ML082130329

U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination

ML082130329
Person / Time
Site:	Byron
Issue date:	08/05/2008
From:	Operations Branch III
To:
References
50-454/08-301, 50-455/08-301
Download: ML082130329 (86)

Text

U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: Facility/Unit: Byron Nuclear Station U1/U2 Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant=s Signature Results RO/SRO-Only/Total Examination Values 75 / 25 / 100 Points Applicant=s Scores / / Points Applicant=s Grade / / Percent

U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: Facility/Unit: Byron Nuclear Station U1/U2 Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant=s Signature Results Examination Value 75 Points Applicant=s Score Points Applicant=s Grade Percent

APPENDIX E POLICIES AND GUIDELINES FOR TAKING NRC EXAMINATIONS Each examinee shall be briefed on the policies and guidelines applicable to the examination category (written, operating, walk-through, and/or simulator test) being administered.

The examinees may be briefed individually or as a group. Facility licensees are encouraged to distribute a copy of this appendix to every examinee before the examination begins.

All items apply to both initial and requalification examinations, except as noted.

Part A: 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 to ask 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 an overall grade of 80.00 percent or greater, with 70.00 percent or greater on the SRO-only items, if applicable. If you only take the SRO portion of the exam (as a retake or with an upgrade waiver of the RO exam), you must achieve an overall grade of 80.00 percent or better to pass. SRO-upgrade applicants who do take the RO portion of the exam and score below 80.00 percent on that part of the exam can still pass overall, but may require remediation. 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the RO exam; 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the 25-question, SRO-only exam; and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the combined RO/SRO exam. Notify the proctor if you need more time.

4. You may bring pens, pencils, and calculators into the examination room; however, programable memories must be erased. Use dark pencil to mark your answer on the answer sheets.

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. You may write on the back side of the pages. Do not use ink. If you are recording your answers on a machine-gradable form that offers more than four answer choices (e.g., Aa@ through Ae@), be careful to mark the correct column.

7. If you have any QUESTIONs concerning the intent or the initial conditions of a QUESTION, do not hesitate to ask them before answering the QUESTION. Note that QUESTIONs asked during the examination are taken into consideration during the grading process and when reviewing applicant appeals. Ask QUESTIONs of the NRC examiner or the designated facility instructor only. A dictionary is available if you need it.

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. Similarly, you should assume that no operator actions have been taken, unless the stem of the QUESTION or the answer choices specifically state otherwise. 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 that includes the examination cover sheet and answer sheets, 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. Leave all other materials at your examination station.

10. After turning 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?

REACTOR OPERATOR Page 1 of 50 QUESTION: 001 (1.00)

The following Unit 1 plant conditions exist:

- Unit 1 has experienced a reactor trip and SI.

- Containment pressure is 27 psig.

- RCS pressure is 300 psig.

- Seven of Eight SX Cooling Tower Fans are running in High Speed.

- 0A fan will NOT start in High Speed.

Which ONE of the following actions is required per 1BEP-0, Reactor Trip or Safety Injection, when aligning the SX Cooling Towers?

a. OPEN all EIGHT riser valves.

b. Restart 0A fan in Low Speed.

c. CLOSE all FOUR Hot Water Basin Bypass valves.

d. Ensure that ONLY the bypass valve associated with the non-running fan is CLOSED.

QUESTION: 002 (1.00)

Unit 1 was at 100% power with all systems normally aligned when annunciator 1-12-B2, PZR PORV OR SAF VLV OPEN, alarms. The following indications are current:

- Actual PZR pressure is 2100 psig and lowering

- Channel 1PT-455 indicates 2500 psig

- PZR level is 62% and rising

- PRT temperature, pressure and level are rising

- All PZR Safety Valve indicator lights are GREEN Action(s) to mitigate this transient is/are to . . .

a. close the PZR PORV block valve(s) for affected PORV(s).

b. manually trip the reactor and actuate SI.

c. verify insertion of control rods at 48 steps per minute.

d. manually trip the reactor, but DO NOT manually actuate SI.

REACTOR OPERATOR Page 2 of 50 QUESTION: 003 (1.00)

Given the following plant conditions for Unit 1:

- Indicated charging flow is 152 gpm.

- Letdown flow is 120 gpm.

- Total seal injection flow is 32 gpm.

- Total seal return flow is 12 gpm.

- Pressurizer level is stable.

Does a primary leak exist, and if so, what is the size of the leak?

There is . . .

a. NO primary leak.

b. a 20 gpm leak.

c. a 32 gpm leak.

d. a 52 gpm leak.

QUESTION: 004 (1.00)

Unit 1 is operating at 100% power with all systems normally aligned:

- 1A CV pump running

- PZR level at 60%

- Charging flow 132 gpm

- Seal injection flows 11 gpm per RCP

- 1CV121, Centrifugal Charging Pump Flow Control Valve, in MANUAL

- 1CV182, Charging Header Back Pressure Control Valve, at 52% demand

- 1A CV impeller begins to slowly degrade In order to maintain pressurizer level at 60% and RCP seal injection flows at 11 gpm, operators must . . .

a. throttle close 1CV121 AND 1CV182.

b. throttle open 1CV182 ONLY.

c. throttle close 1CV121 AND throttle open 1CV182.

d. throttle open 1CV121 AND maintain 1CV182 at 52% demand.

REACTOR OPERATOR Page 3 of 50 QUESTION: 005 (1.00)

Unit 1 is operating at 100% power with all systems normally aligned. During the shift, the following 1A RCP indications have been observed:

- Motor amps slowly rising

- Shaft vibrations slowly rising

- Seal inlet temperature 116°F, stable

- #1 seal outlet temperature 132°F, stable

- High range seal leakage recorder reads 4.2 gpm, stable

- #1 seal delta P greater than 400 psid (pegged high)

- 1A RCP loop flow slowly lowering These cumulative conditions indicate the . . .

a. #1 seal is failing.

b. #2 seal is failing.

c. shaft has sheared.

d. thrust bearing is failing.

QUESTION: 006 (1.00)

Based upon time in core life, which of the following statements describes the transient expected during an ATWS?

a. At BOL conditions, the MORE negative doppler temperature coefficient results in higher peak power.

b. At BOL conditions, the LESS negative moderator temperature coefficient allows higher peak temperatures and pressures.

c. At EOL conditions, the LESS negative doppler temperature coefficient allows higher peak temperatures and pressures.

d. At EOL conditions, the LESS negative moderator temperature coefficient results in higher peak power.

