
QUESTION: 91 21433 (1 point(s))
Two temporary bladder tanks have been installed as a temporary surge volume for radwaste.
The bladders are located inside the protected area but have no retaining walls. Radwaste effluent is currently being discharged to bladder tanks A and B.
Chemistry report that following data following a sample of the tanks:
% Tank A contains 5,000 gallons and 12 Ci of activity (excluding tritium and noble gas).
% Tank B contains 6,000 gallons and 9 Ci of activity (excluding tritium and noble gas).
What action(s) is/are required?
- a. Immediately suspend addition of radwaste to bladder tank A.
- b. Immediately suspend addition of radwaste to both bladder tanks.
- c. Immediately initiate action to reduce the curie content of both tanks.
- d. Shutdown per TS LCO 3.0.3.
ANSWER: 91 21433
- a. Immediately suspend addition of radwaste to bladder tank A.
Provide the Candidate with TS5.5.8 and D3.1.4.
Explanation:
Tanks without retaining walls are limited to 10 Ci each. In this case the tank A is greater than 10 curies. IAW DLCO 3.1.4 Action A Immediately suspend the addition of radioactive material to the tank.
Distractors:
- b. is incorrect because tank B is lower than the 10 Ci limit.
- c. is incorrect because tank B has no requirement to reduce the Ci content of the tank.
- d. is incorrect because no 3.0.3 shutdown is not required. DLCO3.0.3 is the applicable vehicle for non-compliance with ODAM limits.
Source New 194
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 92 16634 00 10/20/1997 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Administrative COR0010502, FIRE PROTECTION TS Required actions Related Lessons COR0010502 FIRE PROTECTION SYSTEM Related Objectives COR00105020010200 Given condition(s) and/or parameters associated with the Fire Protection system, determine if related Technical Requirements Manual Limiting Condition for Operation are met.
Related References 0.23 CNS Fire Protection Plan 6.FP.101 Fire Pump 31 Day Operability Test 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
Related Skills (K/A) 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR:
43.2 / 45.2) (3.4/4.1) **EXAM USE ONLY**
195
QUESTION: 92 16634 (1 point(s))
Following the performance of 6.FP.101, Fire Pump 31 Day Operability Test, it is noted that Electric Fire Pump C would NOT automatically start and would NOT start when the START pushbutton was pressed on local control panel.
What is action is the required action?
- a. Restore the fire pump to OPERABLE status within 7 days.
- b. Initiate a Fire Protection System impairment for the fire pump.
- c. Establish an hourly Fire Watch patrol in the affected areas within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- d. Establish another fire suppression water system as a backup within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ANSWER: 92 16634
- b. Initiate a Fire Protection System impairment for the fire pump.
"C" fire pump is NON-TRM. Only action is to initiate a fire impairment. There are no specific actions specified in 0.23 for the impairment.
REFERENCE:
TRM: T 3.11.2, TLCO 6.FP.101: Step 7.1 0.23: 6.2, 6.4, and Attachment 1 Distracter a:
T3.11.2 TLCO statement is still met. Conditions and actions are not entered.
Distracter d:
Actions if certain detection systems become inoperable.
Distracter a:
Correct response if E fire pump or diesel fire pump inoperable and not restored within 7 days or if the system is inoperable for other reasons than an inoperable fire pump (E or diesel).
Provide to Candidate: TLCO 3.11.2 Source: Direct 196
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 93 21435 00 07/09/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Administrative INT0070601, Determine consequences of thermal limit out of limits and determine required actions when thermal limit is exceeded.
Related Lessons INT0070601 TRM - Overview, Reactor Power Distribution, Reactor Coolant and Refueling Related Objectives INT0070601001040A Given plant conditions and the TRM, determine ACTIONS required per the following TLCOs: T.3.2.1 Linear Heat Generation Rate INT0070601001020A Discuss the applicable Bases associated with each of the following TRM Limiting Conditions for Operation (TLCOs): T.3.2.1 Linear Heat Generation Rate INT0070601001030A Given plant conditions, determine if the following TLCOs are met:
T.3.2.1 Linear Heat Generation Rate Related References 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
Related Skills (K/A) 290002.A2.05 Ability to (a) predict the impacts of the following on the REACTOR VESSEL INTERNALS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of...: (CFR: 41.5 / 45.6)
?Exceeding thermal limits (3.7/4.2) 197
QUESTION: 93 21435 (1 point(s))
The plant is operating at 95% power when a loss of feedwater heating occurs. Feedwater temperature drops several degrees and reactor power increases to 98%. A Gardel Periodic case is demanded and the following were noted for the most limiting core locations:
MAPRAT 0.992 MFLPD 1.001 MFLCPR 0.988 What potential consequence could result from continued operation with these conditions?
