ML032480039

Initial SRO Examination 08/2003

ML032480039
Person / Time
Site:	Palisades
Issue date:	08/01/2003
From:	NRC/RGN-III
To:	Nuclear Management Co
References
50-255/03-301
Download: ML032480039 (85)

Text

U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: MASTER SRO Region: III Date: August 1, 2003 Facility/Unit: PALISADES License Level: SRO Reactor Type: CE Start Time: 0810 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 __________ Points Applicants Score __________ Points Applicants Grade __________ Percent

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.

PART B - WRITTEN EXAMINATION GUIDELINES

1. [Read Verbatim] 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.

2. 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.

3. For an initial examination, the nominal time limit for completing the examination is six hours; extensions will be considered under extenuating circumstances.

4. You may bring pens, pencils, and calculators into the examination room. Use black ink to ensure legible copies; dark pencil should be used only if necessary to facilitate machine grading.

5. Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identification.

6. Mark your answers on the answer sheet provided and do not leave any question blank.

Use only the paper provided and do not write on the back side of the pages. If you are using ink and decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change.

7. 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.

8. 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.

9. When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC 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. The scrap paper 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 of 71 QUESTION: 001 (1.00)

To ensure that in the event of a Large Break LOCA the Low Pressure Safety Injection Pump P-67A bearings and seals will be cooled adequately, which of the following is true for the valves listed below?

- CV-0913, CCW Inlet to Safeguards

- CV-0950, CCW Outlet from Safeguards These valves are normally maintained ...

a. CLOSED to prevent seal leakage when pump is idle, but automatically open on a SIAS.

b. CLOSED to reduce the potential for draining CCW to the lake, but automatically open on a SIAS.

c. OPEN due to single failure criteria concerns, even though they receive an open signal on a SIAS.

d. OPEN because they do NOT receive an open signal on a SIAS.

QUESTION: 002 (1.00)

Which of the following describes the design basis of the ATWS circuitry and its interface with the rod position indicating system?

a. Circuitry is designed to provide a completely independent trip from the RPS, therefore the Plant Process Computer rod position indication must be used, since ATWS actuation does not affect LED display on panel C-02.

b. Circuitry is designed to provide a completely independent trip from the RPS.

ATWS actuation will cause LED display on panel C-02 to change from 131" to 0".

c. Circuitry design does not require it to provide a completely independent trip from the RPS. ATWS actuation affects rod position indication exactly the same as an RPS trip.

d. Circuitry is designed to provide a completely independent trip from the RPS.

ATWS actuation will cause Plant Process Computer rod position indication to change from 0" to 131".

SENIOR REACTOR OPERATOR Page 5 of 71 QUESTION: 003 (1.00)

Given the following conditions:

- The plant is at full power.

- There are TWO licensed operators in the Control Room, one NCO, and the Control Room Supervisor.

- The Turbine NCO is in the restroom.

- The Shift Engineer is in the Tech. Support Center.

- The Main Turbine and Generator spuriously trip.

- The Reactor does NOT automatically trip.

- The Reactor NCO has a seizure and is rendered unable to function as a licensed operator.

As the Control Room Supervisor, what is your required action, and what procedure specifies this action?

a. Contact the Turbine NCO to manually trip the Reactor, as required by Admin Proc. 4.00, "Operations Organization, Responsibilities, and Conduct".

b. You must manually trip the Reactor, as required by Admin Proc. 4.14, "Conduct of Operations".

c. Contact the Shift Engineer to call out the EMTs, and then manually trip the Reactor, as required by Admin Proc. 4.14, "Conduct of Operations"

d. You must manually trip the Reactor, as required by Admin Proc. 4.02, "Control of Equipment."

SENIOR REACTOR OPERATOR Page 6 of 71 QUESTION: 004 (1.00)

A Station Blackout has occurred. As the Control Room Supervisor, how do you use the procedures to determine which relays should be checked in order to help you evaluate which power source to restore first?

a. Use SOP-32, "345KV Switchyard" for an attachment which lists all Switchyard relays and expected status of those relays for a Station Blackout event.

b. EOP Supplement 28, "Supplementary Actions for Loss of Power" will direct you to use EOP Supplement 22, "Switchyard Relay/Target List".

c. EOP Supplement 21, "Restoration of F or R Buses" will direct you to use EOP Supplement 22, "Switchyard Relay/Target List".

d. At Step 16 of EOP-3.0, "Station Blackout Recovery" you will be directed to use EOP Supplement 29, "Restore Buses 1C, 1D, 1E Power from Off- Site Source".

SENIOR REACTOR OPERATOR Page 7 of 71 QUESTION: 005 (1.00)

A Station Blackout has occurred concurrent with THREE stuck rods. Both Diesel Generators are running and have failed to automatically load. Both Steam Generator levels are at approximately -10% and lowering with NO Auxiliary Feedwater flow.

Which one of the following describes the required sequence of mitigation strategy?

a. Sequential Actions:

1. Open RPS breakers 42-1 and 42-2.

2. Close D/G output breakers.

3. Start P-8B Auxiliary Feedwater Pump.

b. Sequential Actions:

1. Start P-8B Auxiliary Feedwater Pump.

2. Close D/G output breakers.

3. Open RPS breakers 42-1 and 42-2.

c. Sequential Actions:

1. Open RPS breakers 42-1 and 42-2.

2. Start P-8B Auxiliary Feedwater Pump.

3. Close D/G output breakers.

d. Sequential Actions:

1. Close D/G output breakers.

2. Start P-8B Auxiliary Feedwater Pump.

3. Open RPS breakers 42-1 and 42-2.

SENIOR REACTOR OPERATOR Page 8 of 71 QUESTION: 006 (1.00)

Given the following conditions:

- The plant is operating at 100% power.

- A small break LOCA occurs inside containment.

- EOP-4.0, "Loss of Coolant Accident Recovery" is being implemented.

- Reactor Vessel Level Monitoring System (RVLMS) has ALL red lights LIT for both channels.

If this condition worsens, how will Primary Coolant System temperature and Main Steam pressure respond, and what action is required to address the condition?

a. As CETs continue to indicate saturated conditions, main steam pressure will LOWER due to code safety operation. Transition from EOP-4.0, to EOP-9.0, "Functional Recovery Procedure."

b. CETs will rapidly rise to indicate superheated conditions, main steam pressure will RISE. Transition from EOP-4.0, to EOP-9.0, "Functional Recovery Procedure."

c. CETs will rapidly rise to indicate superheated conditions, main steam pressure will RISE. Remain in EOP-4.0 and implement EOP Supplement 26, "PCS Void Removal".

d. As CETs continue to indicate saturated conditions, main steam pressure continues to RISE. Remain in EOP-4.0 and implement EOP Supplement 20, "Hot Leg Injection Via PZR".

SENIOR REACTOR OPERATOR Page 9 of 71 QUESTION: 007 (1.00)

Given the following conditions:

- The reactor has been manually tripped due to a small break LOCA.

- The operating crew has just begun carrying out the Immediate Actions of EOP-1.0, "Standard Post-Trip Actions".

- A fire is reported in Cable Spreading Room and large amounts of smoke and noxious fumes are entering the Control Room.

- Pressurizer pressure is 1300 psia and lowering and NO Safety Injection equipment has actuated, and cannot be actuated manually.

- You have issued the order to evacuate the Control Room.

To ensure safety injection flow to the core, you need to dispatch ...

a. ONE operator to Bus 1D to manually start LPSI P-67A and open at least one loop injection MOV with local switch.

b. ONE operator to Bus 1C to manually start LPSI P-67B and open at least one loop injection MOV with local switch.

c. TWO operators; ONE to Bus 1D to manually start HPSI P-66A and ONE to Panel C-150A to open at least one loop injection valve with local switch.

d. TWO operators; ONE to Bus 1C to manually start HPSI P-66B and ONE to Panel C-33 to open at least one loop injection valve with local switch.

SENIOR REACTOR OPERATOR Page 10 of 71 QUESTION: 008 (1.00)

Which of the following describes the Technical Specification applicability for PCS Specific Activity, including the basis?

This Technical Specification is applicable in ...

a. MODES 1, 2, and 3 with Tave > 300°F based on the lift settings for the ADVs and the Turbine Bypass Valve.

b. MODES 1, 2, and 3 with Tave > 500°F based on the lift settings for the ADVs and the main steam safety valves.

c. MODES 1, 2, 3, and 4 with Tave > 500°F based on the analyzed failure mode of the ADV and Turbine Bypass Valve controller.

d. MODES 1, 2, and 3 with Tave > 300°F based on the lift settings for the ADVs and the main steam safety valves.

SENIOR REACTOR OPERATOR Page 11 of 71 QUESTION: 009 (1.00)

Given the following:

- The plant is at full power.

- Testing of the Main Turbine Protective Trips is in progress.

- The operator at the front pedestal inadvertently causes the Main Turbine to trip, and immediately notifies the Control Room of what happened.

- All plant equipment functions as designed.

Which of the following is the correct Trip Classification of this event, and what procedures will be implemented?

a. Implement EOP-1.0, "Standard Post Trip Actions", and then EOP-2.0, "Reactor Trip Recovery". This is a Condition I trip, and does NOT require a PRC review prior to restart.

b. Implement EOP-1.0, "Standard Post Trip Actions", and then EOP-9.0, "Functional Recovery Procedure". This is a Condition II trip, and DOES require a PRC review prior to restart.

c. Implement EOP-2.0, "Reactor Trip Recovery", and then EOP-9.0, "Functional Recovery Procedure". This is a Condition I trip, and does NOT require a PRC review prior to restart.

d. Implement EOP-1.0, "Standard Post Trip Actions", and then EOP-2.0, "Reactor Trip Recovery". This is a Condition II trip, and DOES require a PRC review prior to restart.

SENIOR REACTOR OPERATOR Page 12 of 71 QUESTION: 010 (1.00)

For full power plant conditions, which one of the following conditions affecting the ability to makeup to the Primary Coolant System requires notification to the Nuclear Regulatory Commission?

a. Charging Pump P-55B is inoperable and will be restored in 4 days.

b. LPSI P-67A is inoperable and will be restored in 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />.

c. Safety Injection Tank T-82A pressure is 180 psig and lowering. It will be restored to normal in 4 days.

d. Boric Acid Pump P-56A spuriously started and was manually stopped. Repairs will require 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br />.

QUESTION: 011 (1.00)

Refer to the provided LCO 3.3.1, page 3.3.1-2.

Which of the following describes a failure of equipment that would require entry into CONDITION D, and what is the basis for the REQUIRED ACTION? (Assume APPLICABILITY conditions exist.)

a. One Source Range channel NI becomes inoperable. The Safety Analysis relies on the Source Range NIs to remove ZPM Bypass for a Continuous Rod Withdrawal.

b. The ZPM bypass key can be removed from the keyswitch in BYPASS. Affected RPS trips are bypassed for reactor protection in the event of a Control Rod Ejection.

c. One Power Range channel NI becomes inoperable. The Safety Analysis relies on the Power Range NIs to remove ZPM Bypass for a Control Rod Ejection.

d. One Wide Range channel NI becomes inoperable. The Safety Analysis relies on the Wide Range NIs to remove ZPM Bypass for a Continuous Rod Withdrawal.

SENIOR REACTOR OPERATOR Page 13 of 71 QUESTION: 012 (1.00)

Given the following conditions:

- A Steam Generator Tube Rupture in "A" S/G has occurred.