REACTOR OPERATOR Page 4 of 50 QUESTION: 007 (1.00)

The CAUTION before Step 1 of BEP ES-3.1, Post SGTR Cooldown Using Backfill, warns that the RCP in the ruptured loop should NOT be started first. Which of the following explains the reason for this CAUTION?

Starting the ruptured loops RCP first could cause . . .

a. an internal pressure surge resulting in further failure of damaged tubes.

b. a slug of cold water to pass through the core and cause a return to criticality.

c. a slug of unborated water to pass through the core and cause a return to criticality.

d. a rapid reactor vessel cooldown that would challenge the Integrity critical safety function.

QUESTION: 008 (1.00)

For a steam line break inside containment, all of the following will indicate a RISE in value on the Narrow Range SPDS iconic, EXCEPT . . .

a. Subcooling Margin.

b. Net Charging.

c. Containment Air Temperature.

d. RVLIS.

REACTOR OPERATOR Page 5 of 50 QUESTION: 009 (1.00)

Unit 1 was operating at 100% power preparing to enter coast down for an end of cycle refueling outage. The 1A FW pump was OOS with all other systems normally aligned.

An inadvertent FW isolation occurred. The following conditions are present 60 seconds after the FW isolation:

- PR NIs are 0%

- IR SUR is -0.5 dpm

- MSIVs are OPEN

- Annunciator 1-18-C3 EMER TRIP HDR PRESS LOW TRIP is NOT lit

- SG 1A -1D Train B AF flow is greater than 160 gpm per SG

- SG WR levels are 25%, lowering

- RCS temperature is 480°F, lowering These cumulative conditions indicate . . .

a. AF is NOT flowing to the SGs. RCS pressure will exceed the accident analysis limits during the event.

b. BOTH reactor and turbine are tripped. RCS pressure will NOT exceed the accident analysis limits during the event.

c. BOTH reactor and turbine are NOT tripped. RCS pressure will exceed the accident analysis limits during the event.

d. the reactor is tripped, but the turbine is NOT tripped. RCS pressure will NOT exceed the accident analysis limits during the event.

REACTOR OPERATOR Page 6 of 50 QUESTION: 010 (1.00)

The reactor has tripped from 100% power due to a Loss of All AC Power, the EDGs failed to reenergize the 4160 volt ESF buses and the following plant indications are noted.

- RCS pressure is 1575 psig and lowering

- RCS hot leg temperatures indicate 604°F and rising

- Pressurizer level indicates 90% and rising

- A steam bubble exists in the reactor vessel Which ONE of the following is most likely to cause these indications and what actions should be taken to mitigate the situation?

a. A pressurizer PORV has failed in the open position. Manually or locally close the PORV block valve when power is available.

b. A Steam Generator PORV has failed open. An operator needs to be dispatched to locally isolate the failed open PORV.

c. Lack of an adequate heat sink has caused a bubble to form in the reactor vessel.

Reduce Steam Generator pressures to lower RCS temperatures to reduce vessel bubble.

d. The steam dump controller has failed with all steam dumps fully closed. Send operators to locally throttle the steam dumps to control the plant heatup rate.

REACTOR OPERATOR Page 7 of 50 QUESTION: 011 (1.00)

After a Loss of All AC Power, the Operations crew is in procedure 1BCA-0.1, Loss of All AC Power Recovery Without SI Required. Verification of natural circulation cooling can be confirmed by which of the following?

a. RCS subcooling is adequate, SG pressure dropping, RCS hot leg temperature rising.

b. SG pressure dropping, RCS hot leg temperature stable, RCS cold leg temperature rising.

c. RCS subcooling is adequate, RCS hot leg temperature dropping, Average of ten highest core exit thermocouples dropping.

d. SG pressure rising, RCS hot leg temperature dropping, RCS cold leg temperature rising.

QUESTION: 012 (1.00)

Why is LOCAL operator action required for a loss of Instrument Bus 214 that resulted in a reactor trip and safety injection?

a. The flow from 2B Auxiliary Feedwater train to all Steam Generators will be excessive.

b. The flow from 2B Auxiliary Feedwater train to all Steam Generators will be too low.

c. 2A Auxiliary Feedwater Pump will need to be LOCALLY started.

d. 2B Auxiliary Feedwater Pump will need to be LOCALLY started.

REACTOR OPERATOR Page 8 of 50 Question #13 Has Been Deleted From This Examination QUESTION: 013 (1.00)

Previously, 125 VDC Bus 211 was crosstied to Bus 111 due to equipment problems with Bus 111 Battery and Charger. Bus 111 Battery and Charger are Out of Service.

Presently,

- U-1 is in MODE 3.

- U-2 is in MODE 1.

Bus 111 conditions are:

- Crosstie loading due to the loading on Bus 111 is 183 Amps.

- Voltage on Bus 111 is 121 VDC.

- Then, a ground of 50 volts is detected on Bus 111.

Based upon the above conditions, which one of the following actions would be CORRECT?

a. Parameters on Bus 111 are normal and within limits. No action is necessary.

b. Enter into BOP, DC-15, DC Ground Isolation, due to an unexpected ground detected on Bus 111.

c. Shed non-essential loads from Bus 111 to lower Amperage to below 180 Amps to meet cross-tie loading restrictions.

d. Disconnect Bus 111 from Bus 211 in accordance with BOP DC-7, 125 VDC ESF Bus Crosstie/Restoration to ensure that the ground does not adversely affect loads on the operating unit.

REACTOR OPERATOR Page 9 of 50 QUESTION: 014 (1.00)

Given the following conditions:

- A unit 1 SX leak in the Aux building has been identified and isolated from the main control room.

- A clearance order is being prepared by an RO.

- A Safety Injection has just occurred on unit 1.

The leak will remain isolated if it is located in which of the following components?

a. Unit 1 Reactor Containment Fan Coolers

b. 1B Containment Chiller Condenser

c. 1A Emergency Diesel Generator

d. 1B Auxiliary Feedwater Pump QUESTION: 015 (1.00)

The operators are performing 1BOA SEC-4, Loss of Instrument Air, with all other conditions normal and the unit at 90% load.

