ML032890603
| ML032890603 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/29/2003 |
| From: | Lanksbury R NRC/RGN-III/DRS/OLB |
| To: | |
| References | |
| 05-266/03-301, 05-301/03-301 | |
| Download: ML032890603 (98) | |
Text
U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: MASTER Region: III Date: September 29, 2003 Facility/Unit: POINT BEACH U1 & U2 License Level: SRO Reactor Type: W Start Time: Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent.
Examination papers will be collected five hours after the examination starts.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicants Signature Results Examination Value 100.0 Points Applicants Score __________ Points Applicants Grade __________ Percent
GENERAL GUIDELINES
- 1. [Read Verbatim] Cheating on any part of the examination will result in a denial of your application and/or action against your license.
- 2. If you have any questions concerning the administration of any part of the examination, do not hesitate asking them before starting that part of the test.
- 3. SRO applicants will be tested at the level of responsibility of the senior licensed shift position (i.e., shift supervisor, senior shift supervisor, or whatever the title of the position may be).
- 4. You must pass every part of the examination to receive a license or to continue performing license duties. Applicants for an SRO-upgrade license may require remedial training in order to continue their RO duties if the examination reveals deficiencies in the required knowledge and abilities.
- 5. The NRC examiner is not allowed to reveal the results of any part of the examination until they have been reviewed and approved by NRC management. Grades provided by the facility licensee are preliminary until approved by the NRC. You will be informed of the official examination results about 30 days after all the examinations are complete.
- 6. After you complete the examination, 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 examination.
- 7. To pass the examination, you must achieve a grade of 80.00 percent or greater; grades will not be rounded up to achieve a passing score. Every question is worth one point.
- 8. For an initial examination, the nominal time limit for completing the examination is six hours; extensions will be considered under extenuating circumstances.
- 9. You may bring pens, pencils, and calculators into the examination room. Dark pencil should be used to facilitate machine grading.
- 10. Print your name in the blank provided on the examination cover sheet and the answer sheet.
- 11. Mark your answers on the answer sheet provided and do not leave any question blank.
- 12. If you have any questions concerning the intent or the initial conditions of a question, do not hesitate asking them before answering the question. Ask questions of the NRC examiner or the designated facility instructor only. When answering a question, do not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question. For example, you should not assume that any alarm has activated unless the question so states or the alarm is expected to activate as a result of the conditions that are stated in the question. Finally, answer all questions based on actual plant operation, procedures, and references. If you believe that the answer would be different based on simulator
operation or training references, you should answer the question based on the actual plant.
- 13. Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating.
- 14. When you complete the examination, assemble a package including the examination cover sheet and your scan-tron sheet and give it to the examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. Leave all other paper and reference materials at your desk, it will be disposed of immediately after the examination.
- 10. After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.
- 11. Do you have any questions?
SENIOR REACTOR OPERATOR Page 4 QUESTION: 001 (1.00)
Given the following conditions:
- Unit 1 is at 100% power.
- Air in-leakage to the condenser has resulted in steadily degrading condenser vacuum.
- A load reduction is directed in order to maintain vacuum.
- With the unit at approximately 85% power, a manual reactor trip is ordered due to the inability to maintain vacuum.
- All systems function as designed.
Based solely on the information given, which of the following describes the notification requirements for this event?
- a. The State/County must be notified within 15 minutes of the trip due to reaching an Emergency Plan classification for an Unusual Event.
- b. Kewaunee Nuclear Power Plant must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in order to ensure grid stability is maintained.
- c. The NRC must be notified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to manual actuation of the Reactor Protection System.
- d. No notifications to any outside agencies are required for these conditions.
SENIOR REACTOR OPERATOR Page 5 QUESTION: 002 (1.00)
Given the following conditions for Unit 1:
- RCS TAVG 547°F
- RCS pressure 2235 psig
- Reactor Trip breakers Both Open Vibrations on 1P-1A, Reactor Coolant Pump, have steadily increased over the shift and the pump has just been secured. Immediately after securing 1P-1A, the following plant conditions are noted:
- RCS TAVG 548°F
- RCS pressure 2240 psig
- SG 'A' pressure 1005 psig
- SG 'B' pressure 1013 psig
- SG 'A' NR level 40%
- SG 'B' NR level 25%
Using the given references, which of the following describes the Technical Specification implications for these conditions?
- a. Only RCS Loop A is inoperable.
It must be restored in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- b. Both RCS Loops are inoperable.
LCO 3.0.3 must be entered immediately since this condition is not addressed.
The plant must be placed in Mode 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
- c. Neither RCS Loop is inoperable.
No actions are required since one RCP is still in operation which satisfies the requirements of both loops.
- d. Both RCS loops are inoperable.
Immediate actions are required to restore one loop to an operable status, verify control rods are incapable of being withdrawn, and verify suspension of operations that may cause an RCS dilution.
SENIOR REACTOR OPERATOR Page 6 QUESTION: 003 (1.00)
Unit 2 is operating at 100% power.
At 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, a major transient results in a Reactor Trip and Safety Injection.
At 0310 hours0.00359 days <br />0.0861 hours <br />5.125661e-4 weeks <br />1.17955e-4 months <br />, the following plant conditions are noted:
- RCS pressure 1400 psig
- RCS TAVG 500°F
- Pressurizer level 75%
- Safety Injection NOT reset All equipment operated per design.
With respect to LCO 3.4.9, Pressurizer, which of the following describes the required actions?
- a. Restore required pressurizer heaters to an operable status by 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />.
- b. Restore Pressurizer water level to within the limit by 0410 hours0.00475 days <br />0.114 hours <br />6.779101e-4 weeks <br />1.56005e-4 months <br />.
- c. The plant is required to be in Mode 4 by 1510 hours0.0175 days <br />0.419 hours <br />0.0025 weeks <br />5.74555e-4 months <br />.
- d. The plant is required to be in Mode 3 by 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> AND Mode 4 by 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />.
QUESTION: 004 (1.00)
Which of the following is the basis for terminating Safety Injection flow when the criteria are satisfied during the performance of EOP-3, Steam Generator Tube Rupture?
- a. Prevent overcooling the RCS.
- b. Prevent solid plant operations.
- c. Prevent exhausting RWST level.
- d. Prevent overfilling the ruptured SG.
SENIOR REACTOR OPERATOR Page 7 QUESTION: 005 (1.00)
Given the following conditions:
- Unit 1 is at 100% power.
- A major fault occurs on 1X-04, Low Voltage Station Auxiliary Transformer, resulting in a Sudden Pressure lockout.
- All four Emergency Diesel Generators have failed to start.
- 1A-05, 4160 VAC bus, has indications of a lockout and major damage.
The crew has entered ECA-0.0, Loss Of All AC Power. They have been unsuccessful in:
- starting an emergency diesel from the control room.
- cross-tying of the 1A-03 and 2A-03 buses.
- cross-tying of the 1A-04 and 2A-04 buses.
The crew has just placed all pump control switches in pullout.
Based on this information, which of the following attachments in ECA-0.0 will be selected to restore a power source?
- a. Attachment A, G-01 Local Manual Start.
- b. Attachment C, G-03 Local Manual Start.
- c. Attachment E, Power Restoration Using Gas Turbine.
- d. Attachment F, Backfeed To 480 VAC Safeguards Buses.
SENIOR REACTOR OPERATOR Page 8 QUESTION: 006 (1.00)
Unit 2 was operating at 75% power when a plant transient resulted in a Reactor Trip and Safety Injection. EOP-0, Reactor Trip Or Safety Injection, has been entered and the crew is carrying out actions of the procedure. The following plant conditions are noted:
- RCS pressure 1100 psig and slowly lowering.
- Pressurizer level 5% and slowly lowering.
- Pressurizer PORVs Closed.
- Spray valves Closed.
- Steam Generator levels Normal.
- Steam Generator pressures Normal.
- Containment pressure Normal.
- Containment radiation Normal.
- Sump A level Normal.
- RE-214, PAB Exhaust Monitor Rising.
- Several PAB area radiation monitors Rising.
After assessing these conditions, the next procedure the DOS will implement is:
- c. ECA-1.2, LOCA Outside Containment.
- d. ECA-3.1, SGTR With Loss Of Reactor Coolant - Subcooled Recovery Desired.
SENIOR REACTOR OPERATOR Page 9 QUESTION: 007 (1.00)
Unit 1 was operating at 100% power with 1P-10A, RHR pump, tagged out for seal replacement.
The following sequence of events occurs:
- Large break LOCA.
- The B SI/RHR train has been fully aligned and placed on containment sump recirculation.
- The crew is currently implementing ECA-1.1, Loss of Containment Sump Recirculation.
- Foldout page criteria for securing ANY pumps has NOT been met.
- Containment pressure is 55 psig.
Given the attached reference from ECA-1.1, Loss Of Containment Sump Recirculation, which of the following indicates the REQUIRED correct combination of Containment Accident Recirculation Fans and Containment Spray Pumps to operate under these conditions?
- a. 1 Accident Fan, 0 Spray Pumps.
- b. 1 Accident Fan, 2 Spray Pumps.
- c. 3 Accident Fans, 1 Spray Pump.
- d. 4 Accident Fans, 0 Spray Pumps.
QUESTION: 008 (1.00)
Unit 1 is at 90% power. In response to alarms and control board indications, the crew has determined that Quadrant Power Tilt Ratio is 1.04. Using the given references, which of the following states the required power reduction?
- a. No power reduction is required.
- b. Power must be reduced to less than or equal to 88%.
- c. Power must be reduced to less than or equal to 78%.
- d. Power must be reduced to less than or equal to 50%.
SENIOR REACTOR OPERATOR Page 10 QUESTION: 009 (1.00)
Unit 1 is operating at 100% power. 1LT-426, Pressurizer Level Transmitter, was removed from service yesterday due to a suspected transmitter failure. All other equipment is in a normal alignment and there are no plant tests or surveillances in progress.
The following conditions now exist:
- 1P-2A, Charging Pump, is in Automatic.
- 1P-2C, Charging Pump, is in Manual.
- 1LI-427, Pressurizer Level, is slowly lowering.
- 1LI-428, Pressurizer Level, is slowly rising.
- 1P-2A, Charging Pump, speed is lowering.
- 1FI-128, Charging Flow, is lowering.
Based on the above indications, which of the following describes the failure that is occurring and the procedure the DOS will use to address the failure?
(AOP-1D, Chemical and Volume Control System Malfunction) (AOP-24, Response To Instrument Malfunctions)
- a. 1HC-428C, C Charging Pump Controller, is failing.
Enter AOP-1D to address the failure.
- b. 1LT-428, Pressurizer Level Transmitter, is failing.
Enter AOP-24 to address the failure.
- c. 1LT-427, Pressurizer Level Transmitter, is failing.
Enter AOP-24 to address the failure.
- d. 1HC-428A, A Charging Pump Controller, is failing.
Enter AOP-1D to address the failure.
SENIOR REACTOR OPERATOR Page 11 QUESTION: 010 (1.00)
An accidental release of a Waste Gas Decay tank is occurring. The release started just over an hour ago and cannot be terminated. After review of the Emergency Plan, the Shift Manager declared an Unusual Event due to meeting the following Emergency Action Level (EAL): "Vent radiation reading(s) exceed the high alarm setpoint for >60 minutes."
NO EALs for an ALERT or higher classification have been reached.
Which of the following indicates the reason an Unusual Event was declared and the Emergency Plan implemented for these conditions:
- a. Protective Action Recommendations are required to protect the health and safety of the public.
- b. Assembly, accountability, and evacuation of unnecessary personnel is required to protect plant personnel.
- c. The release indicates a degradation in plant control and a potential degradation in the level of safety.
- d. The conditions are indicative of radiation limits at the site boundary in excess of 10CFR20 limits.
SENIOR REACTOR OPERATOR Page 12 QUESTION: 011 (1.00)
Given the following plant conditions:
- A loss of coolant accident occurred about 15 minutes ago.
- During the initial phases of the accident, containment pressure peaked at 15 psig and containment radiation dose rate peaked at 106 R/hr.
- The DOS has directed that adverse containment numbers be used during EOP implementation.
- Approximately 30 minutes later, containment pressure lowered to 6 psig and containment radiation dose rate lowered to 8E4 R/hr.
The DOS must direct that adverse containment numbers:
- a. still be used until containment pressure is less than 5 psig.
- b. not be used since containment pressure is no longer indicative of adverse containment conditions.
- c. still be used until relaxed by Technical Support Center personnel.
- d. not be used since the radiation level is no longer indicative of adverse containment conditions.