- The actions of EOP-5.0, "Steam Generator Tube Rupture Recovery" are being implemented.

- "A" S/G has not yet been isolated.

What direction should be given concerning control of PCS pressure, what is the basis for it?

a. Within the limits of EOP Supplement 1, "Pressure Temperature Limit Curves" to minimize PCS dilution and maintain Shutdown Margin.

b. Within the limits of EOP Supplement 2, "PCS Cooldown Strategy" to reduce potential lifting of a Main Steam Code Safety valve.

c. Less than 940 psia to minimize the potential for a radiation release to the environment.

d. Less than 940 psia to minimize PCS dilution, and maintain Shutdown Margin.

SENIOR REACTOR OPERATOR Page 14 of 71 QUESTION: 013 (1.00)

Given the following plant conditions:

- PCS temperature is 420°F.

- LTOP System is armed in LTOP Mode.

- Charging pump P-55A is in operation.

- Letdown is in service.

- 125 VDC Panel D11-1 has deenergized due to a fault.

To address these conditions, the Control Room Supervisor will ...

a. direct the crew to reestablish Charging and Letdown flow per SOP- 2A, "Chemical and Volume Control" since letdown flow automatically isolated due to loss of D-11-1.

b. implement ONP-23.1, "Primary Coolant System Leak" since RV-2006 has lifted due to closure of CV-2009 (Letdown Containment Isolation Valve), and will not reseat.

c. direct the crew to bypass the CVCS purification demineralizers due to CV-0909, Letdown Hx CCW Outlet, failing CLOSED.

d. implement ONP-23.1, "Primary Coolant System Leak" since a PORV has lifted due to loss of D-11-1.

QUESTION: 014 (1.00)

Which one of the following describes the mitigation strategy for the Safety Function "Maintenance of Vital Auxiliaries - Air" during the performance of EOP-9.0, "Functional Recovery Procedure"?

a. There is only ONE Success Path and it requires availability of 2400 VAC safety related power.

b. There is only ONE Success Path and it does NOT require availability of 2400 VAC safety related power.

c. There are TWO Success Paths and they BOTH require availability of 2400 VAC safety related power.

d. There are TWO Success Paths and only ONE requires availability of 2400 VAC safety related power.

SENIOR REACTOR OPERATOR Page 15 of 71 QUESTION: 015 (1.00)

Given the following conditions:

- Steam Generator Nozzle Dams are installed.

- Fuel is being moved from the core to the Spent Fuel Pool.

- EK-1349 and EK-1350, Containment Sump Hi Hi Level, alarm annunciates.

- Containment Radiation Monitors, RIA-2316 and RIA-2317, indicate rising radiation levels.

Which of the following procedures has IMMEDIATE ACTIONS which must be performed for these conditions?

a. ONP-11.1, Fuel Cladding Failure

b. ONP-11.2, Fuel Handling Accident

c. ONP-17, Loss of Shutdown Cooling

d. ONP-23.3, Loss of Refueling Water Accident QUESTION: 016 (1.00)

Which one of the following regulating rod group configurations requires entry into a Technical Specification Limiting Condition of Operation?

a. Group 1 at 90 inches Group 2 at 5 inches At least one action of Tech. Spec. 3.1.4, "Control Rod Alignment" applies.

b. Group 2 at 110 inches Group 3 at 35 inches At least one action of Tech. Spec. 3.1.6, "Regulating Rod Group Position Limits" applies.

c. Group 3 at 131 inches Group 4 at 45 inches At least one action of Tech. Spec. 3.1.6, "Regulating Rod Group Position Limits" applies.

d. Group 1 at 131 inches Group 2 at 35 inches At least one action of Tech. Spec. 3.1.5, "Shutdown and Part Length Rod Group Insertion Limits" applies.

SENIOR REACTOR OPERATOR Page 16 of 71 QUESTION: 017 (1.00)

Per Technical Specification 3.7.6, "Condensate Storage and Supply", the Condensate Storage Tank (T-2) and the Primary Makeup Storage Tank (T-81) are required to contain a minimum inventory for applicable plant modes. The BASIS for this requirement is to provide enough cooling water to remove decay heat for at least ...

a. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following a Reactor trip from 100% Rated Thermal Power.

b. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a Reactor trip from 100% Rated Thermal Power.

c. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following a Reactor trip from 105% Rated Thermal Power.

d. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a Reactor trip from 102% Rated Thermal Power.

QUESTION: 018 (1.00)

Refueling operations are in progress when the following alarm annunciates:

- EK-1364, GASEOUS WASTE MONITORING HI RADIATION The alarm is due to a valid high alarm condition on RIA-5712, Fuel Handling Area Vent monitor.

In response to this condition, the Control Room Supervisor will use ...

a. SOP-24, "Ventilation and Air Conditioning Systems", to direct a manual shutdown of the Fuel Handling Area ventilation system.

b. ONP-11.2, "Fuel Handling Accident" which will direct verification of the automatic tripping of V-69 Supply Fan, and to direct certain manual actions.

c. SOP-38, "Gaseous Process Monitoring System" which will direct verification of the automatic tripping of V-70A and V-70B Exhaust Fans.

d. SOP-24, "Ventilation and Air Conditioning Systems" which will direct verification of the automatic tripping of only ONE of the V-70 Exhaust Fans.

SENIOR REACTOR OPERATOR Page 17 of 71 QUESTION: 019 (1.00)

For a Loss of Load event, which one of the following describes the impact of a failure of the Reactor Protection System to automatically trip the reactor, and what procedure is used to mitigate the condition? (Assume NO operator action.)

Pressurizer pressure rises to ...

a. 2235 psia and the reactor automatically trips. Implement EOP-1.0, "Standard Post Trip Actions".

b. 2235 psia and the reactor automatically trips. Implement EOP-9.0, "Functional Recovery Procedure".

c. 2375 psia and the reactor automatically trips. Implement EOP-1.0, "Standard Post Trip Actions".

d. 2375 psia and the reactor automatically trips. Implement EOP-9.0, "Functional Recovery Procedure".

QUESTION: 020 (1.00)

A Containment Purge is to be performed with the plant in MODE 5, using the Containment Purge Exhaust Valves, CV-1805, 1806, 1807, and 1808.

Which document is used to provide guidance for monitoring or controlling the Containment Purge, and what is required?

a. Health Physics 6.14, "Containment Purge" prescribes a purge flow rate of LESS than 100 scfm.

b. SOP-24, "Ventilation and Air Conditioning System" requires logging the times of Containment Purge Exhaust valves operation.

c. The Offsite Dose Calculation Manual allows a flow rate (up to a maximum of 100 scfm) that results in a nuclide sum fraction of < 10.0.

d. The Operating Requirements Manual requires the Containment Purge Exhaust valves to be open for NO MORE THAN a total of 30 minutes.

SENIOR REACTOR OPERATOR Page 18 of 71 QUESTION: 021 (1.00)

Which document delineates the operational responsibilities for breaker operations in the Palisades Plant 345 KV Switchyard?

a. Admin Procedure 4.00, "Operations Organization, Responsibilities, and Conduct".

b. Admin Procedure 4.14, "Conduct of Operations".

c. Admin Procedure 4.28, "Control of Palisades Switchyard Activities".

d. SOP-30, "Station Power".

QUESTION: 022 (1.00)

For the following conditions:

- The reactor is not critical.

- Tave is at 532°F

- Group A and Group B Shutdown Rods are fully withdrawn.

- All part length rods are withdrawn.

Which of the following describes the resulting mode change when withdrawing the FIRST regulating rod?

This is a mode change from ...

a. MODE 3 to MODE 2 and requires authorization from the Plant Manager.

b. MODE 3 to MODE 2 and requires authorization from the Reactor Engineering Manager.

c. MODE 2 to MODE 1 and requires authorization from the Reactor Engineering Manager.

d. MODE 2 to MODE 1 and requires authorization from the Site Vice President.

SENIOR REACTOR OPERATOR Page 19 of 71 QUESTION: 023 (1.00)

During an outage, a portion of the water in the hotwell is to be released to the lake, per COP-31, "Non-Radiological Environmental Operating Procedure".

For reviewing and approving this release, the Shift Supervisor (SS) is required to ensure that the

a. batch volume has not changed, required dilution flow is met. After the release, SS forwards Discharge Authorization to the Environmental Coordinator.

b. Discharge Authorization was prepared by a qualified Chemistry Technician.

Check at least ONE Dilution Water pump in service. After the release, SS forwards Discharge Authorization to the Certified Waste Treatment Plant Operator (WTPO).

c. required dilution flow is met. After the release , SS forwards Discharge Authorization to the Chemistry and Rad Services Supervisor.

d. Discharge Authorization was prepared by the Environmental Coordinator, batch volume has not changed, required dilution flow is met. After the release, SS forwards Discharge Authorization to the Certified Waste Treatment Plant Operator.

SENIOR REACTOR OPERATOR Page 20 of 71 QUESTION: 024 (1.00)

Given the following conditions:

- A reactor trip has occurred due to a loss of offsite power.

- Subsequently, Startup Power has been restored to the plant.

- It is now desired to restart a Primary Coolant Pump per EOP Supplement 3, "Starting Primary Coolant Pumps".

One of the start criteria in this EOP Supplement is that the average of Qualified CETs must be at least 25°F subcooled. The BASIS for this requirement is to prevent ...

a. emptying the Pressurizer.

b. overpressurizing the PCS.

c. pump cavitation and damage.

d. reactor head voiding.

QUESTION: 025 (1.00)

The plant is operating at full power on "A" Shift when the NCO informs the CRS of a rising Charging flow rate and lowering Pressurizer level.

- At 0113 hours0.00131 days <br />0.0314 hours <br />1.868386e-4 weeks <br />4.29965e-5 months <br /> the NCO informs the CRS that "A" Charging Pump is at full speed and that "B" and "C" Charging Pumps have started.

- Pressurizer level is continuing to lower.

- At 0117 hours0.00135 days <br />0.0325 hours <br />1.934524e-4 weeks <br />4.45185e-5 months <br /> the CRS directs a manual reactor trip.

By which one of the following times should the current emergency be classified in accordance with the Emergency Plan?

a. 0128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br />

b. 0132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br />

c. 0147 hours0.0017 days <br />0.0408 hours <br />2.430556e-4 weeks <br />5.59335e-5 months <br />

d. 0213 hours0.00247 days <br />0.0592 hours <br />3.521825e-4 weeks <br />8.10465e-5 months <br />

SENIOR REACTOR OPERATOR Page 21 of 71 QUESTION: 026 (1.00)

During a planned power reduction from full power Group 4 rod positions are indicating as follows:

- Rod 41 - 122.0"

- Rod 40 - 123.3"

- Rod 39 - 112.2"

- Rod 38 - 114.1" What adverse consequences are of concern with these rod positions?

a. Rod 41 and 40 CRDM motors will overheat due to continuous drivedown.

b. Power peaking limits may have been exceeded.

c. Excessive negative reactivity has been inserted into the core.

d. Uncontrollable Xenon oscillations will be induced.

QUESTION: 027 (1.00)

The plant was at 100% power and operating normally when the following occurred:

- P-50A Primary Coolant Pump (PCP) ammeter pegged HIGH and then dropped to 0 amps and remained at 0.

- The reactor automatically tripped approximately 1-2 seconds later.