If the reactor is NOT manually tripped as instrument air pressure continues to lower, then the initial automatic reactor trip is the result of...

a. Main Feedwater Regulating Valves FW510-540 failing CLOSED.

b. Letdown Orifice Isolation Valves CV8149A, B, and C failing CLOSED.

c. Main Feedwater Regulating Valves FW510-540 failing OPEN.

d. Centrifugal Charging Pumps flow control valve, CV121, failing OPEN.

REACTOR OPERATOR Page 10 of 50 QUESTION: 016 (1.00)

Given the following conditions:

- An inadvertent Safety Injection signal tripped the reactor.

- All equipment responded as required with the exception that 1CV8160, Letdown Line Inboard Containment Isolation Valve, failed to close.

- Pressurizer level fell to 25% and is now at 30% and rising.

- Pressurizer Pressure is 2310 psig and rising.

With the above conditions, the CVCS letdown line suddenly breaks just outside of the containment penetration and upstream of 1CV8152, Letdown Line Outboard Containment Isolation Valve.

Based upon the information above, what conditions would be expected following this event?

a. A leak will continue outside containment with NO valves closed to isolate it.

b. 1CV8149A, B & C, Orifice Isolation Valves, will immediately close on a Safety Injection Signal to isolate any leakage.

c. 1CV8149A, B & C, Orifice Isolation Valves, will immediately close on a Phase A Isolation Signal to isolate any leakage.

d. A leak would be present outside containment until 1CV8149A, B & C, Orifice Isolation Valves, fail CLOSED.

REACTOR OPERATOR Page 11 of 50 QUESTION: 017 (1.00)

During implementation of 1BCA-1.1, Loss of Emergency Coolant Recirculation, which of the following Unit 1 plant PARAMETERS/CONDITIONS are evaluated to determine the number of required Containment Spray pumps to operate?

1. RCFCs running in Accident Mode

2. Containment Pressure

3. RWST Level

a. 1, 2, and 3

b. 1 and 2 ONLY

c. 1 and 3 ONLY

d. 2 and 3 ONLY QUESTION: 018 (1.00)

The following Unit 1 plant conditions exist:

- Unit 1 has experienced a reactor trip

- A loss of all AF pumps has required an entry into 1BFR-H.1, Response to Loss of Secondary Heat Sink

- Containment pressure is 1 psig

- RCS temperature is 557°F Which of the following plant conditions would require going to RCS Bleed and Feed immediately after tripping all RCPs?

a. Wide range SG levels are as follows: 1A 20%, 1B 30%, 1C 25%, 1D 30%.

b. PZR pressure is 2300 psig due to the loss of secondary heat sink.

c. RCS subcooling is unacceptable per the Iconic Display.

d. Loss of BOTH centrifugal charging pumps.

REACTOR OPERATOR Page 12 of 50 QUESTION: 019 (1.00)

The following Unit 1 plant conditions exist:

- Unit 1 is at 40% power

- Control Bank D is at 180 steps

- Rods are in AUTO The following then occurs:

- The control rods are observed to be stepping out.

- A control rod has dropped into the core by observation of the associated rod position indicator.

1) What parameters are used to determine that the control rod has dropped into the core?

2) How is control rod motion stopped during the event?

a. 1) Associated Rod Bottom light lit ONLY

2) Rod motion is stopped by placing rod control in MANUAL

b. 1) Associated Rod Bottom light lit ONLY

2) Rod motion is stopped by allowing Control Bank D to reach its auto rod stop setpoint of 223 steps

c. 1) Associated Rod Bottom light lit AND Tave initially dropping

2) Rod motion is stopped by placing rod control in MANUAL

d. 1) Associated Rod Bottom light lit AND Tave initially dropping

2) Rod motion is stopped by allowing Control Bank D to reach its auto rod stop setpoint of 223 steps

REACTOR OPERATOR Page 13 of 50 QUESTION: 020 (1.00)

The following plant conditions exist:

- Unit 1 is at 100% power.

- A control rod was determined to be misaligned by 25 steps from its Group Step counter position.

- It has been determined that it will take approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to realign the control rod to its Group Step counter position.

When the control rod realignment to its Group Step Counter position begins

1. reactor power is limited to less than 50%.

2. reactor power must be no more than 75%.

3. the affected control rod must be moved into alignment at a rate of 3 steps/hour.

The crew must adhere to

a. 1 ONLY

b. 2 ONLY

c. 1 and 3 ONLY

d. 2 and 3 ONLY

REACTOR OPERATOR Page 14 of 50 QUESTION: 021 (1.00)

The following plant conditions exist:

- Unit 1 experienced a Reactor Trip

- 1 BEP ES-0.1, Reactor Trip Response, has been entered

- Tave is 550°F and dropping

- Two control rods do NOT have their rod bottom lights LIT For the above plant conditions, what is the MINIMUM amount of emergency boration from the Boric Acid Storage Tank that is required at this time?

a. 1600 gallons

b. 2920 gallons

c. 3480 gallons

d. 5780 gallons QUESTION: 022 (1.00)

Select the set of conditions that would cause INDICATED Pressurizer Level to rise. (Assume leaks on variable and reference legs are NOT in the vapor space.) Consider each condition separately.

1. Leak on Variable Leg

2. Leak on Reference Leg

3. "Delta P" diaphragm ruptures

4. Containment temperature increases

a. 1, 3, and 4 ONLY

b. 1, 2, and 3 ONLY

c. 1, 2, and 4 ONLY

d. 2, 3, and 4 ONLY

REACTOR OPERATOR Page 15 of 50 QUESTION: 023 (1.00)

The following plant condition exist:

- Unit 1 is in MODE 2.

- Reactor thermal power is currently above P-6, but less than P-10.

- Calibration of N-31 is in progress with N-31 and N-35 Level Trip Switches in BYPASS.

Numerous alarms are subsequently received in the control room indicating a loss of Instrument Bus 112. Which of the following action(s) is (are) required IMMEDIATELY and why?

a. Suspend positive reactivity additions AND reduce power to less than P-6 to ensure protection against an uncontrolled rod withdrawal from a low power condition.

b. ONLY suspend positive reactivity additions to ensure protection against a continuous and uncontrolled RCCA bank rod withdrawal from a low power condition during startup, boron dilution and rod ejection events.

c. ONLY reduce power to less than P-6 to ensure protection against an uncontrolled rod withdrawal from a low power condition.

d. Verify the Reactor Trip Breakers are open to ensure protection against a continuous and uncontrolled RCCA bank rod withdrawal from a low power condition during startup, boron dilution and rod ejection events.