SENIOR REACTOR OPERATOR Page 13 QUESTION: 012 (1.00)
Given the following plant conditions:
- A loss of power to all non-safety 4160 Volt buses has resulted in a Unit 1 Reactor Trip.
- Off-site power is available to all safety related buses.
- The appropriate Emergency procedures were implemented.
- RCS temperature is currently 325°F.
- All safety related equipment is operable.
The current procedure in use is EOP-0.2, Natural Circulation Cooldown. While implementing the steps of this procedure, a check is made to determine whether or not one train of Safety Injection (SI) should be removed from service.
Which of the following indicates the actions the DOS will direct when implementing this step?
The DOS will direct:
- a. the 'A' SI Train to be isolated, the 'B' SI Train is the preferred train to keep operable since it can inject via two flowpaths.
- b. the 'B' SI Train to be isolated, the 'A' SI Train is the preferred train to keep operable since it can be used to fill the accumulators.
- c. that neither SI Train be isolated since Technical Specifications require both trains operable for these conditions.
- d. that either SI Train be isolated, one train is not preferred over the other since off-site and on-site power to the safety related buses is available.
SENIOR REACTOR OPERATOR Page 14 QUESTION: 013 (1.00)
Unit 2 is operating at 100% power and is in a normal full power alignment. The Unit 2 CO reported that 2CV-142, Charging Flow Control Valve, has failed full open.
Which of the following is the expected plant response for this valve failing open, including the proper procedure to address this failure?
(AOP-24, Response To Instrument Malfunctions) (AOP-1D, Chemical and Volume Control System Malfunctions)
- a. Reactor Coolant Pump labyrinth seal delta-P will rise.
Pressurizer level will steadily rise.
Implement AOP-1D.
- b. Reactor Coolant Pump labyrinth seal delta-P will rise.
Pressurizer level will remain constant.
Implement AOP-24.
- c. Reactor Coolant Pump labyrinth seal delta-P will lower.
Pressurizer level will steadily rise.
Implement AOP-24.
- d. Reactor Coolant Pump labyrinth seal delta-P will lower.
Pressurizer level will remain constant.
Implement AOP-1D.
SENIOR REACTOR OPERATOR Page 15 QUESTION: 014 (1.00)
The Residual Heat Removal System, along with other ECCS subsystems, ensures that the ECCS Acceptance Criteria of 10CFR50.46 is met. All of the following are part of this criteria EXCEPT:
- a. The total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
- b. The maximum fuel pellet centerline temperature shall not exceed 2000°F.
- c. The total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount generated if all of the metal in the cladding were to react.
- d. Changes in core geometry shall be such that the core remains amenable to cooling.
SENIOR REACTOR OPERATOR Page 16 QUESTION: 015 (1.00)
Given the following plant conditions:
- Unit 1 was operating at 100% power when a loss of coolant event occurred.
- The operating crew has just entered EOP-1.3, Transfer To Containment Sump Recirculation - Low Head Injection.
- Shortly after initiating EOP-1.3, it is identified that the CCW surge tank is at 8%
and continuing to lower.
- The DOS directs that both CCW pumps be placed in pullout.
- All other equipment is operating per design.
Based on these conditions, which of the following correctly describes the impact of these events during subsequent actions to establish containment sump recirculation?
(EOP-1.3, Transfer To Containment Sump Recirculation - Low Head Injection)
(AOP-9B, Component Cooling System Malfunction)
(ECA-1.1, Loss Of Containment Sump Recirculation)
EOP-1.3 will be utilized as a secondary priority until CCW is restored.
A transition to ECA-1.1 must be made immediately.
AOP-9B will be utilized as a secondary priority to address the CCW leak.
SENIOR REACTOR OPERATOR Page 17 QUESTION: 016 (1.00)
Unit 1 was operating at 100% power. The Reactor/Turbine tripped on Low Pressurizer Pressure due to a design basis loss of coolant accident. The following conditions are noted:
- Subcooling margin is less than 0°F.
- Operators are responding using EOP-0, Reactor Trip Or Safety Injection.
- 1SW-2907, Containment Ventilation Cooler Outlet Emergency Flow Control Valve, cannot be opened manually or locally.
- All other equipment has operated per design.
- The third license (BOP) has just secured 1P-14A, Containment Spray Pump, per Attachment A of EOP-0.
Which of the following statements is correct with respect to the above conditions?
(CSP-Z.1, Response to High Containment Pressure)
- a. Containment design pressure will NOT be exceeded.
Only two Containment Accident Fan Coolers are available.
The BOP will continue with Attachment A.
- b. Containment design pressure will be exceeded.
No Containment Accident Fan Coolers are available.
The BOP will re-start 1P-14A and remain in Attachment A.
- c. Containment design pressure will be exceeded.
No Containment Accident Fan Coolers are available.
An immediate transition to CSP-Z.1 will be made.
- d. Containment design pressure will NOT be exceeded.
All four Containment Accident Fan Coolers are available.
The BOP will continue with Attachment A.
SENIOR REACTOR OPERATOR Page 18 QUESTION: 017 (1.00)
Given the following plant conditions:
- Unit 1 is at 29% power awaiting release from a Secondary side chemistry hold.
- All equipment is in a normal lineup for this power level.
- There are no plant evolutions in progress.
- Control rods are in Manual.
- The RED TAVG meter and the RED DT meter are both observed to be steadily rising, and both eventually peg high.
- All other TAVG and DT meters are steady
- No operator actions have occurred.
Using the given references, which of the following states the impact of this failure, including the Technical Specification implications?
- a. Pressurizer level will rise and stabilize at a higher level.
Technical Specification 3.3.1 is NOT met.
Technical Specification 3.3.2 is met.
- b. Pressurizer level will rise until a high level trip occurs.
Technical Specification 3.3.1 is NOT met.
Technical Specification 3.3.2 is met.
- c. Pressurizer level will remain unchanged.
Technical Specification 3.3.1 is NOT met.
Technical Specification 3.3.2 is NOT met.
- d. Pressurizer level will rise and stabilize at a higher level.
Technical Specification 3.3.1 is met.
Technical Specification 3.3.2 is NOT met.
SENIOR REACTOR OPERATOR Page 19 QUESTION: 018 (1.00)
Which one of the following is considered to be the most limiting event (time critical) concerning operation of the Atmospheric Dump Valves?
- a. Small break loss of coolant accident without a loss of off-site power.
- b. Steam generator tube rupture accident with a loss of off-site power.
- c. Large break loss of coolant accident without a loss of off-site power.
- d. Main Steam Line break accident inside containment with a loss of off-site power.
QUESTION: 019 (1.00)
Both units are at 100% power. The mid-shift crew is at minimum shift crew composition. At 0015, the Shift Manager (SM) is required to leave due to an emergency at home. Which of the following actions, if any, must be taken?
- a. No action is required as long as the Duty Operating Supervisor (DOS) and the Operating Supervisor (OS) remain in the Control Room.
- b. The DOS can assume the Shift Manager duties until the next shift arrives.
- c. The Duty and Call Superintendent (DCS) must report to the Control Room until minimum staffing requirements are met.
- d. The Shift Technical Advisor (STA) must report to the Control Room and shall remain there until minimum staffing requirements are met.
SENIOR REACTOR OPERATOR Page 20 QUESTION: 020 (1.00)
Given the following plant conditions:
- Both Units are operating at 100% power.
- Unit 2 is operating in the ice melt mode.
- Due to rising lake temperatures, the DOS (who has Command & Control) directs the third license to secure ice melt.
- The fourth license is in the WCC and unavailable.
- When securing ice melt, the third license requests a peer check for the required valve manipulations.
The DOS should:
- a. direct the Unit 1 CO to provide a peer check.
- b. direct the Unit 2 CO to provide a peer check.
- c. direct the OS to provide a peer check.
- d. provide the peer check.
SENIOR REACTOR OPERATOR Page 21 QUESTION: 021 (1.00)
Given the following plant conditions:
- A Unit 1 Reactor startup is about to begin.
- Preparations are being made to begin pulling the Shutdown Bank rods and entering Mode 2.
- A power supply fails on 1LT-460A, Steam Generator A Wide Range Level Transmitter.
Using the given references, which of the following statements correctly describes the Technical Specification requirements for entry into Mode 2 for this condition?
- a. Mode 2 can be entered and the startup continued.
The channel must be restored to an operable status within 30 days.
If the channel is NOT restored after 30 days, the plant must submit a special report within 14 days.
- b. Mode 2 can be entered and the startup continued.
The channel must be restored to an operable status within 30 days.
If the channel is NOT restored after 30 days, the plant must be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. Mode 2 CANNOT be entered.
Entry into Mode 2 is prohibited by LCO 3.0.4.
- d. Mode 2 CANNOT be entered.
Entry into Mode 2 is prohibited by LCO 3.0.2.
SENIOR REACTOR OPERATOR Page 22 QUESTION: 022 (1.00)
Given the following conditions:
- Maintenance has requested a tagout for 1A52-60, G-01 Diesel Generator To Bus 1A-05 Breaker, in order to perform an inspection.
- Bus 1A-05 was re-aligned to G-02, Emergency Diesel Generator, per OI-35A.
- The tagout consisted of taking the control switch to PULLOUT and racking out the breaker.
- A visual inspection showed that NO work is required.
- NO disassembly work was performed on the equipment.
In addition to ensuring the control switch for 1A52-60 is in Auto, which of the following indicates the minimum requirement(s) for restoring operability of 1A52-60?
1A52-60 can be considered operable when it is:
- a. racked in and the breaker is independently verified to be racked in.
- b. racked in and the breaker is administratively verified it was tested 28 days ago.
- c. racked in and TS-81, EDG G-01 Monthly Test, is performed.
- d. racked in and the green indicating light is ON.
SENIOR REACTOR OPERATOR Page 23 QUESTION: 023 (1.00)
Given the following plant conditions:
- Unit 2 is in Mode 5.
- A containment purge is in progress.
- 2RE-212, Containment Noble Gas Monitor, has just been declared inoperable and removed from service for calibration.
Using the given references, which of the following statements describes the impact, if any, of removing 2RE-212 from service?
- a. There are no required actions provided RE-305, Containment Purge Exhaust Noble Gas Monitor, is operable.
- b. Grab samples are required to be taken and analyzed every twelve hours in order for the purge to continue.
- c. Continuous sampling using auxiliary equipment is required in order for the purge to continue.
- d. The purge must be immediately secured, the requirements for monitoring the purge effluent cannot be met.
QUESTION: 024 (1.00)
The plant has experienced a major plant transient. An ORANGE path CSP is currently being implemented. The implementation of the ORANGE path CSP must be suspended for all of the following conditions EXCEPT when:
- a. a RED path CSP is identified.
- b. a higher priority ORANGE path CSP is identified.
- c. the ORANGE path condition clears.
- d. a total loss of onsite and offsite AC power occurs.
SENIOR REACTOR OPERATOR Page 24 QUESTION: 025 (1.00)
Given the following plant conditions:
- An accident has occurred that has resulted in a Reactor Trip and Safety Injection.
- The EOPs are currently being implemented.
- Prior to the trip, an AOP was being implemented to address plant equipment problems.
- A second SRO is now coordinating remaining actions in the AOP while the EOPs are being implemented.
- A conflict has arisen between the EOP and the AOP regarding an electrical lineup.
Which of the following describes the proper resolution of this conflict?
- a. Guidance in the EOP must be followed since it is the controlling procedure.
The conflicting guidance in the AOP will NOT be performed.
The conflicting guidance in the EOP will NOT be performed.
- c. The SRO will consult with the DCS and follow guidance in either the AOP or the EOP, depending on the specific situation.
- d. The SRO will invoke 10CFR50.54(x) and perform the procedure most appropriate for the situation.
SENIOR REACTOR OPERATOR Page 25 QUESTION: 026 (1.00)
A Unit 1 reactor trip has occurred for undetermined reasons. The operator is carrying out the Immediate Actions of EOP-0, Reactor Trip or Safety Injection. While performing Step 3, "Verify Safeguards Buses Energized", the operator notes the following:
- All off-site power has been lost to Unit 1.
- All four Emergency Diesel Generators have failed to auto-start.
- Annunciator C02 D 3-4, Unit 1 4.16kV Bus Lockout, is lit.
The operators next action is to:
- a. Immediately transition to ECA-0.0, Loss of All AC Power.
- c. Attempt to restore power to 1A-05 or 1A-06 by fast starting and loading either G-01 or G-03, Emergency Diesel Generator.
- d. Continue on to step 4, "Check if SI is Actuated", when immediate actions are complete, concurrently enter AOP-19, Safeguards Bus Restoration.