- PCPs P-50B, C, D remain operating normally.

Which of the following would account for all of the above conditions?

a. P-50A has a seized pump shaft and the reactor tripped due to low Primary Coolant flow.

b. P-50A has a sheared pump shaft and the reactor tripped due to low Primary Coolant flow.

c. The feeder breaker to Bus 1A tripped due to a fault which caused a Main Generator protective relay actuation and a reactor trip.

d. P-50A was cavitating which caused at least 2 out of 4 TM/LP channels to actuate an automatic reactor trip.

SENIOR REACTOR OPERATOR Page 22 of 71 QUESTION: 028 (1.00)

The plant was at 50% power when the power supplies for all Primary Coolant Pumps became de-energized. Which one of the following describes an operating behavior characteristic of the Primary Coolant System (PCS) that is present during natural circulation?

a. Loop T LESS than normal full power T due to the reduced core power level.

b.

Loop T GREATER than normal full power T since a higher thermal driving head is required.

c. Rate of steam generator pressure reduction is LESS than rate of PCS temperature reduction due to HIGHER thermal driving head.

d. Rate of steam generator pressure reduction is LESS than rate of PCS temperature reduction due to the REDUCED thermal driving head.

QUESTION: 029 (1.00)

The crew is initiating Emergency Manual Boration per EOP 1.0, "Standard Post Trip Actions" following a reactor trip with 2 full length control rods not fully inserted.

- Bus 1C and Bus 1D are both energized.

- One HPSI pump running.

- Boric Acid Pump P-56A is running.

- Charging Pump P-55A is operating.

Which of the following emergency boration methods should be selected if VCT outlet valve (MO-2087) is open and will NOT close from the Main Control Board?

a. Open MO-2169 and MO-2170, Gravity Feed Valves.

b. Open MO-2160, SIRWT to Charging Pump Suction.

c. Open MO-2140, Pumped Feed Valve.

d. Open MO-3072, CVCS to HPSI Train 2.

SENIOR REACTOR OPERATOR Page 23 of 71 QUESTION: 030 (1.00)

Which one of the following Engineered Safeguards Features will result in a loss of Component Cooling Water (CCW) to the Primary Coolant Pumps (PCPs), and what is the reason for the alignment?

a. Containment High Pressure (CHP) - ensures containment building performs its design function.

b. Containment High Radiation (CHR) - reduces radiation release potential by isolating CCW headers.

c. Safety Injection Signal (SIS) on Low Pressurizer Pressure - Since operators will be manually tripping PCPs for Low Pressurizer Pressure events, this provides more cooling for SIS actuated components.

d. Recirculation Actuation Signal (RAS) - ensures adequate cooling capability for the hotter containment sump water following a Large Break LOCA, by isolating CCW to the PCPs.

QUESTION: 031 (1.00)

Given the following plant parameters during a Primary Coolant System (PCS) heatup:

- PCS temperature is at 195°F

- Pressurizer pressure is 250 psia

- Pressurizer temperature is 380°F

- Pressurizer level is 100%

- P-50B PCP is the only Primary Coolant Pump in service.

A Pressurizer Pressure Control System malfunction causes both spray valves to open and remain open. Why does the reactor operator NOT expect to see an immediate reduction in Pressurizer pressure?

a. At this pressure and temperature, heat input is matching spray flow effectiveness.

b. Subcooling conditions are indicated in the pressurizer.

c. Pressurizer spray flow is inadequate for given conditions.

d. CVCS backpressure regulator will automatically open to maintain pressure.

SENIOR REACTOR OPERATOR Page 24 of 71 QUESTION: 032 (1.00)

The Plant has tripped and the immediate actions of EOP-1.0 are in progress. It is noted that BOTH Steam Generator pressures are approximately 760 psia and lowering slowly.

Which one of the following actions is the operator required to perform?

a. Manually trip BOTH Main Feedwater pumps.

b. Manually initiate Safety Injection Actuation signal

c. Verify main steam isolation signal

d. Close BOTH Main Steam Isolation Valves QUESTION: 033 (1.00)

Given the following:

- A Main Steam Line Break has occurred upstream of the "B" S/G MSIV.

- Main Steam Line Isolation has automatically actuated.

Which one of the following is of concern if a steaming path from the unaffected steam generator is not established immediately following dryout of the affected steam generator?

a. Void formation in the Reactor Vessel upper head region.

b. Rise in core exit temperatures causing a loss of natural circulation.

c. Rapid rise in Tcold of the unaffected loop which would result in a loss of natural circulation.

d. Rapid repressurization of the Primary Coolant System and subsequent pressurized thermal shock.

SENIOR REACTOR OPERATOR Page 25 of 71 QUESTION: 034 (1.00)

Given the following conditions:

- The plant is operating at 85% power.

- Cooling Tower Pump B trips.

- Main Condenser vacuum begins lowering, as prescribed by ONP-14, "Loss of Condenser Vacuum".

- The crew begins lowering power using ONP-26, Rapid Power Reduction.

- When power level reaches 55% during the power reduction, EK-0111, VACUUM LO, alarms due to vacuum at 24" Hg.

- Vacuum is at 22" and CONTINUES LOWERING.

Which of the following actions are required to be taken?

a. Trip the turbine, verify the reactor automatically trips, and go to EOP-1.0, Standard Post-Trip Actions.

b. Trip the reactor, verify the turbine automatically trips, and go to EOP-1.0, Standard Post-Trip Actions.

c. Continue the rapid power reduction until condenser vacuum stabilizes.

d. Continue the power reduction, using normal de- escalation rates, until condenser vacuum stabilizes.

SENIOR REACTOR OPERATOR Page 26 of 71 QUESTION: 035 (1.00)

Given the following conditions:

- The plant is in MODE 2.

- Charging Pump P-55A is operating.

- Charging Pumps P-55B and P-55C are in AUTO.

- All control systems are aligned normally and functioning properly.

- A loss of Instrument AC Bus Y-01 occurs.

Which of the following describes why ONP-24.5, "Loss of Instrument AC Bus Y01" directs the operators to isolate PCP bleedoff from the Volume Control Tank (VCT) and realign it to the Primary System Drain Tank?

a. Minimizes the likelihood of gas intrusion in to the PCP seals.

b. Minimizes the amount of pressure reduction in the VCT.

c. Prevents a complete draining of the VCT.

d. Prevents overfilling of the VCT.

SENIOR REACTOR OPERATOR Page 27 of 71 QUESTION: 036 (1.00)

Given the following plant conditions:

- During power operations a Large Break Loss of Coolant Accident (LOCA) has occurred.

- A Safety Injection has occurred as designed.

- A Containment High Pressure (CHP) also occurred as designed.

- 30 minutes later the Safety Injection & Refueling Water Tank level is at 1.8%.

(Refer to the attached drawing.) What is the expected response of the highlighted valves (CV-0823 and CV-0826) and what is the reason for that response?

a. Failed AS IS to provide a gradual cooling of the containment sump fluid.

b. Modulating to provide a gradual cooling of the containment sump fluid.

c. Opened to a preset hardstop to provide maximum cooling for containment spray and safety injection recirculation flow.

d. Closed to a preset hardstop to ensure maximum cooling water flow is maintained for the Containment Air Coolers.

SENIOR REACTOR OPERATOR Page 28 of 71 QUESTION: 037 (1.00)

Given the following conditions:

- There is a fire at the north end of the Service Building.

- This fire is generating a significant amount of smoke.

- NO radiological event is in progress.

- Wind direction is from due North at 5 to 10 miles per hour.

For these conditions, there is a need to operate the Control Room HVAC system in the Recirculation mode because it will ...

a. pressurize the Control Room envelope, thereby preventing entry of smoke and other contaminants.

b. filter out smoke through the charcoal filters, thereby preserving habitability of the Control Room envelope.

c. minimize smoke intake in the Control Room envelope, and also protect the charcoal filters.

d. maximize fresh air intake and purge the Control Room envelope of any smoke particles.

SENIOR REACTOR OPERATOR Page 29 of 71 QUESTION: 038 (1.00)

Given the following initial plant conditions:

- The plant is at 100% power.

- The Personnel Air Lock between the seals test (DWO-13) has just been completed and BOTH seals have FAILED.

Subsequent plant conditions:

- A Main Steam Line Break inside containment occurs on the "A" Steam Generator (S/G).

- EOP-1.0, "Standard Post Trip Actions" is in progress.

- P-8A Aux. Feedwater Pump is in service providing 165 gpm to each S/G.

- A Containment High Pressure (CHP) has actuated.

- SIRW tank level is at 85% at lowering.

- CV-0510 "A" S/G MSIV is stuck OPEN and will not close by any means.

- CV-1359 (Non-Critical Service Water Isolation) did NOT automatically close.

- P-52C Component Cooling Water Pump did NOT automatically start.

Which one of the following operations will result in a REDUCTION of leakage out of the failed Personnel Air Lock seals?

a. Manually start P-52C Component Cooling Water Pump.

b. Manually close CV-1359 (Non-Critical Service Water Isolation).

c. RAISE Auxiliary Feedwater flow to "A" S/G to 200 gpm.

d. LOWER Auxiliary Feedwater flow to "B" S/G to 100 gpm.

SENIOR REACTOR OPERATOR Page 30 of 71 QUESTION: 039 (1.00)

Given the following:

- During a power escalation at 18% power, Containment Radiation Monitors are indicating as follows:

RIA-1805 = 8 R/hr. RIA-1806 = 11R/hr.

RIA-1807 = 10.5R/hr. RIA-1807 = 8 R/hr.

- A fuel cladding failure has been verified.

- The crew has entered ONP-11.1, "Fuel Cladding Failure".

Which one of the following describes the reason for any required actions?

a. Manual alignment of Control Room HVAC to Emergency mode is required to maintain Control Room habitability.

b. Operators are required to close Letdown Orifice Stop Valves, since letdown has automatically isolated.

c. Operators must closely monitor Pressurizer level and maintain it at less than 78.2% since letdown has automatically isolated.

d. Since the Stack Gas Monitor is expected to be in high alarm, operators are required to start an additional Main Exhaust Fan, V-6A/B.

QUESTION: 040 (1.00)

During a critical approach, all Group 4 Regulating Rods start to continuously withdraw with NO operator action. What Control Room indications can be used to determine that a continuous rod withdrawal is occurring?

a. Associated core matrix indicating lights change from RED to AMBER.

b. EK-0911, "ROD POSITION 4 INCHES DEVIATION" annunciates.

c. Rod Deviation (RED) light is ON for Group 4 rod indicating lights on control panel C-02.

d. PPC GREEN indicating bars on Page 410 for Group 4 rod positions are getting SHORTER.

SENIOR REACTOR OPERATOR Page 31 of 71 QUESTION: 041 (1.00)

Which one of the following conditions would PREVENT retrieval of a dropped control rod?

a. A valid EK-0916, "CONTROL RODS OUT OF SEQUENCE" is annunciating.

b. Motor Control Center 10 is de-energized for diagnostic testing.

c. Rod Drive Control System power supply switch is in "Bus #2" position.

d. A valid EK-0605, "VARIABLE HIGH POWER LEVEL CHANNEL PRE-TRIP" is annunciating.