REACTOR OPERATOR Page 16 of 50 Question #24 Has Been Deleted From This Examination QUESTION: 024 (1.00)

Unit 1 is starting up with the following conditions:

- Reactor power is at 7%.

- Due to IR Channel N35 reading a full decade lower than IR Channel N36, Channel N35 has been placed in BYPASS.

While withdrawing rods in Control Bank D, IR Channel N36 fails low and the LOSS OF DETECTOR VOLT light on the N36 drawer is lit.

Which one of the following is a required response for this condition?

a. Immediately trip the reactor and follow required actions in 1BEP-0, Reactor Trip or Safety Injection.

b. Immediately reduce power to less than P-6.

c. Immediately stop control rod withdrawal and suspend any other positive reactivity additions.

d. Continue power ascension to greater than P-10.

QUESTION: 025 (1.00)

During refuel operations on Unit 1, the following Area Radiation Monitors indications are received on the RM-11:

- 1AR11J, CONTAINMENT FH ICDT is flashing RED.

- 0AR55J, FUEL HANDLING BLDG FH ICDT is flashing YELLOW.

Which of the following Immediate Actions are to be taken?

Verify automatic . . .

a. FHB Charcoal Booster Fan OVA04CA start.

b. Train A Containment Purge Isolation.

c. FHB Crane upward hoist movement inhibited.

d. Control Room Outside Air Intake A suction realignment to TB and Make-Up Fan start.

REACTOR OPERATOR Page 17 of 50 QUESTION: 026 (1.00)

The following plant conditions exist:

- SG 1A has just developed a tube leak.

- The leak rate in SG 1A has been determined to be 80 GPD.

- SG 1C has a previously determined tube leak of 40 GPD whose leak rate has been stable.

- SG 1D has a previously determined tube leak of 40 GPD whose leak rate has been stable.

Per 1 BOA SEC-8, Steam Generator Tube Leak, is a reactor shutdown required and what is the reason for the required shutdown, if required?

a. Yes, since SG 1A leak rate exceeds the Technical Specification limit.

b. Yes, since total SG leak rate on all 4 SGs exceeds the Technical Specification limit.

c. No, unless the SG 1A leak rate rises at a rate greater than allowable.

d. Yes, since SG 1A leak rate is above the procedural limit, though below the Technical Specification limit, for continued operation.

REACTOR OPERATOR Page 18 of 50 QUESTION: 027 (1.00)

Unit 1 is conducting a start-up, and with Reactor Power at 16%, had just closed the generator output breakers, when the following occurred:

- Main Condenser Vacuum lowered to the Low Vacuum Alarm setpoint

- 1BOA SEC-3, Loss of Condenser Vacuum was entered

- As vacuum reached 7.8 inches HgA, the Reactor was tripped.

- 1 circulating water pump is supplying the Unit 1 condenser.

Which one of the following describes the current status of the steam dumps?

a. Steam dumps are armed.

b. Steam dumps are NOT armed because C-9 is NOT satisfied due to having only one circulating water pump running.

c. Steam dumps are NOT armed because C-9 is NOT satisfied due to low condenser vacuum.

d. Steam dumps are NOT armed because C-7 is NOT satisfied due to insufficient load rejection.

REACTOR OPERATOR Page 19 of 50 QUESTION: 028 (1.00)

Unit 2 start up is in progress with Reactor Power at 16% and all systems normally aligned.

- An electrical transient causes the 2A and 2C RCP Breakers to trip open.

- 2B and 2D RCPs remain running

- The RCP Breaker Position Reactor Trip Circuit malfunctioned and NO Reactor trip occurred.

If NO operator action is taken, what will happen within 2 minutes?

The reactor will . . .

a. NOT automatically trip. RCS overpressure condition will NOT result.

b. NOT automatically trip. Excessive KW/ft condition will NOT result.

c. automatically trip. DNB condition will NOT result.

d. automatically trip. Loss of heat sink condition will result.

QUESTION: 029 (1.00)

While following 1BCA-0.0, Loss of All AC Power, an operator locally closes 1CV8100, RCP Seal Water Return Containment Isolation Valve.

The reason 1CV8100 was closed was to . . .

a. prevent a failure of the RCP seals.

b. prevent steam formation in the CC lines of the Seal Water Heat Exchanger.

c. conserve RCS inventory.

d. protect the RCPs from seal and shaft damage that may occur when a charging pump is started as part of the recovery.

REACTOR OPERATOR Page 20 of 50 QUESTION: 030 (1.00)

BOP CV-1a(b), Startup of the CV System, states that transferring between the Loop A and Loop B Cold Leg charging return valves 1CV8146, RC CL Loop 2 Charging Water Inlet Valve, and 1CV8147, RC CL Loop 1 Charging Water Inlet Valve, should normally be done under Cold Shutdown conditions.

Failure to perform this evolution under Cold Shutdown conditions could result in . . .

a. over-pressurization of the cold leg return lines.

b. runout of the charging pumps.

c. thermal shock to the cold leg return lines.

d. an unplanned reactivity change.

REACTOR OPERATOR Page 21 of 50 QUESTION: 031 (1.00)

Unit 1 was operating at full power with all systems lined up in a normal full power configuration when a transient in the Chemical and Volume Control System occurred, resulting in the following indications:

- Letdown Relief Valve Temperature (TI125) 90°F

- Regenerative Heat Exchanger Letdown Temperature (TI127) 300°F

- Letdown Heat Exchanger Outlet Temperature (TI130) 150°F

- Letdown Line Pressure (PI131) 230 psig

- Letdown Flow (FI132) 150 gpm Which of the following malfunctions caused the above indications?

a. Letdown Orifice Downsteam Relief Valve 1CV8117 failing open.

b. Letdown Low Pressure Control Valve 1CV131 failing open.

c. Low Pressure Control Valve Downstream Relief Valve 1CV8119 failing open.

d. Letdown Demineralizer High Temperature Divert Valve1CV129 failing in the divert position.

REACTOR OPERATOR Page 22 of 50 QUESTION: 032 (1.00)

Unit 2 has experienced a large break LOCA. The following conditions exist:

- 2A RH pump is out-of-service.

- 2B RH is in hot leg recirculation.

- A complete loss of instrument air has occurred.

Which one of the following describes the CORRECT response of the RHR system to the loss of instrument air with respect to the following valves?