SENIOR REACTOR OPERATOR Page 26 QUESTION: 027 (1.00)
Given the following plant conditions:
- Unit 1 has just tripped from 100% power due to a small break LOCA that was caused by a stuck open Pressurizer PORV and Block Valve.
- During implementation of EOP-1, Loss of Reactor or Secondary Coolant, subcooling lowers to 15°F.
- The operating crew has just tripped both Reactor Coolant Pumps (RCPs).
Which of the following indicates the reason the RCPs were tripped by the crew?
- a. To prevent excessive RCS inventory loss.
- b. To minimize the cooldown rate.
- c. To prevent RCP damage from cavitation.
- d. To prevent an RCP motor over-current condition.
QUESTION: 028 (1.00)
Given the following Unit 1 plant conditions:
- The Unit has tripped from 100% due to a small break LOCA.
- Conditions have stabilized and operators are evaluating the criteria for terminating Safety Injection.
- Adverse containment conditions do NOT exist.
Which one of the following conditions would PREVENT SI termination per EOP-1.2, "Small Break LOCA Cooldown and Depressurization"?
- a. RCS subcooling is 40°F.
- b. Both Steam Generator levels are 40%.
- c. RCS pressure is 2050 psig.
- d. Pressurizer level is 9%.
SENIOR REACTOR OPERATOR Page 27 QUESTION: 029 (1.00)
Given the following plant conditions:
- A Large Break LOCA has occurred on Unit 1.
- EOP-1.3, Transfer To Containment Sump Recirculation
- Low Head Injection, is in progress.
- Containment sump recirculation utilizing RHR Pump B has just been established.
After sump recirculation is established, Safety Injection Pump B is required to be started within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> of the onset of the event.
Which of the following indicates the reason that SI Pump B is started?
- a. To prevent the formation of thermal stratification layers in the core.
- b. To reduce thermal stresses on RHR injection nozzles.
- c. To address boron precipitation concerns within the reactor vessel.
- d. To prevent excessive voiding in the reactor vessel head.
SENIOR REACTOR OPERATOR Page 28 QUESTION: 030 (1.00)
Unit 1 is operating at 100% power. The following indications are noted:
- Annunciator "1P-1A or B RCP Upper or Lower Sump Oil Level High or Low" is lit.
- Points 2 and 4 on recorder 1TR-2001 (1P-1A RCP Thrust Bearing Upper and Lower Shoe temperatures) are in alarm and are currently reading 92°C and rising.
- Unit 1 Component Cooling Water Surge Tank level is 49% and lowering.
- 1P-1A RCP seal injection flow is 6 gpm.
- 1P-1A RCP No. 1 seal leakoff is 1.2 gpm.
Which of the following describes the required action and the reason for the action that would explain all of the above abnormal conditions?
- a. Unit 1 Reactor must be tripped and 1P-1A RCP stopped because oil has leaked out of the pump resulting in poor bearing lubrication.
- b. The position of 1CC-761A, Thermal Barrier Outlet AOV, must be checked to ensure it is shut because the thermal barrier is leaking.
- c. Unit 1 Reactor must be tripped and 1P-1A RCP stopped because its oil has been emulsified with CCW and this has affected bearing lubrication.
- d. The Seal Return Heat Exchanger must be bypassed because CCW is diluting the RCS through a leak in the heat exchanger.
SENIOR REACTOR OPERATOR Page 29 QUESTION: 031 (1.00)
Unit 1 was operating at 100% power when a disk failure occurred on 1CV-304C, 1P-1A RCP Seal Injection Check Valve. The failure resulted in a complete loss of seal injection to 1P-1A.
1P-1A should be run no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in order to:
- b. prevent exceeding a TSAC for loss of flow path to RCP seals.
- c. minimize damage due to overheating 1P-1A #1 seal.
- d. prevent high flow damage to 1P-1B #1 seal.
QUESTION: 032 (1.00)
Given the following plant conditions:
- Unit 1 is in day 10 of a refueling outage.
- Unlatching of rods is in progress.
- Reactor Coolant System temperature (RHR inlet) is 90°F.
- The running RHR pump trips.
- The other RHR pump is tagged out for minor maintenance, but can be restored if needed.
Using the given references, which one of the following indicates the minimum time (number of hours) at which RCS boiling will occur?
- a. 165 minutes
- b. 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
- c. 18.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />
- d. 180 minutes
SENIOR REACTOR OPERATOR Page 30 QUESTION: 033 (1.00)
Given the following sequence of events:
- Unit 2 is operating at 100% power.
- 2P-11A, Component Cooling Water Pump, is running.
- 2P-11B, Component Cooling Water Pump, is in standby.
- A manual Reactor Trip was initiated due to a large feedwater leak in the Turbine Building.
- During performance of Immediate Actions of EOP-0, a manual Safety Injection is initiated due low Pressurizer Level.
- After completing all immediate actions, a lockout on Low Voltage Station Transformer 1X-04 occurs.
- All automatic actions occur and the safeguards buses are re-energized from the Emergency Diesel Generators.
Which of the following indicates the status of the Component Cooling Water Pumps after the buses are re-energized?
- b. Neither CCW Pump is running.
- d. Both CCW Pumps are running.
SENIOR REACTOR OPERATOR Page 31 QUESTION: 034 (1.00)
Given the following plant conditions:
- Unit 1 is operating at 100% power.
- Normal Letdown has been secured for maintenance.
- Excess Letdown is in service with one Charging Pump running in Manual at minimum speed per OP-5E, Establishing and Securing Excess Letdown.
- Charging and Excess Letdown flow have been balanced and Pressurizer level is stable.
- All other equipment is in a normal alignment.
- No other equipment is out of service.
1LT-428, Pressurizer Level Transmitter, fails low.
Assuming no operator action, what effect will this transmitter failure have on the following Pressurizer parameters?
- a. Pressurizer pressure rises.
Saturation temperature rises.
- b. Subcooling rises.
Saturation temperature lowers.
- c. Subcooling lowers.
Saturation temperature rises.
- d. Pressurizer pressure lowers.
Saturation temperature lowers.
SENIOR REACTOR OPERATOR Page 32 QUESTION: 035 (1.00)
Unit 1 was operating at 100% power when a Steam Generator Tube Rupture occurred.
The following plant conditions exist:
- RCPs Both secured
- Ruptured S/G identified and isolated
- RCS cooled down to target temperature and depressurized
- SI pumps secured
- Charging in service
- Letdown in service Preparations are being made to start a Reactor Coolant Pump (RCP). Which one of the following parameters is required to be verified acceptable prior to restarting the RCP in EOP-3, Steam Generator Tube Rupture?
- a. VCT Pressure
- b. # 2 Seal Delta-P
- c. #2 Seal Leakoff flow
- d. Standpipe level
SENIOR REACTOR OPERATOR Page 33 QUESTION: 036 (1.00)
Given the following Unit 1 plant conditions:
- The Unit has tripped from 100% due to a steam line break inside containment.
- Containment pressure peaked at 28 psig.
- All equipment functioned as designed.
- The crew has completed EOP-0, Reactor Trip or Safety Injection, and Attachment A of EOP-0.
- A transition to EOP-2, Faulted Steam Generator Isolation, has just been made.
- All procedural steps were performed without error.
The current status of the Containment Spray system is:
- a. both pumps running.
- b. both pumps secured.
- c. one pump secured with its suction valve shut.
- d. one pump secured with its discharge valves shut.
SENIOR REACTOR OPERATOR Page 34 QUESTION: 037 (1.00)
Given the following sequence of events:
- Unit 2 is operating at 15% power.
- Feedwater control has been transferred to the Main Feedwater Regulating Valves (operating in Auto) and all equipment is in a normal alignment for the given power level.
- The only running Main Feedwater Pump then trips.
- Steam Generator water levels are observed to be lowering.
- The standby Main Feedwater Pump cannot be started.
- S/G A level = 35% NR.
- S/G B level = 48% NR.
Which of the following describes the status of Feedwater System components for these conditions?
- a. Both Main Feed Regulating Bypass Valves shut.
Both Main Feed Regulating Valves open.
- b. Both Main Feed Regulating Bypass Valves open.
Both Main Feed Regulating Valves shut.
- c. Both Main Feed Regulating Bypass Valves open.
S/G A Main Feed Regulating Valve is open, B is shut.
- d. S/G A Main Feed Regulating Bypass Valve is open, B is shut.
Both Main Feed Regulating Valves open.
SENIOR REACTOR OPERATOR Page 35 QUESTION: 038 (1.00)
Given the following plant conditions:
- Unit 1 is responding to a loss of offsite power.
- Both Emergency Diesel Generators G-01 and G-03 have started and loaded onto their respective buses.
- Safety Injection did NOT actuate.
- Pressurizer level is 25%.
The Control Operator is attempting to control Pressurizer pressure. What must be done to energize 1T-1C, Backup Group C Heaters?
- a. Reset the 1B-03 Non-Safeguards Equipment lockout.
Leave the 1T-1C control switch in AUTO.
- b. Restore power to 1B-01.
Then turn the 1T-1C control switch to ON.
- c. Turn the 1T-1C control switch to OFF.
Then turn the 1T-1C control switch to ON.
- d. Place 1HC-431K, Pressurizer Pressure Controller, in Manual and increase setting.
Leave the 1T-1C control switch in AUTO.
SENIOR REACTOR OPERATOR Page 36 QUESTION: 039 (1.00)
Unit 1 and Unit 2 are at 100% power. Service Water Pumps P-32A, B, and F are running. The following annunciators are received:
- Service Water Strainers Delta-P High.
- North or South Service Water Header Pressure Low.
- G-01 Emergency Diesel Cooler Delta-P Low.
- G-02 Emergency Diesel Cooler Delta-P Low.
- Unit 2 Turbine Building Sump Level High.
Which of the following indicates the cause of these alarms and the appropriate remedial action?
(OI-70, Service Water System Operation) (AOP-9A, Service Water System Malfunction)
- a. The Unit 2 Turbine Building Zurn Strainer is clogged, utilize OI-70 to backwash the strainer.
- b. There is a leak in the North Service Water Header, utilize AOP-9A to isolate the leak.
- c. The North Service Water Header Strainer is clogged, utilize OI-70 to backwash the strainer.
- d. There is a leak in the South Service Water Header, utilize AOP-9A to isolate the leak.
SENIOR REACTOR OPERATOR Page 37 QUESTION: 040 (1.00)
Given the following plant conditions:
- A total loss of the Instrument Air System has occurred due to compressor failures.
- The Service Air to Instrument Air backup valves have failed to open and the Instrument Air System headers have completely depressurized.
Which of the following indicates the impact on 1P-29 Turbine Driven AFP Mini-Recirc Valve, 1AF-4002, and its recirculation line capability?
1AF-4002 will. . .
- a. continue to operate normally due to Service Air backup.
- b. continue to operate normally due to Nitrogen backup.
- c. operate for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> since it has an IA accumulator.
- d. NOT open, recirculation line capability is immediately lost.
SENIOR REACTOR OPERATOR Page 38 QUESTION: 041 (1.00)
Unit 1 was operating at 100% power, with no equipment out of service.
The following conditions exist:
- A LOCA has occurred outside Unit 1 containment.
- Both Safety Injection Trains have actuated.
- Both SI Pumps are running.
- Both RHR Pumps are running.
- The crew is currently in ECA-1.2, LOCA Outside Containment.
- Annunciator "Unit 1 or 2 RHR Pump Rooms Level High" is in alarm.
- The amber light above 1WL-4100, 1P-10B RHR Pump Drain To Sump, is lit.
- 1WL-4100 switch is in the OPEN position.
- P-40A, -19 Sump Pump, is running continuously.
(1P-10B, Residual Heat Removal Pump)
(1P-15B, Safety Injection Pump)
Based on the above indications, the crew is required to:
- a. Continue operating with all Safety Injection and RHR pumps.
All SI and RHR pumps are required to maintain core cooling.
Neither pump is required to maintain core cooling since the A Train SI and RHR pumps are operating.
- c. Immediately stop 1P-10B only.
1P-15B along with both A Train SI and RHR pumps are required to maintain core cooling.
- d. Immediately stop all Safety Injection and RHR pumps.
Only the Charging Pumps are required to ensure core cooling for these conditions.
SENIOR REACTOR OPERATOR Page 39 QUESTION: 042 (1.00)
Following a LOCA, the inability to open either Containment Sump B suction valve has resulted in a loss of containment sump recirculation. All other equipment has functioned normally.
ECA-1.1, Loss of Containment Sump Recirculation, is the procedure in effect. The following indications are noted:
- Containment Pressure = 15 psig
- VCT level = 52%
- RWST level = 4%
- RCS Pressure = 26 psig The crew has just determined that 100 gpm is the required minimum injection flow.