QUESTION: 042 (1.00)

Following a reactor trip caused by a loss of feedwater to the Steam Generators, one of the Pressurizer code safety valves is stuck slightly open. The following parameters are noted:

- PCS pressure = 900 psia

- PZR vapor space temperature = 532°F

- Quench Tank level = 50%

- Quench Tank pressure = 20 psig What is the expected tail pipe temperature for the above plant conditions?

a. 532°F

b. 360°F

c. 315°F

d. 212°F

SENIOR REACTOR OPERATOR Page 32 of 71 QUESTION: 043 (1.00)

Following a small break LOCA, the following conditions are observed:

- Core exit thermocouple temperatures are approximately 650°F and stable.

- PCS hot leg temperatures are approximately 550°F and stable.

- Pressurizer pressure is 1100 psia.

- PCS cold leg temperatures are approximately 330°F and lowering slowly.

What is the status of PCS inventory and core cooling? The core is ...

a. covered and being cooled by natural circulation.

b. partially uncovered and being cooled by natural circulation.

c. covered and being cooled by reflux boiling.

d. partially uncovered and being cooled by reflux boiling.

QUESTION: 044 (1.00)

The plant was on Shutdown Cooling when a loss of Shutdown Cooling occurred due to a seized bearing on P-67A Low Pressure Safety Injection (LPSI) Pump. The crew is now ready to start the alternate LPSI Pump (P-67B) to restore shutdown cooling flow.

Which one of the following flow rates for P-67B is the MINIMUM acceptable flow rate for pump protection?

a. 150 gpm

b. 250 gpm

c. 500 gpm

d. 2810 gpm

SENIOR REACTOR OPERATOR Page 33 of 71 QUESTION: 045 (1.00)

Source/Wide Range NI - 1/3A must be taken out of service. Prior to removing NI - 1/3A from service, which of the following conditions regarding the High SUR Trip RPS channels would be acceptable? (Assume all other Technical Specification requirements are met.)

RPS A RPS B RPS C RPS D

a. NORMAL BYPASS NORMAL TRIP

b. BYPASS NORMAL TRIP NORMAL

c. TRIP NORMAL NORMAL BYPASS

d. NORMAL TRIP BYPASS NORMAL

SENIOR REACTOR OPERATOR Page 34 of 71 QUESTION: 046 (1.00)

The alarm "PROCESS LIQ MONITORING HI RADIATION" annunciates due to a high alarm condition on RIA-0707, Steam Generator Blowdown Radiation Monitor.

The following valve positions are subsequently noted:

CV-0704, Blowdown Tank Discharge to Mixing Basin CLOSED CV-0738, B S/G Surface Blowdown CLOSED CV-0739, A S/G Surface Blowdown CLOSED CV-0770, B S/G Bottom Blowdown OPEN CV-0771, A S/G Bottom Blowdown OPEN Which of the following is the correct diagnosis of the above valve positions?

a. Per design. S/G sampling capability is maintained through bottom blowdown CVs, and secondary plant contamination is minimized by closing the surface blowdown CVs.

b. Per design. S/G sampling capability is lost since the surface blowdown CVs are closed. Bottom blowdown CVs remain open to allow further trending of RIA-0707.

c. NOT per design. ALL valves should be CLOSED to prevent secondary plant contamination.

d. NOT per design. ALL valves should remain OPEN to maintain full S/G sampling capability.

SENIOR REACTOR OPERATOR Page 35 of 71 QUESTION: 047 (1.00)

Given the following conditions:

- EOP-7.0, Loss of All Feedwater, actions are in progress.

- The crew is implementing a cooldown in order to use the Condensate Pumps for feeding the Steam Generators (S/G).

- Feed Reg Bypass Valves (CV-0734, CV-0735) have been positioned to 10%

open as read on valve position indicators on panel C-01.

Given the following information:

- Feed pump discharge pressure = 500 psia

- S/G pressure = 420 psia How much flow is being delivered to EACH S/G from the Condensate Pumps?

a. 300 gpm

b. 100 gpm

c. 140 gpm

d. 125 gpm QUESTION: 048 (1.00)

Which one of the following operations is performed at Palisades that reduces the potential consequences if an Accidental Liquid Radwaste Release event were to occur?

a. Maintaining the Spent Fuel Pool level above the low level alarm setpoint.

b. Minimizing the amount of weir overflow from the Makeup Basin to the Mixing Basin.

c. Recirculating T-91 Utility Water Storage Tank through demineralizers if the normal value of gamma (in µCi/ml) is exceeded.

d. Maximizing Cooling Tower blowdown in order to ensure the amount of tritium (in

µCi/ml) is maintained below allowable limits.

SENIOR REACTOR OPERATOR Page 36 of 71 QUESTION: 049 (1.00)

For a rupture of the Volume Control Tank and subsequent gaseous release, what operational requirement ensures required dose limits are not exceeded?

a. Limiting primary coolant gross gamma activity to less than 100 µCi/gm.

b. Maintain a hydrogen overpressure on the Volume Control Tank in MODE 1.

c. Ensuring primary coolant lithium concentration is less than 1.0 ppm.

d. Operating with no more than 3% failed fuel in all plant modes.

QUESTION: 050 (1.00)

Given the following conditions:

- The plant is at full power.

- Train "A" of Control Room HVAC is in service in Normal Mode.

- The following alarm annunciates:

EK-0239, "CRHVAC TRAIN A RIA-1818A HI RAD/FAIL"

- It is determined that RIA-1818A has failed and is inoperable.

What are the consequences of continuing to operate the Control Room HVAC system in these conditions?

a. If a CHP/CHR occurs ONLY the operating train will FAIL to automatically swap to Emergency Mode.

b. If a CHP/CHR occurs NEITHER train will automatically swap to Emergency Mode.

c. Due to the loss of RIA-1818A, any radioactive contamination entering the Control Room from the outside will not be detected.

d. Due to the loss of RIA-1818A, Train "A" CRHVAC automatically swaps to Purge Mode and the Control Room depressurizes.

SENIOR REACTOR OPERATOR Page 37 of 71 QUESTION: 051 (1.00)

The crew is implementing ONP-25.1, "Fire Which Threatens Safety-Related Equipment" for a fire inside containment. Why does ONP-25.1 refer the operators to EOP-9.0, "Functional Recovery Procedure" Attachment G-1 for these conditions?

a. Determine which Success Paths are available for mitigating the event.

b. Establish the hierarchy of which safety functions to address first.

c. Establish which Continuing Actions to perform after the fire is out.

d. Determine actions for inoperable instrumentation inside containment.

QUESTION: 052 (1.00)

Given the following conditions:

- Plant is at 100%.

- Level Controller LIC 0101B is in CASCADE.

- Level Control Selector is in Channel B.

- Pressurizer level transmitter LT-0101B diaphragm ruptures.

What is the resulting effect on actual Pressurizer level and the reason for it?

a. Level LOWERS due to the controllers normal level control signal ramping to letdown.

b. Level LOWERS due to the unselected controllers backup signal overriding the failed signal.

c. Level RISES due to the backup signal being calculated from Tave and overriding the failed signal.

d. Level RISES due to the controllers normal level control signal ramping to charging.

SENIOR REACTOR OPERATOR Page 38 of 71 QUESTION: 053 (1.00)

Given the following plant conditions:

- The plant is on shutdown cooling (SDC) in reduced inventory.

- P-67A LPSI Pump is in service. P-67B is NOT running.

- All SDC system controls are aligned normally.

- A leak develops in the Primary Coolant System.

- A moment or two later LPSI injection flow becomes erratic and the following valid alarm then annunciates:

- EK-1157, LO PRESS SI PUMPS P-67A & P-67B TRIP Which one of the following describes any required operator actions for these conditions?

a. Closely monitor operating parameters of P-67B since it has auto started, and there are potential cavitation concerns.

b. Since the LPSI Pump STANDBY auto start feature is not used, the operator must manually start P-67B to maintain SDC flow.

c. Since the LPSI Pump STANDBY auto start feature is not used, the operator must report that a Loss of Shutdown Cooling event has occurred.

d. The operator should attempt only one restart of P- 67A to prevent exceeding motor starting duty limitations and to avoid a loss of shutdown cooling.

QUESTION: 054 (1.00)

What is the concern for two adjacent control rods that are determined to be untrippable, but moveable?

a. On a reactor trip a portion of the core would have excess reactivity until the two rods could be inserted.

b. For an emergency downpower the Axial Shape Index (ASI) could not be maintained within the prescribed band.

c. On a reactor trip the required Shutdown Margin could not be achieved using Emergency Boration.

d. For an emergency downpower the Power Dependent Insertion Limits would be violated for the two affected rods.

SENIOR REACTOR OPERATOR Page 39 of 71 QUESTION: 055 (1.00)

Assume the Plant is in MODE 3 with both Steam Generators available. Which statement describes the effect on the Primary Coolant System (PCS) of the number of operating Primary Coolant Pumps (PCPs)?

a. Operating ALL PCPs raises PCS flow rate, but results in a reduction in DNB margin due to pump heat input.

b. Fifteen minutes after shutting off ALL PCPs there will be NO flow in the PCS, and margin to DNB will be reduced.

c. Reducing the number of operating PCPs lowers the PCS flow rate which causes a RISE in DNB margin.

d. Reducing the number of operating PCPs lowers the PCS flow rate which causes a REDUCTION in DNB margin.

QUESTION: 056 (1.00)

The crew is implementing EOP-8.0, "Loss of Forced Circulation Recovery" and is using Auxiliary Spray to control Pressurizer pressure. TWO Charging Pumps are in service.

If the operator desires to REDUCE Pressurizer pressure, which of the following methods should be used?

a. Start a third Charging Pump and open the Main Spray valves.

b. Start a third Charging Pump and close the Main Spray valves.

c. Shutoff one Charging Pump and open the Main Spray valves.

d. Shutoff one Charging Pump and close the Main Spray valves.

SENIOR REACTOR OPERATOR Page 40 of 71 QUESTION: 057 (1.00)

Given the following conditions:

- A small break LOCA has occurred and the Control Room crew is performing the actions of EOP-4.0, "Loss of Coolant Accident Recovery".

- Pressurizer pressure is 980 psia and very slowly lowering.

- A plant cooldown has been initiated using the Steam Generators and Auxiliary Feedwater.

- Safety Injection throttling criteria have been met and the operator is ready to throttle Safety Injection by shutting off one HPSI pump.

What plant response should the operator expect when throttling Safety Injection for the above conditions?

a. The cooldown rate will LOWER unless the operator raises the steaming rate.

b. The cooldown rate will RISE unless the operator lowers the steaming rate.

c. The resulting unbalanced loop injection flows will interrupt natural circulation unless the operator raises the steaming rate.

d. The resulting Pressurizer pressure reduction may result in core voiding unless the operator lowers the steaming rate.

QUESTION: 058 (1.00)

Which of the following lists the normal power supplies for the indicated Nuclear Instruments?

NI-1 NI-2 NI-3 NI-4 NI-5 NI-6 NI-7 NI-8

a. Y10 Y20 Y10 Y20 Y30 Y40 Y30 Y40

b. Y40 Y30 Y40 Y30 Y10 Y20 Y10 Y20

c. 30 Y40 Y30 Y40 Y10 Y20 Y30 Y40

d. Y10 Y20 Y30 Y40 Y10 Y20 Y30 Y40

SENIOR REACTOR OPERATOR Page 41 of 71 QUESTION: 059 (1.00)

The Cutler-Hammer Interface has failed and therefore QCET indication is not available on the PPC.