What action is required?
- a. Fuel clad cracking due to high stress.
Immediately initiate action to restore operation to within limits.
- b. Fuel clad cracking due to lack of cooling.
Restore operation to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- c. Fuel clad cracking due to high stress.
Restore operation to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- d. Fuel clad cracking due to lack of cooling.
Immediately initiate action to restore operation to within limits.
ANSWER: 93 21435
- a. Fuel clad cracking due to high stress.
Immediately initiate action to restore operation to within limits.
Provide TLCO 3.2.1, TS 3.2.1 and TS 3.2.2.
EXPLANATION OF ANSWER: With MFLPD > 1.0, LHGR is exceeding its limit. TLCO 3.2.1 requires immediate action to restore operation to within limits.
Distractors:
- b. is incorrect because the failure mode when LHGR is exceeded is high stress. and immediate action is required by TLCO 3.2.1. .
- c. is incorrect because with LHGR out of limits TLCO 3.2.1 requires entry..
- d. is incorrect because the failure mechanism for LHGR exceeding its limit is clad cracking due to high stress.
Source: New 198
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 94 5574 04 07/02/2004 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Technical Specifications, INT0070601, TRM Chemistry - 7% power ascension continue?
ODAM, TRM Related Lessons INT0070601 TRM - Overview, Reactor Power Distribution, Reactor Coolant and Refueling Related Objectives INT00706010010100 Explain the application of Improved Technical Specifications ITS sections 1.0 and 3.0 to the CNS Technical Requirements Manual.
INT0070601001020B Discuss the applicable Bases associated with each of the following TRM Limiting Conditions for Operation (TLCOs): T 3.4.1 RCS Chemistry INT0070601001030B Given plant conditions, determine if the following TLCOs are met:
RCS Chemistry INT0070601001040B Given plant conditions and the TRM, determine ACTIONS required per the following TLCOs: T.3.4.1 RCS Chemistry Related References 2.1.1 Startup Procedure T3.4.1 RCS Chemistry 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
Related Skills (K/A) 2.1.34 Ability to maintain primary and secondary plant chemistry within allowable limits. (CFR: 41.10 / 43.5 / 45.12) (2.3/2.9) 199
QUESTION: 94 5574 (1 point(s))
The plant is in the process of a reactor startup, at approximately 7% power when the following occur:
% At 08:00 on 7/2 the Reactor Coolant Continuous Conductivity Monitor indicates that conductivity is 2.3 micro mho/cm and lowering slowly.
Does Procedure 2.1.1 allow the power ascension to continue?
How long is allowed to restore conductivity to within limits?
- a. No. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />
- b. No. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- c. Yes. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />
- d. Yes. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ANSWER: 94 5574
- a. No. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Provide TRM Table 3.4.1-1 to the Candidate.
TRM 3.4.1 Chemistry limits are exceeded for condition 2 of Table 3.4.1-1, < 10% power and >
212°F and CONDITION A requires restoring RCS Chemistry to within limits with a completion time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time the limit was exceeded. The note in the actions section states TLCO 3.0.4 is not applicable so the power ascension could continue, however, Procedure 2.1.1 directs you to hold the startup until chemistry is within spec.
Source: Direct 200
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 95 19366 00 09/15/2003 10/07/2006 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Administrative INT0231001, S/D Risk Mngt - determine which activity can continue Related Lessons INT0231001 OPS Shutdown Risk Management Related Objectives INT02310010000300 Given specific plant configurations, determine if any deviations from outage guidelines exist Related References 0.50 Outage Management Program Related Skills (K/A) 2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations. (CFR: 43.5 / 45.13) (2.3/3.6) 201
QUESTION: 95 19366 (1 point(s))
The plant is in day 21 of a scheduled outage with the following conditions.