Based on these indications, which of the following actions will the crew take to maintain core cooling?
- a. Charging Pumps must be aligned to the VCT and started to establish 100 gpm charging flow.
- b. One Safety Injection Pump must be started and 100 gpm injection flow established by throttling the respective 1SI-866A/B (SI Pump Discharge Header MOV).
- c. Both Safety Injection Pumps must be started and 50 gpm each established by throttling both 1SI-866A and B.
- d. One RHR Pump must be started and 100 gpm established by throttling 1RH-625, RHR Hx Outlet Flow Control Valve.
SENIOR REACTOR OPERATOR Page 40 QUESTION: 043 (1.00)
Unit 1 was operating at 100% power when a Reactor Trip and Safety Injection occurred.
Subsequent failures have also resulted in a total loss of all Auxiliary Feedwater. The crew has transitioned to CSP-H.1, Response to Loss of Secondary Heat Sink.
While attempting to restore 1P-29, Turbine Driven AFW Pump, using CSP-H.1, the following conditions are noted:
- 1MS-2019, 1P-29 Steam Supply valve - green light off, red light lit.
- 1MS-2020, 1P-29 Steam Supply valve - green light off, red light lit.
- 1MS-2082, 1P-29 Low Suction / - green light off, red light lit.
- Overspeed Trip valve operator
- 1MS-2082 Trip Valve Position - amber light lit, red light off.
- 1P-29 AFP Low Suction Pressure Trip annunciator is clear.
- Unit 1 Auxiliary Feedwater System Disabled annunciator is lit.
Which of the following is the reason 1P-29 is NOT running?
- d. 1P-29 attempted to start but did not because Trip Valve 1MS-2082 was manually shut from the Control Room.
SENIOR REACTOR OPERATOR Page 41 QUESTION: 044 (1.00)
With the Unit 1 Reactor operating at 60% power and turbine in IMP IN, the following indications are observed:
- Rising Steam Generator pressures.
- Rising Pressurizer pressure.
- TAVG greater than TREF and rising.
- Turbine Impulse Pressure constant.
- Rising NI Power.
Assuming no operator action, which of the following would initially explain the above indications?
- a. Turbine runback.
- b. Main steam line leak.
- c. Inadvertent AFW actuation.
- d. Uncontrolled rod withdrawal.
SENIOR REACTOR OPERATOR Page 42 QUESTION: 045 (1.00)
Given the following Unit 1 plant conditions:
- An inadvertent reactor trip has just occurred.
- Both reactor trip breakers indicate open.
- Control rod K-7 IRPI reads 220 steps, its rod bottom light is NOT lit.
- Control rod L-6 IRPI reads 35 steps, its rod bottom light is NOT lit.
- Control rod G-3 IRPI reads 10 steps, its rod bottom light is lit.
The current procedure in effect is EOP-0.1, "Reactor Trip Response".
Which of the following describes the minimum amount of boration required for these conditions?
- a. No boration is required since the reactor trip breakers are open.
- b. A 1200 gallon boration is required since only one control rod is considered not fully inserted.
- c. A 2400 gallon boration is required since only two control rods are considered not fully inserted.
- d. A 3600 gallon boration is required since three control rods are considered not fully inserted.
SENIOR REACTOR OPERATOR Page 43 QUESTION: 046 (1.00)
Given the following plant conditions:
- Unit 1 is in Mode 3, Hot Standby.
- 1CV-350, Emergency Boration Valve, is tagged shut for repair.
- An inadvertent dilution of the Reactor Coolant System has just been detected.
- The dilution has been terminated, however, a shutdown margin calculation has determined that the required shutdown margin is NOT met.
- After initiating boration, 1CV-110A, Boric Acid to Blender Flow Control Valve, is observed to be shut and cannot be re-opened.
Which one of the following boration flowpaths is immediately available to re-establish shutdown margin per OP-5B, Blender Operation/Dilution/Boration?
- a. Borate using the Charging Pumps and the RWST.
- b. Borate using the Charging Pumps and the Blender.
- c. Borate using the Safety Injection Pumps and the RWST.
- d. Borate using the Boric Acid Transfer Pumps, the Charging Pumps and the in-service Boric Acid Tank.
SENIOR REACTOR OPERATOR Page 44 QUESTION: 047 (1.00)
Given the following:
- A Unit 1 reactor startup is in progress per OP-1B, Reactor Startup.
- Reactor power is less than the P-6 interlock and the Intermediate Range Neutron Flux instruments are currently NOT required to be operable per Technical Specifications.
- 1N-35, Intermediate Range, then fails low.
Which of the following indicates the impact of this failure on reactor startup?
- a. Reactor power can be raised and maintained greater than the P-10 interlock until the channel is restored.
- b. Reactor power is limited to a maximum of 5% until the channel is restored.
- c. The failed channel can be bypassed and the startup can continue without restriction.
- d. Reactor power must be maintained less than the P-6 interlock until the channel is restored.
SENIOR REACTOR OPERATOR Page 45 QUESTION: 048 (1.00)
Unit 1 was operating at 100% power when a Steam Line Break occurred inside containment on Steam Generator A.
Several Containment Spray and Cooling system failures have resulted in the crew entering CSP-Z.1, Response To High Containment Pressure.
Which of the following actions in CSP-Z.1 helps to limit the containment pressure transient?
- a. Feed S/G A at only 50 gpm to minimize liquid inventory.
- c. Manually isolate feedwater to S/G A to minimize the mass and energy release.
- d. Perform a rapid cooldown of the RCS using S/G B to reduce the energy input to containment.
SENIOR REACTOR OPERATOR Page 46 QUESTION: 049 (1.00)
Given the following plant conditions:
- Unit 1 is operating at 100% power.
- A small amount of Steam Generator tube leakage is present on both Steam Generators. However, the leakage is below the Technical Specification limit.
- A fuel cladding defect has been detected via multiple indications of rising RMS monitors.
- The severity and magnitude of the defect is being evaluated.
- An inadvertent Containment Isolation signal is then generated during I&C testing.
Which of the following radiation monitors would be expected to continue to have a stable or increasing trend?
- a. 1RE-231, Steam Line A Monitor
- b. 1RE-109, Failed Fuel Monitor
- c. 1RE-116, Demineralizer Valve Gallery Monitor
- d. 1RE-219, Steam Generator Blowdown Liquid Monitor QUESTION: 050 (1.00)
Assume all automatic actions occur.
Which of the following will result in entering EOP-0, Reactor Trip or Safety Injection, and transition to EOP-1.1, SI Termination, without the implementation of any other EOPs?
- a. Unit 1 is at 100% power when a break develops on the RTD Bypass Line, causing containment pressure to rise to 6 psig.
- b. Unit 1 is at 30% power when a Pressurizer PORV sticks open and cannot be isolated, causing Pressurizer pressure to lower to 1700 psig.
- c. Unit 1 is at 100% power when a steam leak develops on the turbine, causing a high-high steam flow condition on both Steam Generators and Steam Generator pressures to lower to 520 psig.
- d. Unit 1 is at 50% power a Steam Generator Safety Valve lifts and sticks open, causing pressure in that Steam Generator to lower to 500 psig.
SENIOR REACTOR OPERATOR Page 47 QUESTION: 051 (1.00)
During recovery from a Unit 1 reactor trip, the crew identifies the following plant conditions at 1200 hour0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />s:
- Steam Generator A pressure is 1075 psig and slowly rising.
- Steam Generator A level is 65% and stable.
At 1205 hours0.0139 days <br />0.335 hours <br />0.00199 weeks <br />4.585025e-4 months <br />, the following plant conditions are noted:
- A Steam Generator pressure begins to lower.
- A Steam Generator level rapidly rises, then slowly lowers.
Which of the following would initiate the conditions noted at 1205 hours0.0139 days <br />0.335 hours <br />0.00199 weeks <br />4.585025e-4 months <br />?
- a. RCS temperature was rapidly lowered.
- b. A Steam Generator feed flow was raised.
- c. A Steam Generator feed flow was lowered.
- d. A Steam Generator Safety Valve has opened.
QUESTION: 052 (1.00)
A small break LOCA has occurred on Unit 1. The crew is performing EOP-1.2, Small Break LOCA Cooldown and Depressurization. Which of the following describes the Reactor Coolant System cooldown rate that is called for in EOP-1.2?
- a. Less than 100°F / hour in order to preclude violating thermal shock limits.
- b. Less than 100°F / hour in order to minimize outsurge from the Pressurizer.
- c. As rapid as possible in order to conserve RWST inventory.
- d. As rapid as possible in order to shorten the time until Residual Heat Removal is placed in service.
SENIOR REACTOR OPERATOR Page 48 QUESTION: 053 (1.00)
Unit 1 is at 15% power and preparing to place the Main Generator on-line. 1P-1A, Reactor Coolant Pump, breaker trips open. Which of the following indicates the effect on the Reactor and Reactor Coolant System?
- a. The reactor will trip automatically.
A loop DT will rise.
- b. The reactor will trip automatically.
A loop DT will lower.
- c. The reactor will not trip.
A loop DT will lower.
- d. The reactor will not trip.
A loop DT will rise.
QUESTION: 054 (1.00)
Unit 1 is operating at 100% power.
- All equipment is in a normal alignment.
- Charging Pump 1P-2A is running in Automatic.
- Charging Pump 1P-2B is in standby.
- Charging Pump 1P-2C is running in Manual.
Charging Pump 1P-2A then trips.
Assuming NO operator action, which of the following correctly describes the change for the given parameters?
- a. Seal injection flow will remain the same, labyrinth seal delta-P will remain the same.
- b. Seal injection flow will remain the same, Charging line flow will lower.
- c. Seal injection flow will lower, Charging line flow will remain the same.
- d. Seal injection flow will lower, labyrinth seal delta-P will lower.
SENIOR REACTOR OPERATOR Page 49 QUESTION: 055 (1.00)
The following conditions exist on Unit 1:
- Pressurizer pressure 1620 psig
- Volume Control Tank pressure 20 psig
- 1P-1A RCP, No. 1 Seal leakage 4 gpm
- 1P-1B RCP, No. 1 Seal leakage 0.5 gpm
- All other seal parameters are normal.
Using the given references, which of the following states the condition of the RCP seal leakage and why it is a concern?
- a. 1P-1A No. 1 seal leakage is high.
The Labyrinth Seal may be damaged by impurities in the RCS.
- b. 1P-1A No.1 seal leakage is high.
The No. 2 Seal does not have enough flow for lubrication.
- c. 1P-1B No. 1 seal leakage is low.
The No. 2 Seal may fail due to higher backpressure from the No. 1 Seal.
- d. 1P-1B seal leakage is low.
The lower pump radial bearing does not have enough cooling flow.
QUESTION: 056 (1.00)
The normal power supply breaker for 1P-10B, Residual Heat Removal Pump, is located on:
- a. 480 Volt Bus B-08
- b. 480 Volt Bus 1B-03
- c. 480 Volt Bus 1B-04
- d. 480 Volt Bus B-09
SENIOR REACTOR OPERATOR Page 50 QUESTION: 057 (1.00)
A small break LOCA has occurred on Unit 1.
- EOP-1.1, SI Termination, is in progress.
- Safety Injection Pump 1P-15A has just been stopped.
- Safety Injection Pump 1P-15B is running.
The following conditions are noted:
- 1TI-970, Subcooling Monitor 200°F and stable
- 1TI-971, Subcooling Monitor 25°F and stable
- RCS Wide Range Pressure 1210 psig
- Core Exit Thermocouple avg 545°F
- Containment pressure 5 psig
- Containment rad levels 4 R/hr After comparing the subcooling readings with RCS pressure and CETs, the crew will determine that:
- a. 1TI-970 is reading accurately, Safety Injection Pump 1P-15A will NOT be started, the crew will continue in EOP-1.1.
- b. 1TI-970 is reading inaccurately, Safety Injection Pump 1P-15A will be started to restore subcooling.
- c. 1TI-971 is reading accurately, Safety Injection Pump 1P-15A will NOT be re-started, the crew will continue in EOP-1.1.
- d. 1TI-971 is reading inaccurately, Safety Injection Pump 1P-15B will be stopped since adequate subcooling exists.
SENIOR REACTOR OPERATOR Page 51 QUESTION: 058 (1.00)
A small break LOCA has occurred on Unit 1. Pressurizer PORVs are being used to reduce RCS pressure per EOP-1.2 Small Break LOCA Cooldown and Depressurization.
- Containment pressure is 10 psig Which of the following is the minimum pressure inside the Pressurizer Relief Tank that will cause the PRT rupture disc to rupture?