QCET temperatures can then be monitored using which one of the following?

a. TYT-0100 or TYT-0200

b. SPI Node

c. PIP Node

d. C-11A recorders QUESTION: 060 (1.00)

Which Containment Air Cooler (CAC) fans have power available if D/G 1-2 is the ONLY available source of AC power?

a. V-1A, V-2A, and V-3A

b. V-4A only

c. ALL CAC "A" fans

d. ALL CAC "A" and "B" fans

SENIOR REACTOR OPERATOR Page 42 of 71 QUESTION: 061 (1.00)

The plant is operating at 100% power when BOTH Condensate Pumps P-2A and P-2B unexpectedly trip.

What is the resulting effect on the Main Feedwater (MFW) Pumps?

a. They trip due to a reduction in MFW pump suction pressure.

b. They trip due to overspeeding of the MFW pump turbines.

c. MFW pumps will experience excessive vibration due to cavitation at the pump suction.

d. MFW pump turbines ramp down to minimum speed to prevent overfeeding Steam Generators.

QUESTION: 062 (1.00)

The plant is operating at 60% power with the Steam Generator Level Control System in automatic, when Annunciator EK-0961, "STEAM GEN E-50A HI LEVEL" alarms.

Which one of the following sets of indications would be expected immediately for the above plant conditions?

CV-0701 position indicator Steam Generator Level POI-0701

a. Lowering 55%

b. Rising 85%

c. Rising 55%

d. Lowering 85%

SENIOR REACTOR OPERATOR Page 43 of 71 QUESTION: 063 (1.00)

Given the following conditions:

- From full power a transient occurs that results in a valid Aux. Feedwater Actuation Signal (AFAS).

- P-8A Aux. Feed Pump is tagged out.

- Plant conditions also require a Natural Circulation plant cooldown.

- P-8C AFW Pump is the ONLY AFW Pp. operating.

- AFW flow to "A" S/G = 80 gpm and stable.

- AFW flow to "B" S/G = 120 gpm and stable.

Which one of the following describes AFW System response in establishing and maintaining a natural circulation cooldown?

a. acceptable since Auxiliary Feedwater flow to at least ONE S/G is greater than 100 gpm.

b. acceptable since Auxiliary Feedwater flow to BOTH S/Gs is greater than 70 gpm.

c. NOT acceptable since P-8B turbine driven pump should have auto started due to flow to ONE S/G at less than 100 gpm.

d. NOT acceptable since P-8B turbine driven pump should have auto started due to flow to BOTH S/Gs at less than 165 gpm.

QUESTION: 064 (1.00)

Containment is normally vented to the ____(1)____ via the _____(2)_____.

(1) (2)

a. Waste Gas Collection Hdr Shield Cooling Surge Tank

b. Main Exhaust Plenum Waste Gas Surge Tank

c. Waste Gas Surge Tank T-64A Clean Waste Receiver Tank

d. Vent Gas Collection Header T-64D Clean Waste Receiver Tank

SENIOR REACTOR OPERATOR Page 44 of 71 QUESTION: 065 (1.00)

Flammable gas mixtures are prevented in the Waste Gas Decay Tanks by ...

a. venting Volume Control Tank hydrogen if Waste Gas Decay Tanks oxygen exceeds 5%.

b. maintaining the Waste Gas Surge Tank at a slightly positive pressure.

c. maintaining the Vacuum Degasifier Tank with a nitrogen overpressure when in standby.

d. placing the Vacuum Degasifier Tank in service during all resin sluices.

QUESTION: 066 (1.00)

Placing the Fuel Handling Area Monitors RIA-2316 and RIA- 2317 cutout switches to the IN position will:

a. Enable automatic closure of selected Containment Isolation valves.

b. Trip the Fuel Handling Area Supply Fan V-7 on one out of two logic.

c. Enable automatic closure of Fuel Handling Area exhaust dampers.

d. Trip the Penetration and Fan Room V-78 and V-79 on high radiation.

SENIOR REACTOR OPERATOR Page 45 of 71 QUESTION: 067 (1.00)

Given the following:

The Plant is in MODE 3. Safety Injection Tank parameters are as follows:

SIT PRESSURE BORON (PPM)

T-82A 205 1750 T-82B 215 1920 T-82C 225 1705 T-82D 220 2150 Which ONE of the Safety Injection Tanks (SITs) will prevent entry into MODE 2 per Technical Specifications?

a. T-82A

b. T-82B

c. T-82C

d. T-82D QUESTION: 068 (1.00)

The plant is operating at 65% power when both pressurizer spray valves fail OPEN. With NO operator action, which of the following automatic actuations is expected to occur?

a. Safety Injection and then a Reactor trip.

b. Reactor trip and then a Safety Injection.

c. ONLY a Reactor trip.

d. ONLY a Safety Injection.

SENIOR REACTOR OPERATOR Page 46 of 71 QUESTION: 069 (1.00)

Given the following plant conditions:

- During a small break LOCA Pressurizer level begins slowly lowering.

- Pressurizer level drops to 34% before the operator notices the trend.

Which of the following describes the status of the Pressurizer heaters?

a. ONLY the proportional heaters are energized.

b. ONLY the backup heaters are energized.

c. ALL heaters are energized.

d. NO heaters are energized.

QUESTION: 070 (1.00)

Which one of the following describes the use of the PIP (Primary Indication Panel) indication as compared to the SPI (Secondary Position Indication) indication when monitoring control rod positions?

a. PIP is a MORE accurate indication since it receives input from the synchro-transmitters.

b. PIP is a LESS accurate indication since it receives input from the reed stack switches.

c. SPI is a LESS accurate indication since it receives input from the synchro-transmitters.

d. SPI is a MORE accurate indication since it receives input from the reed stack switches.

SENIOR REACTOR OPERATOR Page 47 of 71 QUESTION: 071 (1.00)

The following plant conditions exist:

- SIRW tank level indicates 23% and is lowering.

- Pre RAS alignment verification is being performed per EOP Supplement 42 "Pre and Post RAS Actions"

- Only ONE Containment Spray Pump is available and is operating.

- Actions were taken per EOP Supplement 42 to secure one HPSI pump and to CLOSE the Containment Spray Header isolation valve CV-3001.

These actions were taken to ensure that when the Recirculation Actuation signal (RAS) does occur...

a. the operating Containment Spray Pump will not be in a runout condition.

b. the operating HPSI pump will not be above its design discharge pressure rating.

c. the operating Containment Spray Pump will not be above its design discharge pressure rating.

d. the operating HPSI pump will not be in a runout condition.

QUESTION: 072 (1.00)

Refer to attached drawing. When transferring water from the Spent Fuel Pool (SFP) to the SIRW Tank using only P-82, Spent Fuel Pool Recirc Booster Pump, MV-SFP113 (T-50 to the SFP) is closed and MV-SFP127 (T-50 to the SIRW Tank) is opened.

When restoring the Spent Fuel Pool Cooling System to its normal lineup and P-82 is shut off, what concern is addressed by closing MV-SFP127 PRIOR to re-opening MV- SFP113?

a. SFP overfill due to backflow from the SIRW tank.

b. SFP low level due to siphoning action.

c. A high level in the SIRW tank due to unplanned transfer from SFP.

d. Elevated temperature of SIRW tank water due to unplanned transfer from SFP.

SENIOR REACTOR OPERATOR Page 48 of 71 QUESTION: 073 (1.00)

Given the following plant conditions:

- Plant is at 80% power and was performing a power escalation to full power when EK-1364, "GASEOUS MONITORING HI RADIATION" annunciated.

- It is determined that this alarm is due to RIA-0631, Condenser Off- Gas Monitor in an alarm condition.

- PCS total gas activity is 0.18 µCi/cc.

- Off Gas flow is 2 cfm

- At 0610 RIA-0631 indicated 1.00 E4 cpm

- At 0710 RIA-0631 indicates 2.00 E4 cpm

- "B" Steam Generator is the affected generator.

What actions should be taken to address the above plant conditions?

a. Trip the reactor and carry out the Immediate Actions of EOP-1.0, "Standard Post-Trip Actions"

b. Plant management must evaluate the need to perform a controlled Plant shutdown per GOP-8, "Power Reduction and Plant Shutdown".

c. Place the Plant in Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per ONP-23.2, "Steam Generator Tube Leak", Step 4.2.

d. Place the Plant in Mode 3 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per ONP-23.2, "Steam Generator Tube Leak", Step 4.2.

SENIOR REACTOR OPERATOR Page 49 of 71 QUESTION: 074 (1.00)

With the plant at full power, a Steam Generator tube leak occurs. With the Offgas flow rate at 2 cubic feet per minute, a reading is obtained from RIA-0631, Off-Gas Monitor that confirms the tube leak.

(Assume the Steam Generator tube leak rate has remained stable.)

If the Offgas flow rate were raised to 7 cubic feet per minute, the RIA-0631 indication would be...

a. unreliable since the detector is overranged.

b. higher due to the higher offgas flow rate.

c. lower due to the higher offgas flow rate.

d. unaffected.

QUESTION: 075 (1.00)

Given the following conditions and the provided references, as needed:

- Battery Chargers #1 and #2 are in service.

- Battery Charger #3 is inoperable and is to be tagged out.

The following sequence of events occur:

- Breaker 52-285 (Station Battery Charger #3) is opened.

- Breaker 72-15 (Charger #1) is mistakenly opened.

Which of the following additional breaker trips will result in a reactor trip?

a. 72-10

b. 72-18

c. 72-36

d. 72-37

SENIOR REACTOR OPERATOR Page 50 of 71 QUESTION: 076 (1.00)

Following a Loss of Coolant Accident, the reactor was tripped and Safety Injection initiated.

- 2400 VAC Bus 1D is being powered by the Safeguards Transformer.

- 2400 VAC Bus 1C has had a load shed.

- EDG 1-1 is running with normal voltage.

- Breaker 152-107, EDG 1-1 output breaker will NOT close.

What is the resulting effect on the DBA load sequencers?

a. Left channel DBA sequencer IS operating, right channel DBA sequencer IS operating.

b. Left channel DBA sequencer is NOT operating, right channel DBA sequencer IS operating.

c. Left channel DBA sequencer IS operating, right channel DBA sequencer is NOT operating.

d. Left channel DBA sequencer IS NOT operating, right channel DBA sequencer IS NOT operating.

QUESTION: 077 (1.00)

Many Process Liquid Monitors have a HIGH alarm and a LOW alarm setpoint. SOP-37, "Process Liquid Monitor System" Plant Requirements prescribes how these alarm setpoints are to be set.

Which one of the following explains how the LOW level alarm setpoints are to be set?

a. Below background so as to act as a circuit failure alarm for the monitor.

b. Above background so as to act as a circuit failure alarm for the monitor.

c. Below background so as to prevent overranging of the monitor.

d. Above background so as to prevent overranging of the monitor..

SENIOR REACTOR OPERATOR Page 51 of 71 QUESTION: 078 (1.00)

Which of the following describes the power supplies for the Service Water Pumps?

P-7A P-7B P-7C

a. Bus 1D Bus 1D Bus 1C

b. Bus 1D Bus 1C Bus 1D

c. Bus 1C Bus 1C Bus 1D

d. Bus 1C Bus 1D Bus 1C QUESTION: 079 (1.00)

Given the following conditions:

- The actions of ONP-7.1, "Loss of Instrument Air" have been implemented.

- Instrument air is being provided from Feedwater Purity Air system, using C-903A air compressor. System header pressure is being maintained at the normal pressure.