% RHR loop A is operating in shutdown cooling with RHR pump A.
% RHR pump C is inoperable due to a breaker malfunction discovered last shift.
% The plant is in MODE 5 with the fuel pool gates removed.
The following activities are scheduled to continue through the upcoming shift:
% Cleaning and inspection of MCC-R.
% 4160 BKR EG2 removal for inspection and repair.
Which of the following activities scheduled to be worked must be postponed?
- a. 250 VDC B bus maintenance.
- c. Replacement of the CS Pump B breaker.
- d. Replacement of CRD drive filter housings.
ANSWER: 95 19366
- a. 250 VDC B bus maintenance.
250 VDC B should not be worked at this time. RHR-MO-18 is de-energized due to the MCC-R work and 250 VDC B powers RHR-MO-17. PCIS requirements would not be met.
- b. No backup reactivity management method is required at this time.
- c. Replacement of the CS-B breaker would be allowed at this time.
- d. CRD can be worked at this time because adequate inventory control systems are available.
Source: Direct 202
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 96 20517 00 08/06/2005 10/07/2006 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Administrative Knowledge of the refueling process.
Related Lessons NONE Related Objectives NONE Related References 10.25 Refueling - Core Unload, Reload, and Shuffle 10CFR55.43 (7) Fuel handling facilities and procedures.
Related Skills (K/A) 2.2.32 Knowledge of the effects of alterations on core configuration. (CFR: 43.6)
(2.2/3.3) 203
QUESTION: 96 20517 (1 point(s))
The reactor has been shutdown for 7 days with core fuel shuffling activities in progress. During the shuffle two (2) bundles are maintained around each operable SRM. In addition to the bundles around the SRMs, additional fuel is maintained in the core during the shuffle.
What purpose does this additional fuel serve during the shuffle?
- a. Ensure the cooling capacity of the FPC system is not exceeded.
- b. Ensures maintenance of the shutdown Keff of the Spent Fuel Storage Pool (SFSP).
- c. Provide neutronic coupling of the core to ensure indication of neutron population at the SRM detectors.
- d. Limits the I-131 inventory in the Spent Fuel Storage Pool (SFSP) so that the limits of 10CFR100 would not be exceeded during a long duration loss of Fuel Pool Cooling.
ANSWER: 96 20517
- c. Provide neutronic coupling of the core to ensure indication of neutron population at the SRM detectors.
Reloading sequences are designed to ensure proper neutron flux monitoring. During core shuffle activities, at least two fuel bundles shall remain around each operable SRM with sufficient fuel remaining in the core at proper locations to ensure adequate core coupling.
Distractors:
- a. is incorrect because sufficient cooling capacity exists with RHR to cool the pool if the core were discharged.
- b. is incorrect as the spent fuel pool would have remain shutdown with the entire core offloaded.
- d. is incorrect as this is not a consideration for leaving fuel in the vessel.
Repeat from the 2005 NRC Exam Source: Direct 204
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 97 20551 0 04/29/2004 10/07/2006 Licensed RO: N Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Emergency Plan Authorization of Dose in Excess of 10CFR20 Limits Related Lessons GEN0030401 Emergency Plan for Licensed Operators Related Objectives GEN0030401E0E0100 State the major differences in 10CFR20 and EPA-400 derived TEDE values.
GEN0030401F0F1200 Discuss precautions and limitations of 5.7.17, EMERGENCY RADIATION EXPOSURE CONTROL.
GEN0030401F0F1300 From memory, given conditions, determine if Emergency Exposures should be authorized, and if so, for whom.
Related References 5.7.12 Emergency Radiation Exposure Control Related Skills (K/A) 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.
(CFR: 41.12 / 43.4. 45.9 / 45.10) (2.6/3.0) 205
QUESTION: 97 20551 (1 point(s))
What minimum level of authority can authorize radiological exposures in excess of 10CFR20 limits? (Select the first position that has the authority to authorize the exposure).