- a. 90 psig
- b. 100 psig
- c. 110 psig
- d. 125 psig QUESTION: 059 (1.00)
Given the following plant conditions:
- Unit 2 is operating at 100% power and all equipment is in a normal alignment.
- Annunciator "2T-12 CC Surge Tank Level High or Low" is lit.
- 2LI-618B, CC Surge Tank Level, indicates 58% and rising.
Which of the following actions will help mitigate the consequences of this event?
- a. swap Component Cooling Water Heat Exchangers.
- b. isolate flow to the SI Pump Seal Coolers.
- c. secure Normal Letdown and place Excess Letdown in service.
- d. shut CC-LW-63 and CC-LW-64, Radwaste System CC Supply and Return Valves.
SENIOR REACTOR OPERATOR Page 52 QUESTION: 060 (1.00)
Unit 1 is operating at 100% power when the following sequence of events occurs:
- 1RC-431C, PORV, opens and sticks open.
- 1RC-515, PORV Block valve, cannot be shut.
- Unit 1 Reactor trips.
- Safety Injection actuates.
- PRT pressure rises to the point that the PRT Rupture Disc ruptures.
Which of the following is an effect of the disc rupturing?
- a. N2 Header pressure lowers.
- b. Pressurizer Relief Valve Outlet temperature lowers.
- c. H2 concentration in containment lowers.
- d. PRT level drains below the sparging nozzles.
QUESTION: 061 (1.00)
Unit 1 is operating at 100% power steady-state.
1RC-516, Pressurizer PORV Block valve, has just been shut due to seat leakage past 1RC-430, Pressurizer PORV.
Which of the following would be an indication that 1RC-516 also has seat leakage?
- a. Primary leak rate remains the same.
- b. Pressurizer Vapor Space temperature is rising.
- c. 1RE-211, Containment Particulate Monitor, indication is rising.
- d. Annunciator 1RC-430 or 431C Pressurizer PWR-Operated Relief Valve Not Shut alarms.
SENIOR REACTOR OPERATOR Page 53 QUESTION: 062 (1.00)
Unit 2 was operating at 100% power when a total loss of Main Feedwater occurred. An automatic Reactor Trip signal was generated. However, both Reactor Trip breakers remained shut.
- All attempts to perform a manual Reactor Trip have been unsuccessful.
- An Urgent Failure alarm is preventing all rod motion.
- Auxiliary Feedwater is operating per design.
- Main Generator Output indicates 530 MWe.
- Reactor power remains near 100% on Nuclear Instrumentation.
Which of the following actions will the operator take to mitigate this transient?
- a. Trip the Main Turbine to avoid an excessive Reactor Coolant System pressure increase after the Steam Generator tubes uncover.
- b. Open the Pressurizer PORVs immediately because the increasing Reactor Coolant System pressure will take the Pressurizer solid, resulting in insufficient water relief.
- c. Align maximum Auxiliary Feedwater flow to one Steam Generator to maintain it as a heat sink for cooldown of the Reactor Coolant System.
- d. Reduce turbine load slowly to avoid a rapid Reactor Coolant System temperature and pressure increase, leading to opening of a Pressurizer Safety Valve.
SENIOR REACTOR OPERATOR Page 54 QUESTION: 063 (1.00)
Given the following plant conditions:
- Unit 1 is in Mode 4, Hot Shutdown.
- Both RHR Pumps are in service for decay heat removal and there is a bubble in the Pressurizer.
- I&C is performing maintenance on a failed Containment Pressure instrument.
The following indications are noted:
- "Containment Pressure High" annunciator is lit on 1C04.
- Several Containment Isolation valves which were previously open are now shut.
- All four Emergency Diesel Generators are running (unloaded).
Which of the following procedures will the crew use to address these indications?
- a. EOP-0, Reactor Trip or Safety Injection.
- b. AOP-24, Response To Instrument Malfunctions.
- c. AOP-26, Recovery From Inadvertent Safety Injection.
- d. SEP-1, Degraded RHR System Capability.
QUESTION: 064 (1.00)
Unit 1 is in Mode 3, Hot Standby, and preparing for reactor startup. Three sets of Containment Cooling and Containment Accident Fans are running. The reason that at least three Containment Accident Fans must be running for these conditions is to:
- a. prevent steam formation in the Service Water coils.
- b. maintain Containment pressure below its limit.
- c. support continuous operation of the Reactor Coolant Pumps.
- d. satisfy the LCO for Containment Spray and Cooling Systems.
SENIOR REACTOR OPERATOR Page 55 QUESTION: 065 (1.00)
A steam line break in Unit 1 Containment caused a Containment Spray actuation. All equipment responded as required. Thirty (30) seconds after the spray actuation, what is the status of the Unit 1 Containment Spray System components?
- a. Both spray pumps running.
All four discharge valves open.
Both spray eductor valves shut.
- b. Both spray pumps running.
All four discharge valves open.
Both spray eductor valves open.
- c. Both spray pumps running.
All four discharge valves shut.
Both spray eductor valves shut.
- d. Both spray pumps secured.
All four discharge valves open.
Both spray eductor valves shut.
QUESTION: 066 (1.00)
Unit 1 was operating at 100% power when a steam line break occurred in the PAB. The break is upstream of 1MS-2019, B Steam Generator Supply to 1P-29 TDAFP.
During the performance of EOP-2, Faulted Steam Generator Isolation, the Control Operator fails to manually shut 1MS-2019 as required by procedure.
Which of the following describes a consequence of this error?
- a. 1P-29 will lose its steam supply because both Steam Generators will blow down through the rupture.
- b. 1P-29 will lose its steam supply because 1MS-2082, 1P-29 Low Suction/Overspeed Trip Valve, will trip shut.
- c. 1P-29 will NOT be affected because 1MS-2017, B Steam Generator Main Steam Stop, will automatically shut.
- d. 1P-29 will NOT be affected because 1MS-2019, B Steam Generator Supply to 1P-29 AFP, is a stop-check MOV.
SENIOR REACTOR OPERATOR Page 56 QUESTION: 067 (1.00)
Unit 1 is operating at 100% power. All equipment is in a normal alignment. 1PT-484, Steam Header Pressure Transmitter, fails high. Which of the following describes an effect of this failure?
- a. The Condenser Steam Dumps will NOT open on a loss of turbine load as long as the Steam Dump Mode Selector is in Auto.
- b. The Condenser Steam Dumps will open immediately.
- c. The Condenser Steam Dumps will NOT open on a Reactor Trip as long as the Steam Dump Mode Selector is in Auto.
- d. The Condenser Steam Dumps will open if the Steam Dump Mode Selector is taken to Manual.
QUESTION: 068 (1.00)
OP-13A, Secondary Systems Startup, contains a note to start the Steam Generator Feed Pump (SGFP) oil pumps prior to starting the Condensate Pumps. The reason for this note is to:
- a. electrically enable the Condensate Pumps to start.
- b. minimize the effects of cold seal water on SGFP bearings.
- c. prevent damage to the SGFPs due to condensate flow spinning the pumps.
- d. allow time for the SGFPs oil to warm up to operating temperature prior to starting the SGFPs.
SENIOR REACTOR OPERATOR Page 57 QUESTION: 069 (1.00)
Unit 1 is operating at 100% power, when the following conditions are noted:
- 1P-28A SG Feed Pump Suction Pressure Low alarm.
- 1P-28B SG Feed Pump Suction Pressure Low alarm.
- Both Steam Generator levels are 63% and lowering.
- 1PI-2273, SG Feed Pump Suction Pressure, is 165 psig.
- 1FI-2255, Heater Drain Pumps Discharge Flow, is 5200 gpm.
- 1PI-2272, Condensate Pump Discharge Pressure, is 200 psig.
Which of the following would cause the above conditions?
- a. One Main Feed Pump has tripped.
- b. One Condensate Pump has tripped.
- c. The standby Heater Drain Tank Pump is rotating backwards.
- d. One Heater Drain Tank Pump has tripped.
QUESTION: 070 (1.00)
Unit 1 is operating at 100% power. During troubleshooting of an electrical problem associated with the Auxiliary Feedwater Actuation circuitry, both Motor Driven Auxiliary Feedwater Pumps auto-start and both Unit 1 AFW Discharge MOVs, AF-4023 and AF-4021, open. No other equipment is affected. Unit 1 Main Feedwater flow will:
- a. lower due to Steam Generator swell.
- b. lower due to rising Steam Generator level.
- c. remain constant due to the higher head of the Main Feed pumps.
- d. remain constant due to having a separate Steam Generator feed line.
SENIOR REACTOR OPERATOR Page 58 QUESTION: 071 (1.00)
Unit 1 has been operating at 100% power for several months. Unit 2 has just reached 100%
power, after being shutdown for several weeks for maintenance. Both Units are tripped due to a Circulating Water System malfunction. Decay heat is being removed with the aid of the Atmospheric Steam Dumps and the Auxiliary Feedwater System. How does the amount of AFW required by Unit 1 compare to the amount required by Unit 2, and why?
- a. Unit 1 will require more AFW because decay heat is dependent on power history.
- b. Unit 1 will require more AFW because the metal of its primary system has absorbed more heat.
- c. The units will require equal amounts of AFW because both units started out at the same temperature.
- d. The units will require equal amounts of AFW because both units started out at the same power level.
QUESTION: 072 (1.00)
Given the following plant conditions:
- Unit 1 is at 100% power.
- Turbine Driven Auxiliary Feedwater Pump, 1P-29, is tagged out for maintenance.
- Engineering has just notified the Shift Manager that both Motor Driven Auxiliary Feedwater Pumps, P-38A and P-38B, have several electrical components installed that were calibrated incorrectly.
- The Shift Manager has declared P-38A and P-38B inoperable.
In addition to initiating action to restore an Auxiliary Feedwater Pump to service, which of the following statements describes an action required for Unit 1?
- a. Stable plant conditions should be maintained.
- b. A reactor shutdown to Mode 2 is required.
- c. A reactor shutdown to Mode 3 is required.
- d. A reactor shutdown to Mode 4 is required.
SENIOR REACTOR OPERATOR Page 59 QUESTION: 073 (1.00)
Both Units are operating at 100% power steady-state. The AC Electrical Distribution System is in a normal alignment, except that TS-81, Emergency Diesel Generator G-01 Monthly, is in progress. G-01 is operating in parallel with offsite power through 1X-04 and 1A-05.
The following ammeter indications are noted on C02:
- G-01 Emergency Diesel Generator 300 amps
- 1X-11 Station Service Transformer to 1B01 60 amps
- 1X-13 Station Service Transformer to 1B03 100 amps
- 1X-14 Station Service Transformer to 1B04 80 amps
- 1X-12 Station Service Transformer to 1B02 70 amps What will the G-01 EDG ammeter indicate if 1A52-57, 1A-03 to 1A-05 Bus Tie breaker, trips open? (Note: No other equipment or alignment is affected.)
- a. 100 amps
- b. 180 amps
- c. 300 amps
- d. 400 amps
SENIOR REACTOR OPERATOR Page 60 QUESTION: 074 (1.00)
Both units are operating at 100% power. The DC electrical distribution system is in a normal alignment.
The following conditions are noted:
- D-07 Battery Charger Trouble alarm.
- D-01/D-03 125V DC Bus Under/Over Voltage alarm.
- D-05-AM, Battery Ammeter, indicates 120 amps discharging.
- D-01-VM, Bus Voltmeter, indicates 123 VDC.
The cause of the above indications is:
- a. D-05 Battery Feed to Bus D-01 breaker tripped open.
- b. D-01 DC Bus has a ground.
- c. D-05 Battery has an internal short in a cell.
- d. D-07 Battery Charger DC Output Breaker tripped open.
QUESTION: 075 (1.00)
Both units were operating at 100% power. Three minutes ago, an inadvertent Safety Injection occurred on Unit 1. All equipment responded as required. For these conditions, which of the following will initiate a trip of G-03, Emergency Diesel Generator?
- a. Low Water Pressure.
- b. Low Bearing Oil Pressure.
- c. High Crankcase Pressure.
- d. High Jacket Water Temperature.
SENIOR REACTOR OPERATOR Page 61 QUESTION: 076 (1.00)
Unit 1 has been operating at 100% power for several weeks. 1SC-938C, 1RE-109 Failed Fuel Monitor Flow Valve, which is normally in a throttled position, was just inadvertently taken to a full open position. 1RE-109, Failed Fuel Monitor, indication will:
- a. rise because now it is more sensitive to small fuel defects.
- b. rise because now more N-16 gammas will reach the detector.
- c. remain the same because the gross specific activity of the reactor coolant has not changed.
- d. lower because of the increased flow which lowers the time N-16 gammas are in the detector.