- Subsequently, a Large Break LOCA inside containment occurs, with all equipment responding per design.

What is the impact , if any, of the air system cross connection for the current plant conditions, and what action, if any, is required?

a. Since C-903A is now operating with reduced cooling flow and elevated temperatures, an Auxiliary Operator must locally raise cooling flow per SOP-19, "Instrument Air System".

b. C-903A continues to operate normally and provide plant instrument air header pressure, and no additional action is required.

c. Since C-903A has tripped due to loss of power; operators will have to implement an EOP Supplement for repowering and restarting C-903A.

d. C-903A is running, but must be manually tripped since it has NO cooling water.

Operators will have to align High Pressure Air System to supply the Instrument Air System.

SENIOR REACTOR OPERATOR Page 52 of 71 QUESTION: 080 (1.00)

The plant is in a heatup from MODE 4 to MODE 3 and drawing a bubble in the Pressurizer.

What is the pressure requirement for the Quench Tank?

a. less than 10 psig.

b. greater than 10 psig.

c. less than 25 psig.

d. greater than 25 psig.

QUESTION: 081 (1.00)

SOP-16, "Component Cooling Water System," contains a precaution that relates to operation of the following valves:

- CV-0937 and CV-0938, Shutdown Cooling Heat Exchanger CCW Inlet Valves.

Which one of the following describes the applicability of this precaution (including the concern it addresses), AND the action required to satisfy the precaution?

a. If only ONE CCW pump is operating, manually start a second CCW pump. This action prevents auto starting of the STANDBY CCW pump and is done prior to OPENING the valves.

b. If only ONE Service Water pump is operating, manually start a second Service Water pump. This action prevents auto starting of the STANDBY Service Water pump and is done prior to OPENING the valves.

c. If more than one CCW pump is operating, shut off one CCW pump prior to CLOSING the valves. This action ensures the valves operate smoothly, due to lowered system flow.

d. If more than Service Water pump is operating, shut off one Service Water pump prior to CLOSING the valves. This action prevents overcooling of the CCW System, since Service Water system flow has been reduced.

SENIOR REACTOR OPERATOR Page 53 of 71 QUESTION: 082 (1.00)

Which one of the following describes the operation of the containment Iodine Removal Fan units (V-940A, V-940B) and associated charcoal filters?

a. Automatically start on a Safety Injection Signal (SIS) to remove I- 131 generated during a Loss of Coolant Accident (LOCA).

b. Manually started during a normal Plant shutdown to remove I-131 for containment habitability.

c. Manually started during a normal Plant startup to minimize potential I-131 release to the environment.

d. Automatically start on a Containment High Pressure (CHP) to assist Containment Spray System in removing I-131 from containment.

QUESTION: 083 (1.00)

Given the following plant conditions:

- From full power, a Large Break LOCA occurred.

- Containment hydrogen concentration is at 3%.

Which one of the following actions should be taken to address these conditions?

a. Re-start all of the Containment Air Cooling "B" fans to ensure adequate mixing of Containment atmosphere.

b. Initiate a containment purge to reduce hydrogen below 1%, thereby minimizing the potential for a hydrogen burn .

c. Operate at least one of the Hydrogen Recombiners, thereby minimizing the potential for a hydrogen burn.

d. Energize Motor Control Center 9 to energize equipment for adequate mixing of Containment atmosphere.

SENIOR REACTOR OPERATOR Page 54 of 71 QUESTION: 084 (1.00)

During a refueling outage, the operator using the Spent Fuel Handling Machine is lifting a fuel bundle for placement in the inspection elevator. After the bundle has been lifted approximately 20" the operator notes the following:

- Hoist upward motion has automatically stopped.

- The CRT screen for the Spent Fuel Handling Machine is displaying "Fuel Overload" in a red box.

- Hoist Load Readout indicates 1712 lbs.

The operator attempts to lower the bundle back into its storage rack, and is successful in doing so. What is the correct assessment of the Spent Fuel Handling Machine operation?

a. The Hoist Emergency Up Limit functioned per design.

b. The Hoist Underload interlock should have prevented lowering the bundle.

c. The Hoist Overload interlock should have prevented any bundle movement.

d. The Hoist Overload interlock functioned per design.

SENIOR REACTOR OPERATOR Page 55 of 71 QUESTION: 085 (1.00)

Given the following conditions:

- The plant is in MODE 3 following a reactor trip from 100% power.

- PCS temperature is being controlled with the Turbine Bypass Valve in AUTO

- The Atmospheric Steam Dumps are closed with the control room C-02 panel Steam Dump Controller, HIC-0780A in AUTO

- The Average Temperature Display Select Switch is in the LOOP 2 position Which of the following describes the effect of a loss of the Tave signal from TYT- 0200 (e.g.,

signal failed LOW) on the plant. (Assume NO operator action has been taken.)

a. The only means of PCS heat removal with the secondary plant is via the Main Steam Code Safety valves.

b. The Turbine Bypass Valve fails closed and will NOT open until the Average Temperature Display Select Switch is placed in LOOP 1 position.

c. The Turbine Bypass Valve fails closed and will NOT open. The ADVs will open on a quick open signal.

d. The TBV will modulate open/closed to maintain S/G pressures at setpoint. The ADVs will NOT modulate open.

QUESTION: 086 (1.00)

Assume the plant is on Shutdown Cooling when a complete loss of Instrument Air occurs.

Which of the following describes the effect on the Shutdown Cooling System and on the Primary Coolant System (PCS)?

a. Since CV-3006, SDC Hx Bypass, fails CLOSED, the PCS will begin to heat up.

b. Since CV-3025, SDC Hx Outlet, fails CLOSED, the PCS will begin to heat up.

c. Since CV-3006, SDC Hx Bypass, fails OPEN, there is a concern for PCS overcooling.

d. Since CV-3025, SDC Hx Outlet fails OPEN, there is a concern for PCS overcooling.

SENIOR REACTOR OPERATOR Page 56 of 71 QUESTION: 087 (1.00)

Which one of the following describes interlock features on the Personnel Air Lock and Escape Air Lock doors which are designed to ensure Containment integrity?

a. Personnel Air Lock doors cannot be opened at the same time as Escape Air Lock doors.

b. Both doors on the Personnel Air Lock and the Escape Air Lock close and lock on a Containment High Pressure (CHP) or Containment High Radiation (CHR) condition.

c. A timer ensures that the Personnel Air Lock and the Escape Air Lock inner and outer doors can be opened at the same time ONLY for a maximum of 30 seconds.

d. The inner door cannot be opened at the same time as the outer door for the Personnel Air Lock and also for the Escape Air Lock.

QUESTION: 088 (1.00)

During a Loss of Coolant Accident inside containment the operator notes that the Plant Process Computer (PPC) displayed value for containment pressure has changed color from MAGENTA to WHITE. How is this information obtained on the PPC and what is its significance?

a. Depress "URGNT" hardkey. Containment pressure is now LESS THAN the alarm level setpoint.

b. Depress "ALARM" hardkey. Containment pressure is now ABOVE the alarm level setpoint.

c. Depress "EVENT" hardkey. A Containment High Pressure (CHP) has just actuated.

d. Depress "UPDATE" hardkey. Criteria for resetting Containment High Pressure are now met.

SENIOR REACTOR OPERATOR Page 57 of 71 QUESTION: 089 (1.00)

Note Step 5.9 of the attached procedure excerpt from GOP-2. Reference will also be made to SOP-7, "Main Steam System".

To perform the operation of verifying that MSIV closure is UNBLOCKED, how are the above procedures to be implemented?

a. You must EXIT GOP-2, and go to SOP-7 to perform the unblocking.

b. You REMAIN in GOP-2 and refer to SOP-7 to perform the unblocking.

c. Unblocking is performed per GOP-2 only. Use of SOP-7 is NOT required.

d. Unblocking is performed per SOP-7 only. Use of GOP-2 is NOT required.

QUESTION: 090 (1.00)

During the performance of a system checklist the position of a valve is found OPEN when the valve is required to be CLOSED by the checklist. Which of the following describes the sequence of actions required to be taken by the operator?

a. 1. Record the valves current position on the checklist.

2. Continue and complete the checklist.

3. Inform the Shift Supervisor.

b. 1. Obtain the Shift Supervisors authorization to reposition the valve.

2. Reposition the valve CLOSED.

3. Record on the checklist the new position.

c. 1. Record the valves current position on the checklist.

2. Obtain the Shift Supervisors authorization to reposition the valve.

3. Position the valve CLOSED.

d. 1. Position the valve to the CLOSED position.

2. Record valves original position on the checklist.

3. Inform the Shift Supervisor.

SENIOR REACTOR OPERATOR Page 58 of 71 QUESTION: 091 (1.00)

During a plant startup, the following conditions exist:

- The crew has taken actions to change Plant mode to MODE 2.

- The NCO was directed to remove all channels of Zero Power Mode (ZPM)

Bypass from operation, but has NOT taken any action.

- All keyswitches for ZPM Bypass are still in the enabled position (fully clockwise).

- Reactor power has risen to 10E-3% on all available indications.

- "A" Steam Generator (S/G) level has lowered to 24%.

What are the consequences of the above plant conditions?

a. The Reactor has tripped since the ZPM Bypass was automatically removed when Reactor power reached 10E-4%.

b. The Reactor has NOT tripped, but it would trip if the ZPM Bypass enable keyswitches were operated to the disable position (fully counterclockwise).

c. The Reactor has NOT tripped, and it would NOT trip even if the ZPM enable keyswitches were operated to the disable position (fully counterclockwise).

d. The Reactor has tripped since the ZPM Bypass was automatically removed when "A" S/G level lowered to 30%.

SENIOR REACTOR OPERATOR Page 59 of 71 QUESTION: 092 (1.00)

Refer to the following list of valve operations:

1. Close discharge valve.

2. Close suction valve.

3. Open discharge valve.

4. Open suction valve.

Which of the following describes the required sequence of valve operations when tagging out and subsequently restoring to service of a centrifugal pump?

TAGOUT RESTORE

a. 1,2 then 4,3

b. 2,1 then 4,3

c. 1,2 then 3,4

d. 2,1 then 3,4 QUESTION: 093 (1.00)

During refueling operations, which one of the following Spent Fuel Pool water levels is the LOWEST level which allows irradiated fuel handling activities?

a. 6" below skimmers

b. 10" below skimmers

c. 14" below skimmers

d. 18" below skimmers

SENIOR REACTOR OPERATOR Page 60 of 71 QUESTION: 094 (1.00)

Which one of the following describes the process of inverse multiplication plotting (1/M plot) during a reactor critical approach?

a. After each rod withdrawal wait until the startup rate reduces to near "0" before obtaining 1/M count rates.

b. Any of the available neutron flux instruments listed on the plot form may be used for any interval.

c. To ensure consistency the same plot form must be used throughout the entire critical approach.

d. 1/M plot data is obtained by dividing the SUBSEQUENT neutron flux reading by the INITIAL neutron flux readings.

QUESTION: 095 (1.00)

During a plant emergency an operator receives a radiation exposure of 7 REM to the lenses of both eyes.

Regarding 10 CFR 20, "Standards for Protection Against Radiation" and Palisades administrative radiation control limits, which, if any, of these limits have been exceeded?

a. BOTH 10CFR20 AND plant admin. limits have been exceeded.

b. NEITHER of the exposure limits listed have been exceeded.

c. NEITHER of the exposure limits apply for the above situation.

d. Plant admin. limits have been exceeded, but NOT 10CFR20 limits.