- a. Control Room Supervisor (non-emergency)
- b. Shift Manager (non-emergency)
- c. Emergency Director (emergency)
- d. Radiological Manager (emergency)
ANSWER: 97 20551
- c. Emergency Director Only the Emergency Director has the authority to authorize exposures in excess of occupational limits.
Distractors:
- a. is incorrect because the only the emergency director can authorize exposures in excess of occupational limits. Although in a very rare circumstance the Control Room Supervisor may act as the emergency director, the question asked what position could authorize the exposure and the only position that can authorize the exposure is the Emergency Director.
- b. is incorrect because the only the emergency director can authorize exposures in excess of occupational limits. Although initially the Shift Manager will be the emergency director, the position of shift manager cannot authorize the exposure.
- d. is incorrect because only the emergency director can authorize the exposure.
SRO Justification: SRO personnel fulfill the role of emergency director initially during an emergency, ROs cannot perform this function.
Source: Direct 206
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 98 20526 0 04/02/2004 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Administrative Liquid Release Authorization/Approval and Actions for a Lost CW Pump During Release Related Lessons INT0320115 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures (Formal Classroom/Pre-OJT Training)
Related Objectives INT0320115B0B0100 State who, by title, authorizes releases of radioactive liquid effluents from CNS.
INT0320115B0B0300 State the number of Circulating Water Pumps required to be in service during liquid radioactive discharges.
Related References 8.8.11 Liquid Radioactive Waste Discharge Authorization 10CFR55.43 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
Related Skills (K/A) 2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g. / waste disposal and handling systems). (CFR: 43.4 /
45.10) (1.8/2.9) 207
QUESTION: 98 20526 (1 point(s))
The plant is operating at low power with 2 Circulating Water pumps running. De-icing is in progress. The Radwaste Operator indicates that the Floor Drain Sample Tank requires discharging.
- 1) Whose approvals/authorizations is/are required in order to accomplish this discharge?
- 2) If one of the two operating circulating water pumps trip during the discharge, what action, if any, is required and why?
- a. Chemistry department authorizes the release and the duty Shift Manager approves the release.
Continue the discharge sufficient dilution flow exists.
- b. Duty Shift Manager authorizes and approves the release.
Continue the discharge sufficient dilution flow exists.
- c. Chemistry department authorizes the release and the duty Shift Manager approves the release.
Terminate the discharge insufficient dilution flow exists.
- d. Duty Shift Manager authorizes and approves the release.
Terminate the discharge insufficient dilution flow exists.
ANSWER: 98 20526
- c. Chemistry department authorizes the release and the duty Shift Manager approves the release.
Terminate the discharge insufficient dilution flow exists.
Answer c is correct because procedure 8.8.11 requires that chemistry authorizes the release and the duty Shift Manager approves the release. The loss of one CW pump would reduce flow to less than the minimum required and the discharge should be terminated.
Distractors:
- a. is incorrect because the discharge should be terminated.
- b. is incorrect because the discharge should be terminated and the chemistry department authorizes the release.
- d. is incorrect because chemistry authorizes the release.
SRO Justification: This is an SRO only item because in accordance with procedure 8.8.11 only the duty Shift Manager can approve liquid radioactive releases.
Source: Direct 208
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 99 21436 00 06/25/2006 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description Abnormal/Emergency INT0320107, Knowledge of abnormal condition procedures.
Procedures Related Lessons INT0320107 ABNORMAL CONDITION PROCEDURES Related Objectives INT03201070000100 Given a list of symptoms, identify the abnormal condition INT03201070000200 Given a specific procedure title, describe the automatic actions listed in the procedure INT03201070000300 Given a specific procedure title, or adequate information of plant conditions and indications, analyze the immediate actions required INT03201070000400 Given a specific procedure title, appraise the key concepts from the discussion section INT03201070000500 Given a specific procedure, explain or analyze any NOTES and CAUTIONS addressed in the procedure Related References 10CFR55.43 (2) Facility operating limitations in the technical specifications and their bases.
10CFR55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Related Skills (K/A) 2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13)
(3.4/3.6) 209
QUESTION: 99 21436 (1 point(s))
A plant startup is in progress with power at 19%. The following data is noted:
% Reactor water conductivity is 0.5 µmho (RWCU-CR-132, REACTOR WATER CONDUCTIVITY)
% Reactor water sulfates are 225 ppb.