- 2RH607, 2B RHR HX Flow Control Valve

- 2RH619, 2B RHR HX Bypass Flow Control Valve

- 2RH611, 2B RH Pump Recirc Valve

a. 2RH607 fails Open, 2RH619 fails As-Is.

b. 2RH607 fails As-Is, 2RH611 fails Open.

c. 2RH607 fails Closed, 2RH611 fails Open.

d. 2RH607 fails Open, 2RH619 fails Closed.

QUESTION: 033 (1.00)

Following a large break LOCA and due to complications while implementing 1BEP ES-1.3, Transfer to Cold Leg Recirculation, the operators have shutdown both CS pumps due to RWST level. The CS pumps control switches are in AFTER TRIP.

The CS pumps are directed to be restarted in step 9.d. RNO.

The 1A CS pump will NOT start from the MCB control switch in this step if . . .

a. 1CS019A, 1A CS Eductor Spray Add Valve, is closed.

b. 1SI8812A, 1A RH Pump RWST Suction Isolation Valve, is open.

c. 1CS007A, 1A CS Pump Discharge Header Isolation Valve, is closed.

d. 1SI8811A, 1A Containment Recirc Sump Outlet Isolation Valve, is closed.

REACTOR OPERATOR Page 23 of 50 QUESTION: 034 (1.00)

Unit 1 was initially operating at 100% power when a safety injection occurred. The plant has entered 1BEP-0, Reactor Trip or Safety Injection, to respond to the event. Present Unit 1 conditions are as follows:

- 1A Safety Injection Pump is Out-of-Service

- Containment pressure is 7.3 psig

- 1B SI Pump, 1A and 1B CV Pumps, and 1A and 1B RH Pumps are all running

- All RCPs are running

- RCS pressure is 1620 psig and slowly lowering

- Both PZR PORVs are closed

- RCS Temperature is 541°F and slowly lowering The crew has learned that the thrust bearing temperature for the 1B SI pump is presently 208°F and rising; therefore, the 1B SI pump was stopped.

While at Step 25 of 1BEP-0, Reactor Trip or Safety Injection, which one of the following actions would be CORRECT in response to the event?

a. Stop RCPs. Stop dumping steam.

b. DO NOT stop RCPs. Establish a maximum cool down rate of 50°F/Hr.

c. DO NOT stop RCPs. Stop dumping steam.

d. DO NOT stop RCPs. Continue to depressurize the RCS by dumping steam to the condenser from intact SGs.

REACTOR OPERATOR Page 24 of 50 QUESTION: 035 (1.00)

Unit 2 is in MODE 4 with a plant cooldown in progress. The following plant conditions exist:

- RCS temperature is 300°F and slowly lowering due to the plant cooldown.

- 2A RH providing shutdown cooling.

- RCS pressure is 310 psig.

- LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) System, is being met, and pressure relief capabilities for LTOP are met by the 2 PZR PORVs.

In these conditions, an inadvertent SI actuation occurred. With NO operator action, what would be the expected plant response? (NOTE: Unit 2 LTOP PORV Setpoint Curve is provided.)

a. One CV pump realigns to its ECCS lineup with the 2A RH suction relief valve being the first relief valve to lift.

b. BOTH CV pumps realign to their ECCS lineup causing pressure in the RCS to rise with the 2A RH suction relief valve being the first relief valve to lift.

c. One CV pump and BOTH SI pumps realign to their ECCS lineup causing pressure in the RCS to rise with the PORVs being the first relief valves to lift.

d. One CV pump realigns to its ECCS lineup with the PORVs being the first relief valves to lift.

REACTOR OPERATOR Page 25 of 50 QUESTION: 036 (1.00)

The following conditions exist for Unit 1:

- Reactor power - 100%

- RCS activity is elevated, but below Technical Specification (ITS) limits

- Pzr pressure - 2225 psig

- Pzr level - 60%

PORV 1RY456 opened due to a Pressurizer pressure instrument channel failure. The operator placed the 1RY456 control switch in the CLOSE position. When conditions stabilize:

- Reactor power - 100%

- Pzr pressure - 2228 psig

- Pzr level - 60%

How would the operator be able to tell if the PORV has fully seated?

a. Level change in RCDT.

b. Verify VCT level trends are normal.

c. Position lights for 1RY-456 showing CLOSE indication.

d. Lowering readings for containment radiation monitors 1AR011/012.

REACTOR OPERATOR Page 26 of 50 QUESTION: 037 (1.00)

A caution in 1BFR-C.1, Response to Inadequate Core Cooling, states, RH pumps should NOT be run longer than 2.4 HOURS without CC flow to the RH heat exchangers. Which of the following statements best describes the basis for this caution?

a. Steam formation in the shell side of the heat exchangers could result in steam binding of the CC pumps and water hammer in CC piping upon restoration of flow.

b. If RCS pressure is above the shutoff head of the RH pumps, pump or motor failure may occur due to pump overheating or cavitation.

c. This limitation will minimize thermal stresses on the RH heat exchangers.

d. CC flow is necessary since it is possible that core cooling restoration will be provided via ECCS recirculation mode, and the sump water must be cooled prior to being delivered to the core or CV/SI pump suctions.

QUESTION: 038 (1.00)

Unit 1 is operating with the following conditions:

- Reactor power is 100%.

- Pressurizer pressure is 2255 psig.

- Pressurizer pressure control is selected to 1PT455/456.

1PT 455 fails low. Assuming NO operator actions, what is the effect on RCS pressure, and the waste gas compressor?

Over the next 20 minutes, RCS pressure will cycle . . .

a. between 2335 and 2315 psig. The waste gas compressor will continue to remove gases from the PRT.

b. between 2335 and 2315 psig. The waste gas compressor will be isolated from the PRT.

c. at 2185 psig. The waste gas compressor will continue to remove gases from the PRT.

d. at 2185 psig. The waste gas compressor will trip on high suction pressure.

REACTOR OPERATOR Page 27 of 50 QUESTION: 039 (1.00)

The Reactor is at 100% power. Which of the following will result in a Solid State Protection System Train B General Warning alarm?

1. The Ground Return Fuse in SSPS cabinet B is OPEN.

2. A loss of 120 VAC Instrument Bus 113.

3. A loss of a 15 VDC power supply to SSPS cabinet B.

a. 1 AND 3 ONLY.

b. 1, 2, AND 3.

c. 1 AND 2 ONLY.

d. 2 AND 3 ONLY.