SENIOR REACTOR OPERATOR Page 61 of 71 QUESTION: 096 (1.00)

All of the following are elements of the Palisades ALARA program EXCEPT:

a. dose estimating.

b. temporary shielding.

c. Hot Spot Program.

d. Consummables Control .

QUESTION: 097 (1.00)

A Waste Gas Decay Tank batch release is planned, but the Waste Gas Monitor, RE-1113 is INOPERABLE.

For this condition, ALL of the following are actions that would allow initiating the release EXCEPT:

a. Setup local portable monitoring equipment at release point.

b. Perform independent verification of the discharge flowpath lineup.

c. Obtain an additional sample of the tank contents.

d. Perform independent verification of the release rate calculations.

SENIOR REACTOR OPERATOR Page 62 of 71 QUESTION: 098 (1.00)

Refer to the attached excerpt (page 8 of 40) from EOP-8.0.

Which of the following describes the type of step 7.a is, and the sequence of when it can be performed?

a. Continuous Step - can be performed anytime during the event.

b. Concurrent Step - must be performed at the same time as Step 6.1.

c. Sequential Step - must be performed right after Step 6 is complete.

d. Non-Sequential Step - can be performed when stated conditions exist.

QUESTION: 099 (1.00)

During a plant startup, the following conditions exist:

- The Main Generator has just been synchronized to the grid.

- A problem with CV-1359, Non-critical Service Water Isolation, occurs such that CV-1359 is failed partially closed an undetermined amount.

- The following alarm has annunciated:

- EK-1165, NON CRITICAL SERV WATER LO PRESS

- No other alarms have annunciated.

Which one of the following actions is required?

a. Trip the Reactor within 10 seconds.

b. Trip the Reactor if Exciter Cooler Hi Temp alarm annunciates.

c. Trip the Main Turbine within 10 seconds.

d. Trip the Main Turbine if Exciter Cooler Hi Temp alarm annunciates.

SENIOR REACTOR OPERATOR Page 63 of 71 QUESTION: 100 (1.00)

Note the following two alarms:

EK-0552, DIESEL GENERATOR NUMBER 1-1 START SIGNAL BLOCKED EK-0742, PRESSURIZER HTR BUS GROUND/UNDERVOLTAGE For a Loss of all Offsite Power event, which one of the following describes the significance of the alarm condition which results in a direct effect on the HIGHER priority safety function?

a. D/G 1-1 will not start automatically, but can be started manually from the Control Room.

b. D/G 1-1 will not start automatically, and cannot be started manually from the Control Room.

c. ALL Pressurizer heaters are deenergized and will remain deenergized.

d. SOME Pressurizer heaters will regain power after a load sequencing.

(********** END OF EXAMINATION **********)

SENIOR REACTOR OPERATOR Page 64 of 71 ANSWER: 001 (1.00) ANSWER: 005 (1.00) ANSWER: 010 (1.00)

c. a. c.

REFERENCE:

M-209, sh.2 SOP-3, EOP-9.0, RC-1,2,3, TS 3.5.1, Action A, C TS , 2.0.3, rev 52 MVAE-AC-1 3.5.2, Action A ORM 3.2 SOP-3, Attachment 10, page NEWHIGH 2.4.6 NEW 13, rev 52 ARP-7, window 55 000055 2.4.6 ..(KAs) HIGH and 56, rev 64 Tech Spec 000022 2.4.30 ..(KAs)

Basis for 3.5.2, Background NEW ANSWER: 006 (1.00)

MEMORY c. ANSWER: 011 (1.00) 000011 A2.07 ..(KAs)

REFERENCE:

d.

EOP-4.0 Basis, page 278 of

REFERENCE:

310, rev 13 EOP Supplement Tech Spec 3.3.1 Basis for ANSWER: 002 (1.00) 26 Condition D, and Action D.1,

b. NEW D.2

REFERENCE:

HIGH NEW DBD-2.05, page 37 of 129, 000074 A2.04 ..(KAs) HIGH rev 3 References Supplied to NEW Candidate: LCO 3.3.1, page MEMORY ANSWER: 007 (1.00) 3.3.1-2 000029 A2.08 ..(KAs) d. 2.2.25 000033 ..(KAs)

REFERENCE:

ONP-25.2, step 19, rev 18 ANSWER: 003 (1.00) NEW ANSWER: 012 (1.00)

b. HIGH c.

REFERENCE:

2.1.30 000074 ..(KAs)

REFERENCE:

AP 4.14, rev 0 EOP 5.0 Basis for Step 17, NEW rev 13 HIGH ANSWER: 008 (1.00) NEW 2.1.2 000029 ..(KAs) b. MEMORY

REFERENCE:

000038 A2.15 ..(KAs)

Tech Spec 3.4.16 ANSWER: 004 (1.00) Applicability Basis

c. NEW ANSWER: 013 (1.00)

REFERENCE:

HIGH b.

EOP-3.0 EOP Supplements 2.1.28 000076 ..(KAs)

REFERENCE:

as mentioned in question ONP-2.3, 6.0.1, rev 12 NEW ARP-4, window 2, rev 56 MEMORY ANSWER: 009 (1.00) MODIFIED 000055 A2.06 ..(KAs) a. HIGH

REFERENCE:

000058 A2.03 ..(KAs)

AP 4.08, 5.3.4, a, rev 5 NEW HIGH 000007 A2.06 ..(KAs)

SENIOR REACTOR OPERATOR Page 65 of 71 ANSWER: 014 (1.00) ANSWER: 019 (1.00) ANSWER: 024 (1.00)

d. c. c.

REFERENCE:

EOP-9.0, Resource ARP-21, Rack D5, rev 48 EOP Supp 3 Basis, page 2, Assessment Tree I, rev 16 NEW rev 9 NEW HIGH BANK HIGH 012 A2.06 ..(KAs) HIGH 000065 2.4.6 ..(KAs) 2.4.18 ..(KAs)

ANSWER: 020 (1.00)

ANSWER: 015 (1.00) b. ANSWER: 025 (1.00)

d.

REFERENCE:

a.

REFERENCE:

SOP-24, 7.2.5, rev 36 ODCM

REFERENCE:

ONP 23.3 BANK HP 6.14 Operating NONE PROVIDED HIGH Requirements Manual BANK 000036 A2.02 ..(KAs) NEW HIGH MEMORY 2.4.40 ..(KAs) 029 A4.01 ..(KAs)

ANSWER: 016 (1.00)

b. ANSWER: 026 (1.00)

REFERENCE:

ANSWER: 021 (1.00) b.

COLR T.S. 3.1.6 SOP-6, 7.5 c.

REFERENCE:

BANK

REFERENCE:

SOP-6, 4.1.8, rev 20 TS MEMORY AP 4.28, 12.5.1, rev 0 3.1.4. Basis References Supplied to NEW ONP-5.1, rev 20 Candidate: TS 3.1.4, 3.1.5, MEMORY NEW 3.1.6 ( do NOT provide basis) 062 A4.01 ..(KAs) HIGHER 001 2.1.33 ..(KAs) 000005 K3.03 ..(KAs)

ANSWER: 022 (1.00)

ANSWER: 017 (1.00) a. ANSWER: 027 (1.00)

d.

REFERENCE:

a.

REFERENCE:

GOP-3, 4.1, Attachment 1,

REFERENCE:

LCO 3.7.6 Basis 1.21, rev 18 Tech. Spec. ARP-5, window 1, rev 65 NEW Table 1.1-1 FSAR 14.7.2.1 MEMORY NEW NEW 061 2.2.25 ..(KAs) MEMORY HIGHER 2.1.22 ..(KAs) 000015/17 K2.10 ..(KAs)

ANSWER: 018 (1.00)

b. ANSWER: 023 (1.00) ANSWER: 028 (1.00)

REFERENCE:

a. a.

ONP-11.2

REFERENCE:

MODIFIED COP-31, 3.1, 5.2, Attachment LP-RHAA EOP-8.0 Basis, HIGH 2, rev 19 page 48, rev 10 072 A3.01 ..(KAs) NEW NEW MEMORY HIGHER 2.3.6 ..(KAs)

CE/A13 A1.02 ..(KAs)

SENIOR REACTOR OPERATOR Page 66 of 71 ANSWER: 029 (1.00) ANSWER: 034 (1.00) ANSWER: 039 (1.00)

c. b. b.

REFERENCE:

DBD 1.04, page 49 ONP-14, 2.0, and Table 4.3-1 ONP-11.1, rev 17 SOP-2A, 7.5.2.b and BANK NEW , 2.2.3, rev 52 HIGH HIGH BANK 000051 A2.02 ..(KA's) 000076 A2.02 ..(KA's)

HIGHER 000024 K2.01 ..(KAs)

ANSWER: 035 (1.00) ANSWER: 040 (1.00)

d. d.

ANSWER: 030 (1.00)

REFERENCE:

a. ONP-24.5, 2.0, and 4.3, rev DBD-2.06

REFERENCE:

19 NEW DBD 1.01, 3.2.1 BANK HIGH NEW MEMORY 000001 A1.07 ..(KA's)

HIGHER 000057 K3.01 ..(KA's) 000026 K3.02 ..(KAs)

ANSWER: 041 (1.00)

ANSWER: 036 (1.00) d.

ANSWER: 031 (1.00) c.

REFERENCE:

b.

REFERENCE:

SOP-6, 7.8.b, rev 20 ARP-5,

REFERENCE:

DBD 1.02, page 22 of 152 windows 16, 17, rev 64 EOP Supplement 1 NEW NEW Steam Tables HIGH HIGH BANK References Supplied to 000003 K2.05 ..(KA's)

HIGH Candidate: Attached drawing 000027 K1.01 ..(KAs) of CV- 0823, CV-0826 000062 K3.02 ..(KA's) ANSWER: 042 (1.00) c.

ANSWER: 032 (1.00)

REFERENCE:

d. ANSWER: 037 (1.00) Steam Tables, Mollier

REFERENCE:

c. diagram EOP-1.0, BOP Operator Aid,

REFERENCE:

BANK rev 12 SOP-24, 7.7.7, rev 36 HIGH BANK M-218 000008 K1.01 ..(KA's)

MEMORY DBD 1.06 000040 A1.01 ..(KAs) NEW HIGH ANSWER: 043 (1.00) 000067 A2.06 ..(KA's) d.

ANSWER: 033 (1.00)

REFERENCE:

d. ANSWER: 038 (1.00) EOP-4.0 Basis, page 4, rev

REFERENCE:

b. 13 EOP-6.0, Basis, step 16

REFERENCE:

Steam Tables BANK EOP-1.0, Primary Operator BANK LOW Aid HIGH CE/A11 K2.02 ..(KAs) NEW 000009 K1.01 ..(KA's)

HIGH 000069 K1.01 ..(KA's)

SENIOR REACTOR OPERATOR Page 67 of 71 ANSWER: 044 (1.00) ANSWER: 049 (1.00) ANSWER: 054 (1.00)

b. a. a.