% Reactor water chloride is 25 ppb.
% Condensate F/D combined outlet reaches is 2.2 µmho on CF-CR-25 (Channel 2)
What procedure(s) are required to be entered?
Why is action required for these plant conditions?
- a. 2.4CHEM and 2.1.5 Reactor Scram Combined F/D outlet conductivity requires immediate action to prevent degradation of Reactor Water Chemistry.
- b. 2.4CHEM ONLY High sulfates increase probability of Inter-granular Stress Corrosion Cracking
- c. 2.4CHEM ONLY Combined F/D outlet conductivity requires immediate action to prevent degradation of Reactor Water Chemistry.
- d. 2.4CHEM and 2.1.5 Reactor Scram High sulfates increase probability of Inter-granular Stress Corrosion Cracking ANSWER: 99 21436 ANSWER:
- a. 2.4CHEM and 2.1.5 Reactor Scram Combined F/D outlet conductivity requires immediate action to prevent degradation of Reactor Water Chemistry.
Answer source: 2.4CHEM The High sulfates require that entry into 2.4CHEM. IAW 2.4CHEM if Combined Filter Demineralizer conductivity reaches 2.0 µmho enter 2.1.5. This prompt action prevents a more rapid degradation of reactor water chemistry.
210
Distractors:
- b. is incorrect because entry into 2.1.5 is also required because combined F/D conductivity is high.
- c. is incorrect because entry into 2.1.5 is also required because combined F/D conductivity is high.
- d. is incorrect because the high sulfates do not require entry into 2.1.5.
Source: New 211
Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 100 19935 02 03/17/2004 10/07/2006 NRC Style RO: N Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?
Level Level Value Time 4 2 1 8 Multiple Choice Topic Area Description Technical Requirements INT0070509, Cascade from 3.8.1 to PAM Manual Related Lessons INT0070501 OPS Introduction to Technical Specifications INT0070509 OPS Tech. Spec. 3.8, Electrical Power Systems INT0070504 CNS Tech. Spec. 3.3, Instrumentation Related Objectives INT00705040010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.
INT00705090010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.8 LCO, determine the ACTIONS that are required.
INT00705010010200 Given plant conditions and a Specification, apply the rules of Section 3.0 to determine appropriate actions.
Related References 3.8.1 AC Sources - Operating 3.3.3.1 Post accident monitoring (PAM) instrumentation Related Skills (K/A) 2.4.3 Ability to identify post-accident instrumentation. (CFR: 41.6 / 45.4)
(3.5/3.8) 212
QUESTION: 100 19935 (1 point(s))
The plant was operating at rated power when the following occurred:
% NBI-LI-85A (Wide Range RPV water level) becomes inoperable at 0900 on 7/06.
% An air leak in the starting air system for DG2 occurs at 1200 on 7/12. Air pressure in both receivers lowers to 100 psig.
IF conditions do not change, what is the EARLIEST date and time that Technical Specifications requires the plant to enter MODE 3?
- a. 1600 on 7/12.
- b. 1600 on 7/17.
- c. 2400 on 7/19.
- d. 0400 on 7/20.
ANSWER: 100 19935
- c. 2400 on 7/19.
Proved 3.8.3 and 3.3.3.1 to the candidate.
3.8.3.F requires DG2 be declared inoperable immediately.
NBI-LI-85A (Wide Range RPV water level PAM instrument powered from CCP-1A) is inoperable. DG #2 becomes inoperable requiring 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> later, the Conditions and Required Actions for both Wide Range RPV water level PAM instruments inoperable (3.3.3.1 Condition "C") must be entered as the inoperable PAM instrument on a division opposite that of the inoperable DG. NBI-LI-85B is powered by CCP which is supported by DG #2. NBI-LI-85A is "an inoperable redundant required feature supported by the other DG".
Enter 3.3.3.1.C at 1600 on 7/12. Enter 3.3.3.1.D at 1600 on 7/19. Enter 3.3.3.1.E at 1600 on 7/19. Be in MODE 3 0400 on 7/20 DG is more limiting (3.8.1). Be in MODE 3 by 2400 on 7/19 Source: Direct 213