REACTOR OPERATOR Page 28 of 50 QUESTION: 040 (1.00)

One hour ago, an SI occurred on Unit 2 due to a LOCA. The following conditions exist:

- Offsite Power is available and supplying power to the ESF buses.

- DGs are running but are NOT connected to the ESF buses.

- Containment pressure is 5.3 psig.

- Pressurizer level is 24% and slowly lowering.

- RCS pressure is 800 psig and lowering.

- ALL CV, SI, and RH pumps started and are running.

- The SI signal has been RESET.

With the above plant conditions, a Loss of Offsite Power occurred. Assuming that NO operator actions are taken, which of the following describes what will occur and the effect that this will have on core cooling? (Note: Figure 2BCA 1.1-1 is provided.)

The DGs will re-energize their respective ESF buses . . .

a. but ONLY the CV pumps will automatically restart causing INADEQUATE cooling of the reactor fuel.

b. but ONLY the CV pumps will automatically restart. However, the CV pumps will still be able to provide ADEQUATE cooling of the reactor fuel.

c. but the CV, SI, and RH pumps will NOT automatically restart causing INADEQUATE cooling of the reactor fuel.

d. and the CV, SI, and RH pumps WILL automatically restart resulting in only a short interruption in core cooling caused by sequencing times.

REACTOR OPERATOR Page 29 of 50 QUESTION: 041 (1.00)

Unit 2 is operating at 100% power when the CNMT PEN CLG FLOW HIGH LOW annunciator alarms due to a loss of electrical power to valve 2CC053, Inside Containment Penetration Cooling Supply valve.

Which one of the following would be an effect of the loss of penetration cooling?

a. Inadequate cooling to the containment electrical penetrations could cause excessive heat and long term degradation to the penetration assemblies.

b. Inadequate cooling to the containment electrical penetrations could result in failure of the penetration assembly due to an electrical fault.

c. Piping penetrations with penetration cooling may overheat causing long term degradation to the containment structure.

d. Thermal expansion of piping with penetration cooling could cause cracking of the containment concrete.

QUESTION: 042 (1.00)

A Main Steamline break occurred in Unit 2 Containment 20 minutes ago. The plant is presently shutdown with the following conditions:

- Containment pressure is 12 psig.

- Containment Spray automatically actuated on containment pressure.

Based upon the above information, when can containment spray be secured?

a. Containment Spray must operate for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after initiation.

b. Containment Spray can be secured immediately.

c. Containment Spray can ONLY be terminated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after initiation AND after the Spray Additive Tank LO-2 level lights are LIT.

d. Containment Spray can ONLY be terminated after the Spray Additive Tank LO-2 level lights are LIT.

REACTOR OPERATOR Page 30 of 50 Question #43 Has Been Deleted From This Examination QUESTION: 043 (1.00)

Reactor power is at 100% when the following events occur:

- The main turbine trips.

- The reactor does NOT automatically trip due to a failure of the Turbine Trip circuitry for the Reactor Trip System.

Assuming NO operator action, the reactor will still eventually automatically trip.

What Reactor Trip System Functions will initiate this reactor trip?

1. Overpower delta T.

2. Lo-Lo S/G Level.

3. Overtemperature delta T.

4. Pressurizer Pressure.

a. 1, 2, AND 3 ONLY.

b. 1 AND 4 ONLY.

c. 2 AND 3 ONLY.

d. 3 AND 4 ONLY.

REACTOR OPERATOR Page 31 of 50 QUESTION: 044 (1.00)

The following plant conditions exist:

Generator Output 652 MW Feedwater Regulating Valves (FWRV) AUTO 1A Motor Driven Main Feed Pump (MFP) Running/Manual Operation 1B Turbine Driven MFP Out of Service 1C Turbine Driven MFP Running/AUTO Operation Which of the following actions would occur if there is a suction line break to the 1A motor driven Main Feed Pump?

1A MFP 1C MFP FWRV

a. continues running speed lowers throttle in open direction

b. continues running speed rises throttle in close direction

c. trips speed rises throttle in open direction

d. trips speed rises throttle in close direction QUESTION: 045 (1.00) 1AF004A, 1A AF Pump Discharge Valve, was closed for a surveillance run of the 1A Auxiliary Feedwater pump.

Before the pump was started, the air supply connection to the 1AF004As air accumulator came loose and over the next 5 minutes SLOWLY depressurized the accumulator completely.

Then, a Safety Injection occurs.

The 1A AF pump will

a. NOT start automatically.

b. start automatically, and deliver designed flow to the Steam Generators.

c. start automatically, and provide ONLY recirc flow.

d. start automatically, and deliver NO flow (deadhead).

REACTOR OPERATOR Page 32 of 50 QUESTION: 046 (1.00)

The following conditions exist in Unit 1:

- Reactor Power is 100%.

- A sudden voltage drop on the 4160 VAC buses occurred 30 seconds ago with voltage levels as follows:

- 3835 volts on Bus 141

- 3820 volts on Bus 142 20 seconds later, an SI signal is received.

Which one of the following correctly describes the automatic response of the AC Power System?

a. The degraded voltage trip relays for both the 141 and 142 buses de-energize, sending a trip signal to all feed breakers to Buses 141 and 142. The bus undervoltage relays send a signal to start the 1A and 1B DGs.

b. The degraded voltage trip relay for the 142 bus de-energizes, sending a trip signal to all feed breakers to Bus 142. The 142 bus undervoltage relay sends a signal to start the 1B DG. Bus 141 remains energized from offsite power.

c. The degraded voltage trip relays for both the 141 and 142 buses de-energize, sending a signal directly to start the 1A and 1B DGs and to trip all feed breakers to Bus 141 and 142.

d. The degraded voltage trip relays for both the 141 and 142 buses de-energize after a 10 second time delay, sending a signal directly to start the 1A and 1B DGs and to trip all feed breakers to Bus 141 and 142.

REACTOR OPERATOR Page 33 of 50 QUESTION: 047 (1.00)

Unit 1 is operating at 100% power with the following conditions:

- Bus 111 is being powered from the 111 battery charger.

The control room receives the 125V DC BATT CHGR 111 TROUBLE Annunciator due to a HIGH DC output voltage.

- Battery 111 has 2.12 Volts per cell in each of 58 cells.