QUESTION: 077 (1.00)
The following plant conditions exist:
- Unit 1 is in Mode 5.
- The RCS is solid at 300 psig and on RHR.
- 1HC-135, Letdown Line Pressure Controller, is in Manual to control RCS pressure.
- All RCS and RHR conditions are stable.
The PAB operator then performs a blowdown of the Service Water side of 1HX-12A and HX-12B, Component Cooling Water Heat Exchangers.
If no other operator actions are taken, RCS pressure will:
- a. lower because RHR temperature will lower.
- c. rise because Service Water blowdown flow will bypass the CCW Heat Exchanger tubes.
- d. lower because the Non-Regenerative Heat Exchanger letdown outlet temperature will rise.
SENIOR REACTOR OPERATOR Page 62 QUESTION: 078 (1.00)
Which of the following is the power supply to Instrument Air Compressor K-2A?
- c. 480 Volt Bus 1B-04.
- d. 480 Volt Bus 2B-04.
QUESTION: 079 (1.00)
Given the following conditions:
- Unit 1 is in Mode 5 with RCS temperature at 100°F.
- RCS time to boil is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- Containment Integrity has been relaxed per OP-3C, Hot Standby To Cold Shutdown.
- Containment purge is in operation per OP-9C, Containment Venting and Purging.
The Unit 1 containment upper personnel airlock has been damaged such that neither door can be shut. Maintenance estimates that it will take 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to return at least one of the doors to service.
Which of the following is a valid concern about the status of the upper airlock?
- a. An unmonitored release to the atmosphere is taking place while both airlock doors are open.
- b. The lower airlock cannot be utilized because one of its bulkhead doors must be locked shut.
- c. The containment closure time requirements of CL-1E, Containment Closure Checklist, are not met.
- d. The Unit 1 upper airlock will not be returned to service before the one hour Technical Specification Required Action time.
SENIOR REACTOR OPERATOR Page 63 QUESTION: 080 (1.00)
Unit 1 is operating at 100% power. A Unit 1 containment inspection is in progress per PC-24, Containment Inspection Checklist. The two Operators and an RP Tech are on the 8 level of containment. The Control Room receives indications of a steam leak inside containment and orders the immediate evacuation of containment.
The three individuals must evacuate via the:
- a. lower personnel hatch because it is the closest exit.
- b. lower personnel hatch because it is the normal entry/exit used during PC-24.
- c. upper personnel hatch because it is the only unlocked exit.
- d. upper personnel hatch because that is where the Designated Airlock Operator is stationed.
QUESTION: 081 (1.00)
Given the following plant conditions:
- Following a series of plant malfunctions, operators are currently implementing ECA-0.0, Loss of All AC Power.
- The operators have reached the point in the procedure where they are to begin depressurization of the Steam Generators.
Which of the following statements indicates the reason that a secondary depressurization is performed?
- a. To ensure the reactor remains subcritical and does not result in a restart accident.
- c. To remove stored energy in the Steam Generators to prevent a secondary side Safety Valve from lifting.
- d. To depressurize the RCS in order to prevent a challenge to the "Integrity" Critical Safety Function Status Tree which is being monitored for implementation.
SENIOR REACTOR OPERATOR Page 64 QUESTION: 082 (1.00)
Given the following plant conditions:
- Unit 1 is at 100% power with all plant equipment in a normal alignment.
- Charging Pump 1P-2A is running in Automatic.
- Charging Pump 1P-2C is running in Manual.
- An electrical fault has resulted in the closure of 1RC-427, Reactor Coolant Loop B Letdown Isolation Valve.
- Attempts to re-open the valve have been unsuccessful.
Which of the following choices is correct regarding the response of the Pressurizer Level Control System to this failure and the actions required to mitigate this event?
- a. 1P-2A Charging pump speed will rise.
1P-2C Charging Pump speed must be lowered manually with 1CV-142 (Charging Line Control Valve) fully opened until Normal Letdown can be re-established.
- b. 1P-2A Charging pump speed will lower.
1P-2C Charging Pump speed must be raised manually with 1CV-142 fully opened until Normal Letdown can be re-established.
- c. 1P-2A Charging pump speed will lower.
Charging flow must be reduced to one pump at minimum speed with 1CV-142 fully closed, and then Excess Letdown placed in service.
- d. 1P-2A Charging pump speed will rise.
1P-2C Charging pump speed must be lowered manually to minimum with 1CV-142 fully closed, and then Excess Letdown placed in service.
QUESTION: 083 (1.00)
Which one of the following inputs cause the rod bottom lights to illuminate?
- a. Bank demand for each control rod bank.
- b. Individual rod position signal via a reed switch.
- c. The Individual Rod Position Indicator (IRPI) signal.
- d. The output signal of the rod control P/A converter.
SENIOR REACTOR OPERATOR Page 65 QUESTION: 084 (1.00)
Unit 1 was operating at 100% power when the following conditions were noted:
- Numerous annunciators are alarming.
- Numerous instruments have failed.
- Unit 1 B Steam Generator level is rising.
- None of the bistable status lights on 1C-04 are lit.
Given the above conditions, which Power Range Nuclear Instrument channel has lost power?
- a. 1N-41
- b. 1N-42
- c. 1N-43
- d. 1N-44 QUESTION: 085 (1.00)
An accident is in progress on Unit 1. The Control Room staff has entered CSP-C.1, Response To Inadequate Core Cooling. Several steps in CSP-C.1 require the value of core exit thermocouples. From what qualified source is this data obtained?
- a. Digital display on 1C-04.
- c. 1TR-00001A and 1TR-00001B on 1C-20.
- d. SPEC-200 racks above the Control Room.
SENIOR REACTOR OPERATOR Page 66 QUESTION: 086 (1.00)
Given the following plant condition:
- A seismic event has occurred that has resulted in a non-isolable leak at the North end of the Spent Fuel Pool.
- The leak is located one foot above the top of the spent fuel racks.
- Level in the pool is slowly lowering.
- The fuel transfer canal doors are open.
What will be the effect of this leak on the Spent Fuel Pool Cooling System and what action will the crew take to mitigate this event?
- a. Cooling will NOT be lost since the leak is above the suction pipe opening, the transfer canal doors are required to be shut to conserve water inventory.
- b. Cooling will NOT be lost since the leak is above the suction pipe opening, makeup to the pool is required to be initiated to provide radiation shielding above the fuel.
- c. Cooling will be lost when level drops below the suction pipe opening, pool cooling will be initiated by recirculating water between the transfer canal and the Spent Fuel Pool with P-9, HUT Recirc Pump.
- d. Cooling will be lost when level drops below the suction pipe opening, make-up to the pool is required to be initiated to control pool temperature and maintain inventory.
SENIOR REACTOR OPERATOR Page 67 QUESTION: 087 (1.00)
Unit 1 is holding reactor power at 28%. The following conditions are noted:
- Main Generator MWe = 150 MWe and stable.
- 1PT-2058, Impulse Pressure, indicates 100%.
- The LOAD REF CHAN monitor light is lit.
Based on the above conditions, the Control Operator must ensure:
- a. the VPL is limiting governor valve position.
- b. EH controls are in Turbine Manual.
- c. the Control Rod Bank Selector switch is in Manual.
- d. the Condenser Steam Dump controller is in Manual.
QUESTION: 088 (1.00)
A discharge of the "A" Monitor Tank is in progress. Which of the following would provide the Control Operator indication that WL-18, Waste Condensate Overboard to SW Header Control Valve, has automatically closed?
- a. The "Unit 1 Process Monitor High" annunciator alarms.
- b. RE-223, Waste Distillate Tank Overboard Monitor, status indication on the RMS server changes from green to blue.
- c. The status light for WL-18 is lit on the Containment Isolation Panel.
- d. RE-218, Waste Disposal System Liquid Monitor, status indication on the RMS server changes from green to red.
SENIOR REACTOR OPERATOR Page 68 QUESTION: 089 (1.00)
Given the following plant conditions:
- Unit 1 is at 100% power.
- The "A" Gas Decay Tank is being discharged per OP-9D, Discharge of Gas Decay Tanks.
- A forced vent of Unit 1 containment is in progress per OP-9C, Containment Venting and Purging.
The following alarms are then received:
- "Containment or Aux Bldg Vent System Air Flow Low" annunciator on 1C04.
- FT-3298A, PAB Flow Stack Velocity, alarms on the RMS System Server and indicates low.
Which of the following actions is required for these conditions?
- a. Secure the Gas Decay Tank discharge.
- b. Secure the Unit 1 containment forced vent.
- c. Start the standby Cavity Cooling Fan.
- d. Start the standby Purge Exhaust Fan.
QUESTION: 090 (1.00)
P-35A, Electric Fire Pump, is OOS. PS-3713, Diesel Fire Pump Start Pressure Switch, is inadvertantly isolated. In the event of a fire, which of the following will NOT be protected by an automatic suppression system?
- a. Vital Switchgear Room.
- b. G-03, Emergency Diesel Generator.
- c. Auxiliary Feed Pump room.
- d. Main Turbine bearings.
SENIOR REACTOR OPERATOR Page 69 QUESTION: 091 (1.00)
Given the following plant conditions:
- Unit 1 is at 100% power.
- Bank D step counter indicates 220 steps.
- Unit 1 containment forced vent in progress.
Which of the following control board indications will require Unit 1 to enter a Technical Specification Action Condition?
- a. 1LI-428, Pressurizer Level, indicates 40%.
- b. Control Rod C-7 (Bank D) IRPI indicates 205 steps.
- c. 1PI-449, Pressurizer Pressure, indicates 2215 psig.
- d. 1PI-945, Containment Pressure, indicates -0.4 psig.
SENIOR REACTOR OPERATOR Page 70 QUESTION: 092 (1.00)
You are the Unit 1 Control Operator. Unit 1 has had a steam line break. The procedure currently in use is EOP-0, Reactor Trip or Safety Injection. The DOS reads the following step to you:
Check If Secondary System Is Intact:
- No S/G pressure dropping in an uncontrolled manner.
AND
- No S/G completely depressurized.
The following conditions are noted:
A S/G pressure 400 psig and lowering B S/G pressure 800 psig and stable What reply is required to be given to the DOS?
- a. "Yes, the secondary system is not intact."
- b. "Yes, the Bravo Steam Generator is intact, but no, the Alpha Steam Generator is not."
- c. "No, Alpha Steam Generator pressure is 400 psig and lowering, Bravo Steam Generator pressure is 800 psig and stable."
- d. "No, Alpha Steam Generator is faulted, Bravo Steam Generator pressure is 800 psig."
SENIOR REACTOR OPERATOR Page 71 QUESTION: 093 (1.00)
Unit 2 has been operating at 50% power for several days due to 2P-28A, Main Feedwater Pump, being OOS for maintenance. A severe plant transient occurs. The result is several automatic trip signals being generated without the reactor trip breakers opening; however, a manual trip is successfully performed. After stabilizing the plant, a Post Trip Review indicated the following simultaneous panel readings occurred during the transient:
- RCS pressure 2400 psig
- Reactor power 52%
- RCS TAVG 640°F
- RCPs Both running Using the given references, which of the following statements is correct?
- a. No safety limits were exceeded.
- b. Only the Reactor Core Safety Limit was exceeded.
- c. Only the RCS Pressure Safety Limit was exceeded.
- d. Both Reactor Core and the RCS Pressure Safety Limits were exceeded.
QUESTION: 094 (1.00)
Given the following plant conditions:
- Unit 1 was operating at 100% power when condenser vacuum was observed to be lowering.
- Power was rapidly reduced and is currently 65%.
- Rod control is in AUTO.
- Control Bank 'D' is observed to be stepping in at minimum speed (8 steps/min) due to a TAVG-TREF deviation of +2.0°F and is currently at 96 steps.
Which of the following is correct with regard to control rod operation for these conditions?
- a. The rod speed proportional controller is malfunctioning.
- b. Control Bank 'D' rods should be stepping out.
- c. Control Bank 'C' and 'D' rods should be stepping in.
- d. The rod control system is operating properly.
SENIOR REACTOR OPERATOR Page 72 QUESTION: 095 (1.00)
Which of the following is a description of how a Rod Control Cluster Assembly (RCCA) is changed in a fuel assembly located inside containment?
- a. The manipulator removes the RCCA from the spent fuel assembly and places it in the basket of the change fixture, the change fixture gripper picks up the RCCA and places it in the receiving fuel assembly in the upender.
- b. The manipulator places a spent fuel assembly with an RCCA in one basket on the change fixture, the change fixture gripper picks up the RCCA from the assembly and places it in the receiving fuel assembly in the upender.