REFERENCE:

ONP-17 COP-1, Att. 10, rev 50 FSAR EOP-1.0 Basis, page 4 of NEW 14.21, rev 23 103, rev 10 MEMORY NEW MODIFIED 000025 A1.09 ..(KAs) MEMORY MEMORY 000060 K1.01 ..(KAs) 001 K3.02 ..(KA's)

ANSWER: 045 (1.00)

b. ANSWER: 050 (1.00) ANSWER: 055 (1.00)

REFERENCE:

c. d.

SOP-35, 7.1.2, rev 14

REFERENCE:

ARP-21, A-2, rev 48 SOP-24, 4.3.3, rev 36 FSAR 14.7.1.1 BANK MODIFIED MODIFIED MEMORY HIGHER HIGH 000032 K2.01 ..(KAs) 000061 K2.01 ..(KAs) 003 K5.05 ..(KA's)

ANSWER: 046 (1.00) ANSWER: 051 (1.00) ANSWER: 056 (1.00)

c. d. b.

REFERENCE:

ARP-8, window 65, rev 62 None Provided EOP Supplement 37, 1.0.7 EOP-5 Basis, page 28, rev 10 NEW NOTE NEW MEMORY NEW HIGH CE/E09 K1.02 ..(KAs) HIGH 000037 K3.10 ..(KAs) 004 K1.17 ..(KA's)

ANSWER: 052 (1.00) a.

ANSWER: 047 (1.00)

REFERENCE:

ANSWER: 057 (1.00)

d. DBD 1.04 a.

REFERENCE:

BANK

REFERENCE:

Provide EOP Supp 41 HIGH EOP-4.0 Basis, p. 75 BANK 000028 K2.03 ..(KA's) NEW HIGH HIGH References Supplied to 013 A1.10 ..(KA's)

Candidate: EOP Supplement ANSWER: 053 (1.00) 41 c.

000054 A2.05 ..(KAs)

REFERENCE:

ANSWER: 058 (1.00)

ARP-7, windows 57 and 58, c.

rev 64 SOP-3, 5.2.3, rev 52

REFERENCE:

ANSWER: 048 (1.00) DBD 2.01, page 67, rev 7 SOP-35, Attachment 2, page

c. BANK 2 of 8, rev 14

REFERENCE:

HIGH BANK COP-31, 33 CE/A16 K2.02 ..(KAs) MEMORY NEW 015 K2.01 ..(KA's)

MEMORY 000059 K1.01 ..(KAs)

SENIOR REACTOR OPERATOR Page 68 of 71 ANSWER: 059 (1.00) ANSWER: 064 (1.00) ANSWER: 069 (1.00)

d. d. d.

REFERENCE:

FSAR 7.6 SOP-34, 4.2, rev SOP-24, 7.2.2, rev 36 ARP-4, windows 63, 64, rev 16 NEW 56 NEW MEMORY NEW MEMORY 068 K1.02 ..(KAs) MEMORY 017 K1.01 ..(KAs) 011 K6.03 ..(KAs)

ANSWER: 065 (1.00)

ANSWER: 060 (1.00) b. ANSWER: 070 (1.00)

a.

REFERENCE:

a.

REFERENCE:

SOP-18A, Source

REFERENCE:

P& ID E-1, sh. 1, Rev BS Documents section, rev 33 DBD 2.06, 3.3.4.1, rev 3 BANK FSAR 11.3 NEW MEMORY BANK MEMORY 022 K2.01 ..(KAs) MEMORY 014 A1.01 ..(KAs) 071 K5.04 ..(KAs)

ANSWER: 061 (1.00) ANSWER: 071 (1.00)

a. ANSWER: 066 (1.00) a.

REFERENCE:

a.

REFERENCE:

ARP-1, window 55 and 60,

REFERENCE:

EOP Supp 42 Basis rev 52 ONP-3, 1.0, 2.0, rev SOP-39, 4.0.b, and 7.3.2.b, BANK 18 rev 11 HIGH MODIFIED BANK 026 K4.08 ..(KAs)

MEMORY MEMORY 056 K1.03 ..(KAs) 072 K1.03 ..(KAs)

ANSWER: 072 (1.00) a.

ANSWER: 062 (1.00) ANSWER: 067 (1.00)

REFERENCE:

d. c. SOP-27 P&ID M-221, sh.2

REFERENCE:

MODIFIED ARP-5, window 61, rev 65 T.S. 3.5.1 References Supplied to BANK BANK Candidate: M-221, sh.2 HIGH MEMORY 033 K1.05 ..(KAs) 059 K1.04 ..(KAs) 006 2.1.33 ..(KAs)

ANSWER: 073 (1.00)

ANSWER: 063 (1.00) ANSWER: 068 (1.00) d.

a. b.

REFERENCE:

ONP-23.2 DBD 1.03, page 50 of 124 ONP-18, 1.0.a, 2.0, 4.2.1.c.1, NEW P&ID (Logic Diagram) E-17, rev 16 ARP-4, window 53, rev HIGH sh. 21, 21A 56 ARP-21, C-1, rev 48 References Supplied to NEW MODIFIED Candidate: ONP-23.2 MEMORY HIGH (excerpt) 061 K4.12 ..(KAs) 010 K1.02 ..(KAs) 035 A2.01 ..(KAs)

SENIOR REACTOR OPERATOR Page 69 of 71 ANSWER: 074 (1.00) ANSWER: 079 (1.00) ANSWER: 084 (1.00)

c. c. d.

REFERENCE:

ONP-23.2 SOER-93-1, EOP Supplement 25 SOP-28, Att. 7, Section 8.0, Supplement KLO 9 ONP-7.1 rev 33 FSAR 9.11-20, rev. 23 BANK NEW SOP-28, Note prior to 7.2.5 HIGH HIGH NEW 055 K1.06 ..(KAs) 079 A2.01 ..(KAs) HIGH 034 A3.02 ..(KAs)

ANSWER: 075 (1.00) ANSWER: 080 (1.00)

b. a. ANSWER: 085 (1.00)

REFERENCE:

d.

ONP-2.3, Att. 1, rev 12 SOP-1, 4.5.3, rev 51

REFERENCE:

BANK BANK ONP-13, rev 7 DBD 1.09, HIGH MEMORY 3.2.2.2 References Supplied to 007 K5.02 ..(KAs) BANK Candidate: ONP-2.3, Att. 1 HIGH 063 K3.02 ..(KAs) 041 K6.03 ..(KAs)

ANSWER: 081 (1.00) a.

ANSWER: 076 (1.00)

REFERENCE:

ANSWER: 086 (1.00)

d. SOP-16, 5.1.3, rev 23 a.

REFERENCE:

MODIFIED

REFERENCE:

DBD 5.05, Drawing E-17, sh. HIGH ONP-7.1, 4.1, rev 13 M-204, 4 008 2.1.32 ..(KAs) sh. 1 BANK NEW HIGH HIGH References Supplied to ANSWER: 082 (1.00) 078 K3.02 ..(KAs)

Candidate: Provide E-17, sh. b.

4

REFERENCE:

064 K3.03 ..(KAs) GOP-8, 2.6.a, rev 19 GOP-5, ANSWER: 087 (1.00) 1.2, rev 28 SOP-24, 7.2.7, d.

rev 36 FSAR 9.8, page

REFERENCE:

ANSWER: 077 (1.00) 9.8-13, rev 23 DBD 5.8.6.2.1, rev 23

a. NEW BANK

REFERENCE:

MEMORY MEMORY SOP-37, 4.0.c, rev 15 027 K5.01 ..(KAs) 103 K4.04 ..(KAs)

NEW HIGH 073 2.1.32 ..(KAs) ANSWER: 083 (1.00) ANSWER: 088 (1.00)

c. a.

ANSWER: 078 (1.00)

REFERENCE:

b. EOP-4.0, step 58 and basis PPC Users Manual Operator

REFERENCE:

NEW Aid 178 P&ID E-1, sh. 1, rev BS HIGH NEW BANK 028 A2.03 ..(KAs) MEMORY MEMORY 2.1.19 ..(KAs) 075 K2.03 ..(KAs)

SENIOR REACTOR OPERATOR Page 70 of 71 ANSWER: 089 (1.00) ANSWER: 093 (1.00) ANSWER: 097 (1.00)

c. b. a.

REFERENCE:

GOP-2, Att.1, step 5.9, rev 24 GOP-11, Att. 2, 1.1.8, rev 35 SOP-18A, 7.5.c, rev 33 Admin 10.51, Att. 4, 5, and BANK NEW 19, rev 13 MEMORY MEMORY NEW 2.2.27 ..(KAs) 2.3.11 ..(KAs)

MEMORY References Supplied to ANSWER: 094 (1.00)

Candidate: GOP-2, step 5.9 a. ANSWER: 098 (1.00) on page 15, rev. 24

REFERENCE:

d.

2.1.23 ..(KAs) GOP-3, 5.1.1.b, Attachment

REFERENCE:

2, rev 18 AP 4.06, page 5 of 27, rev 12 NEW NEW ANSWER: 090 (1.00) MEMORY MEMORY

b. 2.2.34 ..(KAs) References Supplied to

REFERENCE:

Candidate: EOP-8.0, page 8 AP 4.02, 5.3.1.a, 7.3, rev 18 (attached)

BANK ANSWER: 095 (1.00) 2.4.14 ..(KAs)

MEMORY d.

2.1.29 ..(KAs)

REFERENCE:

10 CFR 20.1201, item ANSWER: 099 (1.00)

(a)(2)(i) - 15R AP 7.04, Att. 1, d.

ANSWER: 091 (1.00) page 2, rev 19 - 6R

REFERENCE:

a. NEW ONP 6.1, rev 11

REFERENCE:

MEMORY NEW SOP-36, 7.2.2.a, rev 8 2.3.1 ..(KAs) HIGH BANK 2.4.24 ..(KAs)

HIGH 2.2.2 ..(KAs) ANSWER: 096 (1.00)

d. ANSWER: 100 (1.00)

REFERENCE:

b.

ANSWER: 092 (1.00) AP 7.02, 6.2

REFERENCE:

a. NEW ARP-3, window 52, rev 58

REFERENCE:

MEMORY ARP-4, window 42, rev 56 AP 4.10, Attachment 1, 3.8.a, 2.3.2 ..(KAs) P&ID E-17, sh. 12 b, rev 13 NEW NEW HIGH MEMORY 2.4.45 ..(KAs) 2.2.13 ..(KAs)

(********** END OF EXAMINATION **********)

SENIOR REACTOR OPERATOR Page 71 of 71 ANSWER KEY MULTIPLE CHOICE 001 c 021 c 041 d 061 a 081 a 002 b 022 a 042 c 062 d 082 b 003 b 023 a 043 d 063 a 083 c 004 c 024 c 044 b 064 d 084 d 005 a 025 a 045 b 065 b 085 d 006 c 026 b 046 c 066 a 086 a 007 d 027 a 047 d 067 c 087 d 008 b 028 a 048 c 068 b 088 a 009 a 029 c 049 a 069 d 089 c 010 c 030 a 050 c 070 a 090 b 011 d 031 b 051 d 071 a 091 a 012 c 032 d 052 a 072 a 092 a 013 b 033 d 053 c 073 d 093 b 014 d 034 b 054 a 074 c 094 a 015 d 035 d 055 d 075 b 095 d 016 b 036 c 056 b 076 d 096 d 017 d 037 c 057 a 077 a 097 a 018 b 038 b 058 c 078 b 098 d 019 c 039 b 059 d 079 c 099 d 020 b 040 d 060 a 080 a 100 b

(********** END OF EXAMINATION **********)