Assuming that the crew takes NO actions, what would be the expected voltage at the output of the 111 Battery 1 minute after the annunciator alarms?

a. Greater than 140 VDC.

b. Greater than 130 VDC but less than or equal to 140 VDC.

c. Greater than or equal to125 VDC but less than or equal to 130 VDC.

d. Less than 125 VDC.

QUESTION: 048 (1.00)

How would the RO know from his Main Control Room indications that the 1A DG is in Local Control?

1. The DG TROUBLE/FAIL TO START annunciator would be lit.

2. The white light located on the DG START/STOP switch would be lit.

3. The lights for the DG START/STOP switch would be dark.

a. 1 ONLY.

b. 2 ONLY.

c. 1 AND 3 ONLY

d. 1 AND 2 ONLY.

REACTOR OPERATOR Page 34 of 50 QUESTION: 049 (1.00)

Unit 2 is operating with the following conditions:

- Reactor power is 100%.

- The Control Room Ventilation system is in a normal alignment with 0A VC train in operation.

- 0PR31J, Control Room Outside Air Train A Radiation Monitor is in OPERATE FAILURE.

- 0PR32J, Control Room Outside Air Train A Radiation Monitor is NORMAL.

- 0PR33J, Control Room Outside Air Train B Radiation Monitor is in OPERATE FAILURE.

- 0PR34J, Control Room Outside Air Train B Radiation Monitor is NORMAL.

With the above conditions, 0PR32J, Control Room Outside Air Train A Radiation Monitor, receives a spurious HIGH RADIATION signal.

Which one of the following correctly describes the response of the Control Room HVAC system?

Control Room HVAC makeup air was shifted to the . . .

a. Turbine Building due to the 0PR31J OPERATE FAILURE.

b. Turbine Building due to the 0PR32J HIGH RADIATION.

c. outside air due to the 0PR33J OPERATE FAILURE.

d. outside air due to the 0PR32J HIGH RADIATION.

REACTOR OPERATOR Page 35 of 50 QUESTION: 050 (1.00)

Given the following plant conditions:

- Bus 141 is deenergized because of a bus ground fault.

- A Safety Injection occurred on Unit 1.

With NO operator action, Containment Chilled Water (WO) flow to the RCFCs is through . . .

a. Train A ONLY.

b. Train B ONLY.

c. BOTH Train A AND Train B.

d. NEITHER Train A NOR Train B.

QUESTION: 051 (1.00)

Given the following plant conditions:

- 1A RCP seal injection flow failed to 0 gpm because its throttle valve clogged.

- 0A, 0B, 0C and 0D SX cooling tower fans all tripped.

Alarms:

- RCP LOWER BRNG TEMP HIGH (1-7-C2) and RCP SEAL OUTLET TEMP HIGH (1-7-D3) for 1A RCP have just LIT at 184°F.

BOTH Lower Bearing AND Seal Outlet Temperatures are rising at 1°F/minute.

In 30 minutes, which, if any, RCP trip setpoint will be exceeded?

a. Pump Lower Radial Bearing Temperature ONLY

b. Pump Seal Outlet Temperature ONLY

c. BOTH Pump Lower Radial Bearing Temperature AND Pump Seal Outlet Temperature

d. NEITHER Pump Lower Radial Bearing Temperature NOR Pump Seal Outlet Temperature

REACTOR OPERATOR Page 36 of 50 QUESTION: 052 (1.00)

Given that:

- Reactor power on both units is 100%.

- The Service Air Compressors (SAC) are lined up as follows:

- 1A SAC Lead

- 1B SAC STDBY 2

- 2A SAC STDBY 1

- 2B SAC Lag The Unit 2 SAC Air Receiver pressure dropped from 114 psig to 101 psig, and is now at 112 psig and rising.

With the above conditions, Bus 243 is lost. Concurrently, the 1A SAC trips due to overcurrent on the motor. Assuming NO operator actions and given the following events, which of the following would be expected and in what sequence would they occur?

1. IA RCVR 1 PRESS LOW alarm.

2. SAC RCVR 1 PRESS LOW alarm.

3. 1B SAC starts.

4. 2A SAC starts.

a. 2, and then 3 ONLY.

b. 2, and then 4.

c. 2, 3, and then 1.

d. 1, 2, and then 3.

REACTOR OPERATOR Page 37 of 50 QUESTION: 053 (1.00)

The following conditions exist on Unit 1:

- DC Bus 111 was lost due to a ground fault.

The Operations crew is performing 1BOA ELEC-1, Loss of DC Bus, and are presently attempting to restore instrument air to the containment by opening 1IA065, IA CNMT outside isolation valve, and 1IA066, IA CNMT inside isolation valve.

Which one of the following is the procedurally required method for restoring instrument air to the containment?

a. Open 1IA065 manually from the Main Control Room; locally unlock and open 1IA066.

b. Open 1IA065 manually from the Main Control Room; open 1IA066 from the Main Control Room.

c. Locally unlock and open 1IA065; open 1IA066 from the Main Control Room.

d. Locally unlock and open 1IA065; locally unlock and open 1IA066.

REACTOR OPERATOR Page 38 of 50 QUESTION: 054 (1.00)

The following conditions exist in Unit 1:

- The Reactor is shut down in Mode 3.

- Containment pressure is 0.7 psig.

- You have made an emergency containment entry to investigate a steam leak, and are presently attempting to exit the containment through the personnel airlock doors.

While attempting to exit, you discover that the interior personnel airlock door will NOT open. Five minutes after mechanically opening the interior equalizing valve, it is discovered that pressure has still NOT equalized across the interior door.

Which of the following could be the reason(s) for this condition? (Consider each condition separately.)

1. The exterior equalizing valve is closed.

2. The exterior equalizing valve is open.

3. Containment pressure is too high to allow the inner airlock door to open.

a. 1 AND 3 ONLY.

b. 2 AND 3 ONLY.

c. 1 ONLY.

d. 2 ONLY.

REACTOR OPERATOR Page 39 of 50 QUESTION: 055 (1.00)

Unit 2 is operating at power with the following conditions;

- 2A and 2C RCFCs are running in HIGH speed

- 2B and 2D RCFCs are in STANDBY Subsequently, a lightning strike results in a sudden pressure trip of SAT 242-2. Which of the following describes the status of the RCFCs one minute after the lightning strike?

a. All four RCFCs will be stopped.

b. The 2A and 2C RCFCs will be running in LOW speed, and the 2B and 2D RCFCs will be in STANDBY.

c. All four RCFCs will be running in LOW speed.

d. The 2A and 2C RCFCs will be running in HIGH speed, and the 2B and 2D RCFCs will be in STANDBY.