- c. The manipulator places a spent fuel assembly with an RCCA in one of the baskets on the change fixture, then places the receiving fuel assembly in the basket next to the spent fuel assembly, the change fixture gripper then picks up the RCCA and shifts it to the receiving fuel assembly.
- d. The manipulator places a spent fuel assembly with an RCCA in one of the baskets on the change fixture. The manipulator operator then removes the RCCA from the assembly with the manipulator and places it in the receiving fuel assembly in the core.
QUESTION: 096 (1.00)
Operations and RP have just completed filling the spent resin High Integrity Container (HIC) with spent resin. The results of a subsequent radiation survey is as follows:
Contact @ 30 cm
- Top of shielded HIC by fill head 2500 mr/hr 1200 mr/hr
- Sides of HIC 100 mr/hr 60 mr/hr Which of the following describes the required radiological postings?
- a. The HIC must be posted as a High Radiation Area with a red flashing light.
- b. The HIC must be posted as a High Radiation Area without a red flashing light.
- c. No postings are required.
- d. The PAB truck bay must be posted as a Very High Radiation Area.
SENIOR REACTOR OPERATOR Page 73 QUESTION: 097 (1.00)
Per NP 4.2.14, Administrative Dose Levels/Dose Level Extension Procedure, an individual at Point Beach has an administrative dose limit of (1) ______ mrem TEDE per year. This can be raised to (2) ______ mrem TEDE per year by the First-line Supervisor.
- a. (1) 1000 (2) 3000
- b. (1) 1000 (2) 4000
- c. (1) 2000 (2) 3000
- d. (1) 2000 (2) 5000 QUESTION: 098 (1.00)
Which one of the following pieces of equipment, if operated, would be considered a monitored release path?
- a. 1MS-2037, Steam Supply to the Priming Air Ejector.
- b. 1P-29, Turbine Driven Auxiliary Feedwater Pump.
- c. 1MS-2016, Atmospheric Steam Dump.
- d. 1MS-2050, Condenser Steam Dump.
QUESTION: 099 (1.00)
A transient has occurred on Unit 1 causing multiple alarms. The unit has remained on-line. The Unit 1 CO has just used the "Mushroom" to silence alarms. Which one of the following individuals is required to be notified that the "Mushroom" has been used?
- a. OS
- b. DOS
- c. Shift Manager
- d. Unit 2 CO
SENIOR REACTOR OPERATOR Page 74 QUESTION: 100 (1.00)
Given the following Unit 1 plant conditions:
- OP-3C, Hot Standby to Cold Shutdown, is in progress.
- 1P-10A, RHR Pump, is running.
- Preparations are being made to start 1P-10B, RHR Pump.
- RCS temperature is 250°F and slowly lowering.
- 1P-10A suddenly trips due to motor failure.
- An attempt to start 1P-10B is made, however, its breaker will NOT close.
- Subsequent attempts to start an RHR pump have failed.
- The procedure currently in effect is SEP-1.1, Alternate Core Cooling.
Which of the following methods available in SEP-1.1 would NOT provide for decay heat removal for these conditions?
- a. Steaming via a Steam Generator (SG) and utilizing Auxiliary Feedwater as makeup to the SG.
- b. Feeding with a Safety Injection Pump and opening a Pressurizer PORV.
(********** END OF EXAMINATION **********)
SENIOR REACTOR OPERATOR Page 75 ANSWER: 001 (1.00) ANSWER: 005 (1.00) ANSWER: 009 (1.00)
- c. b. b.
REFERENCE:
REFERENCE:
REFERENCE:
DCS Handbook 2.1.1 ECA-0.0, Loss Of All AC Logic Sheet 18, Pressurizer Appendix A Power PBNP Electrical Power Pressure and Level Control FUNDAMENTAL Distribution diagram AOP-1D, Chemical and NEW HIGH Volume Control System 2.4.30 ..(KAs) NEW Malfunction, purpose 055.EA2.03 ..(KA's) AOP-24, Response to Instrument Malfunctions, ANSWER: 002 (1.00) purpose
- d. ANSWER: 006 (1.00) HIGH
REFERENCE:
- c. NEW T.S. 3.4.5 and bases
REFERENCE:
028.AA2.12 ..(KA's)
HIGH INPO Bank 19554, Cook 1, NEW ExamDate 05/21/2001 2.1.12 ..(KAs) EOP-0, Reactor Trip Or ANSWER: 010 (1.00)
Safety Injection c.
HIGH
REFERENCE:
ANSWER: 003 (1.00) BANK EPIP 1.2, Emergency
- a. Classification, EAL 4.1.1.1
REFERENCE:
FUNDAMENTAL Logic Sheet 8, Logic Diagram ANSWER: 007 (1.00) NEW Safeguard Sequence Logic d. 060.AK3.01 ..(KA's)
Sheet 9, Safeguards
REFERENCE:
Sequence Logic T.S. 3.4.9, INPO Bank 20674, PBNP 1, Pressurizer ExamDate 02/02/2002 ANSWER: 011 (1.00)
HIGH EOP-1.3, Transfer To c.
NEW Containment Sump
REFERENCE:
027.AA2.10 ..(KA's) Recirculation ECA-1.1, Loss INPO Bank 20596, PBNP 1, Of Containment Sump ExamDate 02/02/2002 OM Recirculation 3.7, AOP and EOP ANSWER: 004 (1.00) HIGH Procedure Sets Use and
- d. BANK Adherence
REFERENCE:
HIGH EOP-3, Steam Generator MODIFIED Tube Rupture, Background ANSWER: 008 (1.00)
INPO Bank 2858 b.
FUNDAMENTAL
REFERENCE:
ANSWER: 012 (1.00)
BANK PBNP ITS Bank a.
038.EK3.09 ..(KA's) 057.02.LP3338.003.009 T.S.
REFERENCE:
3.2.4, Quadrant Power Tilt BG-EOP-0.2, Natural Ratio (QPTR) Circulation Cooldown, R20, HIGH page 37 MODIFIED FUNDAMENTAL 005.AK1.02 ..(KA's) NEW
SENIOR REACTOR OPERATOR Page 76 ANSWER: 013 (1.00) ANSWER: 017 (1.00) ANSWER: 020 (1.00)
- d. a. c.
REFERENCE:
REFERENCE:
REFERENCE:
WEST 684J741 Sh. 2, 0-SOP-IC-002, Tech Spec Bank Chemical & Volume Control LCO - Instrument Cross LOR2002ExamQuestions P&ID AOP-1D, Chemical and Reference T.S. 3.3.1, RPS 024.00.LP3694.001.001 Volume Control System Instrumentation T.S. 3.3.2, OM-1.1, Conduct of Plant Malfunctions ESFAS Instrumentation Operations HIGH WEST 883D195 Sh. 18, HIGH NEW Pressurizer Pressure and BANK 004.A1.11 ..(KAs) Level Control logic 2.1.9 ..(KAs)
HIGH NEW ANSWER: 014 (1.00) 011.A1.04 ..(KAs) ANSWER: 021 (1.00)
- b. a.
REFERENCE:
REFERENCE:
Bank - ANSWER: 018 (1.00) T.S. 3.0, LCO Applicability TRCR31_0BNK.LXRBANK, b. Bank, 031.01.LP0159.009.001 T.S.
REFERENCE:
LOR2002ExamQuestions.LX Bases 3.5.2 ECCS - INPO 20682, PBNP, RBANK, Operating ExamDate 02/02/2002 T.S. 057.02.LP3341.002.002 T.S.
FUNDAMENTAL Bases 3.7.4, ADV Flowpaths 3.3.3, PAM Instrumentation BANK FUNDAMENTAL T.S. 5.6, Reporting 2.2.25 ..(KAs) BANK Requirements 041.2.1.10 ..(KA's) HIGH BANK ANSWER: 015 (1.00) 2.2.22 ..(KAs)
- d. ANSWER: 019 (1.00)
REFERENCE:
d.
EOP-1.3, Transfer To
REFERENCE:
ANSWER: 022 (1.00)
Containment Sump INPO 20679, PBNP, c.
Recirculation, step 13 ExamDate 02/02/2002
REFERENCE:
BG-EOP-1.3 OM-3.1, Operations Shift Bank, Kewaunee 2002 Exam, HIGH Staffing Requirements #91 OI-35A, Standby NEW FUNDAMENTAL Emergency Power Alignment 008.A2.02 ..(KAs) BANK FUNDAMENTAL 2.1.5 ..(KAs) MODIFIED 2.2.24 ..(KAs)
ANSWER: 016 (1.00) d.
REFERENCE:
ANSWER: 023 (1.00)
INPO Bank 20611, PBNP, a.
ExamDate 02/02/2002 T.S.
REFERENCE:
Bases 3.6.6, Containment RECM Section 3.0 Spray and Cooling Systems HIGH HIGH MODIFIED MODIFIED 2.3.9 ..(KAs) 026.A1.01 ..(KAs)
SENIOR REACTOR OPERATOR Page 77 ANSWER: 024 (1.00) ANSWER: 028 (1.00) ANSWER: 033 (1.00)
- c. d. b.
REFERENCE:
REFERENCE:
REFERENCE:
INPO 20685, PBNP, INPO bank 2866 PBNP exam None Provided 02/02/2002 OM 3.7, AOP and 8/2/1999 EOP-1.2 foldout HIGH EOP Procedure Sets Use page BANK and Adherence HIGH 026.AK3.02 ..(KA's)
FUNDAMENTAL BANK BANK 009.EA2.34 ..(KA's) 2.4.14 ..(KAs) ANSWER: 034 (1.00) d.
ANSWER: 029 (1.00)
REFERENCE:
ANSWER: 025 (1.00) c. TRHB 10.3, Pressurizer,
- a.
REFERENCE:
Pressure Control and Relief
REFERENCE:
EOP-1.3 and BG Document, System WEST 883D195 OM 3.7, AOP and EOP LP3340 pg 13 Sh.18, Pressurizer Pressure Procedure Sets Use and FUNDAMENTAL and Level Control Logics Adherence, section 4.8 BANK HIGH FUNDAMENTAL 011.EK2.02 ..(KA's) NEW NEW 2.4.16 ..(KAs) 027.AK1.01 ..(KA's)
ANSWER: 030 (1.00)
ANSWER: 026 (1.00) c. ANSWER: 035 (1.00)
- a.
REFERENCE:
a.
REFERENCE:
REFERENCE:
EOP-0, Reactor Trip or HIGH EOP-3 Safety Injection Bank: NRC NEW FUNDAMENTAL Y2K Bnk 015.AK3.02 ..(KA's) NEW 031.02.LP0405.006.002 038.EA2.17 ..(KA's)
PBNP Electrical Bus Diagram ARB C02 D 3-4, 4.16kV Bus ANSWER: 031 (1.00)
Lockout a. ANSWER: 036 (1.00)
FUNDAMENTAL
REFERENCE:
c.
BANK AOP-1D
REFERENCE:
007.2.4.31 ..(KA's) FUNDAMENTAL EOP-0, Attachment A, NEW Automatic Action Verification 022.AK3.02 ..(KA's) HIGH ANSWER: 027 (1.00) NEW
- a. 040.AA2.05 ..(KA's)
REFERENCE:
ANSWER: 032 (1.00)
LP1829, EOP Generic Issues b.
FUNDAMENTAL
REFERENCE:
ANSWER: 037 (1.00)
BANK None Provided d.
008.AK3.03 ..(KA's) HIGH
REFERENCE:
BANK NRC Y2K BNK.LXRBANK 025.AK1.01 ..(KA's) WEST 883D195 sh. 10 HIGH MODIFIED 054.AA2.05 ..(KA's)
SENIOR REACTOR OPERATOR Page 78 ANSWER: 038 (1.00) ANSWER: 042 (1.00) ANSWER: 046 (1.00)
- c. a. a.
REFERENCE:
REFERENCE:
REFERENCE:
AOP-18A, Train "A" ECA-1.1, Loss Of NRC Y2K BANK.LXRBank Equipment Operation Containment Sump 055.00.LP0000.000 005 HIGH Recirculation, ARB C01 B EOP-0.1 / OP 5B NEW 3-9, STPT 11.1 FUNDAMENTAL 056.AA1.31 ..(KAs) HIGH BANK NEW 024.AK2.04 ..(KA's)
W/E11.EK1.3 ..(KAs)
ANSWER: 039 (1.00)
- b. ANSWER: 043 (1.00) ANSWER: 047 (1.00)
REFERENCE:
- a. d.
REFERENCE:
REFERENCE:
3-5, AOP-9A LP0169, Auxiliary Feedwater OP-1B, Reactor Startup HIGH System HIGH NEW HIGH BANK 2.4.31 ..(KAs) NEW 033.2.1.33 ..(KA's)
W/E05.EA1.1 ..(KAs)
ANSWER: 040 (1.00) ANSWER: 048 (1.00)
- c. ANSWER: 044 (1.00) c.
REFERENCE:
- d.
REFERENCE:
AOP-5B Rev 22, NOTE
REFERENCE:
TRCR43_0BNK.LXRBank before and step 24 NRC Y2K Bnk.LXRBank 043.03.LP2000.006 003 FUNDAMENTAL 055.00.LP0000.000 003 CSP-Z.1 / BG-EOP-2 NEW AOP-6C Symptoms or Entry FUNDAMENTAL 065.AK3.03 ..(KA's) Conditions BANK HIGH 069.AK3.01 ..(KA's)
BANK ANSWER: 041 (1.00) 001.AK1.08 ..(KA's)
- b. ANSWER: 049 (1.00)
REFERENCE:
a.
ECA-1.2, LOCA Outside ANSWER: 045 (1.00)
REFERENCE:
Containment c. WEST 883D195 sh 21 HIGH
REFERENCE:
HIGH NEW INPO bank #2691(S), NEW W/E04.EK1.3 ..(KAs) #9134(R) EOP-0.1 and 076.AA1.04 ..(KA's)
BG-EOP-0.1 HIGH MODIFIED 005.AA2.03 ..(KA's)
SENIOR REACTOR OPERATOR Page 79 ANSWER: 050 (1.00) ANSWER: 054 (1.00) ANSWER: 058 (1.00)
- c. d. c.
REFERENCE:
REFERENCE:
REFERENCE:
EOP-0, Reactor Trip or WEST 685J175 Sh. 2, LP0078, Pressurizer, Level Safety Injection INPO 17388, Chemical & Volume Control Control, Pressure Control, Salem, ExamDate P&ID and Relief System 01/23/1998 WEST 883D195 HIGH HIGH Sh. 7, Safeguards Actuation NEW NEW Signals Logic 003.A3.01 ..(KAs) 007.K1.01 ..(KAs)
HIGH BANK ANSWER: 055 (1.00) ANSWER: 059 (1.00)
- d. c.
ANSWER: 051 (1.00)
REFERENCE:
REFERENCE:
REFERENCE:
Pump Malfunction, Figure 1 HIGH LP0446, Steam Generator OP-4B, Reactor Coolant NEW Transient Response Pump Operation 008.A1.04 ..(KAs)
HIGH HIGH NEW NEW W/E13.EA1.2 ..(KAs) 004.K6.31 ..(KAs) ANSWER: 060 (1.00) b.
ANSWER: 052 (1.00)
REFERENCE:
- a. ANSWER: 056 (1.00) LP0413, Second Law of
REFERENCE:
REFERENCE:
HIGH Cooldown and OI-112, Aligning Equipment NEW Depressurization LP3339, to Appendix R Power Supply 010.K6.04 ..(KAs)
Reactor Coolant System. LO FUNDAMENTAL 057.02.LP3339.001 LP 1829 NEW EOP Generic issues L.O 050.K2.01 ..(KAs) ANSWER: 061 (1.00) 031.02.LP1829.005 a.
FUNDAMENTAL
REFERENCE:
NEW ANSWER: 057 (1.00) P&ID 541F091 Sh. 2, Reactor W/E03.EA2.2 ..(KAs) b. Coolant System
REFERENCE:
FUNDAMENTAL OP 4B, EOP-1.1, SI NEW ANSWER: 053 (1.00) Termination Steam Tables 010.A4.03 ..(KAs)
- c. HIGH
REFERENCE:
NEW WEST 883D195 Sh. 15, RCS 006.K6.18 ..(KAs)
Trip Signals WEST 883D195 Sh. 12, Nuclear Instr.
Permissives & Blocks STPT 3.1, Reactor Trip Interlock Setpoints LP0408 HIGH NEW 003.K3.01 ..(KAs)
SENIOR REACTOR OPERATOR Page 80 ANSWER: 062 (1.00) ANSWER: 066 (1.00) ANSWER: 070 (1.00)
- a. d. b.
REFERENCE:
REFERENCE:
REFERENCE:
INPO 20653, PBNP, P&ID M-201 Sh. 1, Main & LP0131, Feedwater Control ExamDate 02/02/2002 Reheat Steam System TRHB System CSP-S.1, Response to 11.1, Secondary Systems HIGH Nuclear Power Descriptions: Main Steam NEW Generation/ATWS HIGH 059.K1.02 ..(KAs)
FUNDAMENTAL NEW BANK 039.K1.07 ..(KAs) 012.A2.06 ..(KAs) ANSWER: 071 (1.00) a.
ANSWER: 067 (1.00)
REFERENCE:
ANSWER: 063 (1.00) d. LP0332, Fundamentals of
- c.
REFERENCE:
Nuclear Physics (Part 2)
REFERENCE:
WEST 883D195 Sh. 17, FUNDAMENTAL AOP-26 Steam Dump Control Logic NEW HIGH Diagram 061.K5.02 ..(KAs)
BANK HIGH 2.4.4 ..(KAs) NEW 039.K3.06 ..(KAs) ANSWER: 072 (1.00) a.
ANSWER: 064 (1.00)
REFERENCE:
- c. ANSWER: 068 (1.00) LP3343, TS 3.7.5
REFERENCE:
- c. FUNDAMENTAL OI 72, Containment Air
REFERENCE:
NEW Recirculation System LP0102, Condensate System 2.2.22 ..(KAs)
FUNDAMENTAL FUNDAMENTAL NEW NEW 022.2.1.27 ..(KA's) 056.K1.03 ..(KAs) ANSWER: 073 (1.00) a.
REFERENCE:
ANSWER: 065 (1.00) ANSWER: 069 (1.00) LP0007, 4160 VAC Electrical
- a. b. Distribution
REFERENCE:
REFERENCE:
HIGH WEST 883D195 Sh. 8, M-202 Sh.1, Condensate NEW Safeguard Sequence Logic System, TRHB 11.2, TRHB 062.A3.01 ..(KAs)
Diagram WEST 883D195 Sh. 11.6 9, Safeguards Sequence HIGH Logic NEW ANSWER: 074 (1.00)
FUNDAMENTAL 056.A2.04 ..(KAs) d.
NEW
REFERENCE:
026.A3.01 ..(KAs) ARP 2C20A 1-1 ARP 2C20A 2-2 0-SOP-DC-001 HIGH NEW 063.A3.01 ..(KAs)
SENIOR REACTOR OPERATOR Page 81 ANSWER: 075 (1.00) ANSWER: 080 (1.00) ANSWER: 085 (1.00)
- b. a. c.
REFERENCE:
REFERENCE:
REFERENCE:
Modified from PC-24, Containment CSP-C.1, Response to TRCR52_BNK.LXRBANK Inspection Checklist Inadequate Core Cooling 054.02.LP0133.004.001 FUNDAMENTAL FUNDAMENTAL TRHB 12.8 NEW NEW FUNDAMENTAL 103.A2.04 ..(KAs) 016.A4.02 ..(KAs)
BANK 064.K4.02 ..(KAs)
ANSWER: 081 (1.00) ANSWER: 086 (1.00)
- b. d.
ANSWER: 076 (1.00)
REFERENCE:
REFERENCE:
- b. ECA-0.0, LP0462. AOP-8F, LP0110 pg 19 L.O
REFERENCE:
FUNDAMENTAL 112.01.LP0110.006 RMSASRB CI 1RE-109 BANK HIGH HIGH 2.4.6 ..(KAs) NEW NEW 033.A2.03 ..(KAs) 073.K5.01 ..(KAs)
ANSWER: 082 (1.00)
ANSWER: 077 (1.00) c. ANSWER: 087 (1.00)
- a.
REFERENCE:
b.
REFERENCE:
REFERENCE:
1-SOP-CC-001 HIGH LP0023, Main Turbine HIGH NEW Controls NEW 011.A2.07 ..(KAs) HIGH 076.K1.08 ..(KAs) NEW 045.A2.17 ..(KAs)
ANSWER: 083 (1.00)
ANSWER: 078 (1.00) c.
- a.
REFERENCE:
ANSWER: 088 (1.00)
REFERENCE:
LP0576, Rod Position d.
LP0338, Instrument and Indication System OP-1B,
REFERENCE:
Service Air, PBNP Electrical Reactor Startup RMSASRB CI RE-218, Dist. Drawing, WEST FUNDAMENTAL Waste Disposal System 883D195 sh9. NEW Liquid Monitor RMSASRB FUNDAMENTAL 014.K4.03 ..(KAs) 1.0, Generic RMS Alarm NEW Response Guidelines 078.K2.01 ..(KAs) FUNDAMENTAL ANSWER: 084 (1.00) NEW
- d. 068.A4.04 ..(KAs)
ANSWER: 079 (1.00)
REFERENCE:
- c. LP3456, DC and Instrument
REFERENCE:
Bus Malfunctions LP2416, CL-1E, Containment Closure Nuclear Instrumentation Checklist System HIGH HIGH NEW NEW 103.K3.01 ..(KAs) 015.K2.01 ..(KAs)
SENIOR REACTOR OPERATOR Page 82 ANSWER: 089 (1.00) ANSWER: 093 (1.00) ANSWER: 097 (1.00)
- a. a. c.
REFERENCE:
REFERENCE:
ARB 1C04 1C 2-9 RMS ARB T.S. 2.0, Safety Limits INPO
REFERENCE:
FT-3298A OI-39 pg 10 20655, PBNP, 02/02/2002 Bank question 19331 from HIGH COLR, Core Operating Limits INPO bank NP 4.2.14, NEW Report Administrative Dose 071.A1.06 ..(KAs) HIGH Levels/Dose Level Extension BANK Procedure ANSWER: 090 (1.00) 2.2.22 ..(KAs) FUNDAMENTAL
- b. BANK
REFERENCE:
2.3.4 ..(KAs)
LP0003, Fire Protection ANSWER: 094 (1.00)
FUNDAMENTAL c.
NEW
REFERENCE:
ANSWER: 098 (1.00) 086.A2.04 ..(KAs) Modified question from KNPP d.
2002 exam. LP1547, Rod
REFERENCE:
Control System. LP0035 LP0135 ANSWER: 091 (1.00) HIGH FUNDAMENTAL
- b. MODIFIED NEW
REFERENCE:
T.S. 3.1.4, Rod Group Alignment Limits FUNDAMENTAL ANSWER: 095 (1.00) ANSWER: 099 (1.00)
NEW c. d.
2.1.33 ..(KAs)
REFERENCE:
REFERENCE:
TRCR112bnk.LXRBank OM 1.1 112.01.LP0259.002 006 FUNDAMENTAL ANSWER: 092 (1.00) LP0259, Fuel Handling NEW
- c. Containment 2.4.31 ..(KAs)
REFERENCE:
FUNDAMENTAL OM 3.7, AOP and EOP BANK Procedure Sets Use And 2.2.27 ..(KAs) ANSWER: 100 (1.00)
Adherence d.
FUNDAMENTAL
REFERENCE:
NEW ANSWER: 096 (1.00) 2.1.17 ..(KAs) a. SEP-3, Loss of All AC Power
REFERENCE:
to a Shutdown Unit, OP-3C, INPO 20657, PBNP, Hot Standby To Cold ExamDate 02/02/2002 HP Shutdown 3.2, Radiological Labeling, HIGH Posting and Barricading NEW Requirements 2.4.9 ..(KAs)
HIGH BANK 2.3.1 ..(KAs)
(********** END OF EXAMINATION **********)
SENIOR REACTOR OPERATOR Page 83 ANSWER KEY 001 c 021 a 041 b 061 a 081 b 002 d 022 c 042 a 062 a 082 c 003 a 023 a 043 a 063 c 083 c 004 d 024 c 044 d 064 c 084 d 005 b 025 a 045 c 065 a 085 c 006 c 026 a 046 a 066 d 086 d 007 d 027 a 047 d 067 d 087 b 008 b 028 d 048 c 068 c 088 d 009 b 029 c 049 a 069 b 089 a 010 c 030 c 050 c 070 b 090 b 011 c 031 a 051 d 071 a 091 b 012 a 032 b 052 a 072 a 092 c 013 d 033 b 053 c 073 a 093 a 014 b 034 d 054 d 074 d 094 c 015 d 035 a 055 d 075 b 095 c 016 d 036 c 056 c 076 b 096 a 017 a 037 d 057 b 077 a 097 c 018 b 038 c 058 c 078 a 098 d 019 d 039 b 059 c 079 c 099 d 020 c 040 c 060 b 080 a 100 d
(********** END OF EXAMINATION **********)