QUESTION: 056 (1.00)

The following conditions exist in Unit 1 following a refueling outage:

- Reactor power is 12% with the turbine offline.

- The Makeup Mode Selector Switch is in AUTO position.

- VCT level is 32%.

With the conditions above, a calculation error resulted in the Boric Acid Flow Control Potentiometer to be set for 10 ppm boron concentration INSTEAD of 1000 ppm boron concentration.

Assuming that NO operator actions are taken, what would be the eventual result?

a. Rods would step in automatically to compensate for the rise in Tave.

b. Rods would step out automatically to compensate for the lowering of Tave.

c. Reactor power would lower with no automatic rod compensation.

d. Reactor power would rise with no automatic rod compensation.

REACTOR OPERATOR Page 40 of 50 QUESTION: 057 (1.00)

The following conditions exist in Unit 1:

- The plant experienced a large break LOCA 30 minutes ago.

- The operators have taken appropriate actions in accordance with 1BEP-0, Reactor Trip or Safety Injection, and 1BEP-1, Loss of Reactor or Secondary Coolant.

- RCS pressure is stable at 300 psig.

- ECCS pumps are operating in the cold leg recirculation mode.

In this mode, core decay heat is being removed PRIMARILY by . . .

a. the condensation of reflux boiling in the S/Gs.

b. heat transfer between the RCS and the S/Gs due to forced circulation flow.

c. the injection of water from the recirculation sump and the removal of steam/water out from the break.

d. heat transfer between the RCS and the S/Gs due to Natural Circulation flow.

QUESTION: 058 (1.00)

The following conditions exist in Unit 2:

- Reactor power is 100%.

- Pressurizer level is 60%.

With the conditions above and NO operator action, Level Transmitter, 2LT460, fails low. How does this event affect letdown and 2CV-121, Charging Flow Control Valve?

a. Letdown isolates, 2CV-121 throttles in the CLOSED direction.

b. Letdown isolates, 2CV-121 throttles in the OPEN direction.

c. Letdown does NOT isolate, 2CV-121 throttles in the CLOSED direction.

d. Letdown does NOT isolate, 2CV-121 throttles in the OPEN direction.

REACTOR OPERATOR Page 41 of 50 QUESTION: 059 (1.00)

An extra NSO has just reported to the Unit Supervisor that while performing a task in the back of 1PM05J he repositioned the DRPI Accuracy Mode Selector Switch from the A+B position to the A Only position.

As a result . . .

a. ALL General Warning LEDs will be flashing and the DRPI Rod Control Urgent Failure Alarm will actuate.

b. DRPI will be in Half Accuracy Mode.

c. Central Control Failure lights will be illuminated.

d. DRPI will be in normal accuracy.

QUESTION: 060 (1.00)

With a power ascension in progress, the Nuclear Instrumentation channels indicate the following:

N-41 9%

N-42 11%

N-43 11%

N-44 8%

- The Overpower Trip Low Range light is lit on Power Range Drawer N-42.

Based on the above conditions, what actions are required to be taken to address these indications?

a. Lower reactor power to below P-10.

b. Enter BEP-0, Response to a Reactor Trip or Safety Injection.

c. Manually Block the Power Range High Flux Low Setpoint Reactor Trip. No Technical Specification entry is required.

d. No actions are required since the Power Range High Flux Low Setpoint Reactor Trip is Automatically Blocked above 10% Reactor Power.

REACTOR OPERATOR Page 42 of 50 QUESTION: 061 (1.00)

The following plant conditions exist on Unit 1 during a normal plant shutdown:

- RCS pressure is 395 psig

- RCS temperature is 345°F

- 1RY455A and 1RY456 are in Low Temperature Overpressure Protection mode.

Pressurizer PORV 1RY456 has just opened. Which of the following caused 1RY456 to open?

a. 1PT-403 has failed high

b. 1PT-405 has failed high

c. 1PT-406 has failed high

d. 1PT-407 has failed high QUESTION: 062 (1.00)

Unit 2 was operating at steady state, 100% power, when 1MS053A, Steam Pressure Regulating valve to the only operating Steam Jet Air Ejector failed CLOSED.

The operating crew entered the appropriate abnormal operating procedure and initiated actions to start the standby Steam Jet Air Ejector.

How does the failure of the operating Steam Jet Air Ejector INITIALLY affect Unit 2, assuming that the MW feedback loop is in MW OUT and first stage impulse pressure feedback loop is IMP IN?

Condenser Pressure Electrical Megawatts

a. RISES LOWERS

b. CONSTANT LOWERS

c. RISES CONSTANT

d. CONSTANT CONSTANT

REACTOR OPERATOR Page 43 of 50 QUESTION: 063 (1.00)

What design features, interlocks and administrative controls are in place to prevent an inadvertent discharge from the station liquid release tanks?

0WX353, Release Tank Pumps Discharge Valve and 0WX896, Release Tank Pumps Discharge Low Flow Valve . . .

a. have administrative clearance orders hung on their control switches.

b. are isolated until the normally closed manual upstream isolation valves 0WX352 and 0WX895 are locally opened.

c. are key locked valves.

d. have robust operational barriers covering the control switches (Plexiglas covers).

QUESTION: 064 (1.00)

Maintenance has just been completed on 0GW01TC, Gas Decay Tank. Which of the following conditions would require the 0C GDT to be purged with Nitrogen?

a. A Hydrogen concentration approaching 2% in the presence of Oxygen.

b. A Hydrogen concentration approaching 4% in the presence of Nitrogen.

c. An Oxygen concentration approaching 0.5% in the presence of Hydrogen.

d. An Oxygen concentration approaching 2% in the presence of Hydrogen.

REACTOR OPERATOR Page 44 of 50 QUESTION: 065 (1.00)

Given the following information:

- A fire has started in the 1A DG Room.

- The thermal detectors in the 1A DG Room are out-of-service for maintenance.

- The fire has activated one of the fire detectors in the room.

Which detector(s) would have activated, and what would be the result of the activation?

a. Ionization Detector(s) resulting in the actuation of the deluge system.

b. Ionization Detector(s) resulting in the actuation of the CO2 suppression system.

c. Ultraviolet Detector(s) resulting ONLY in an alarm on Fire Detection Panel 1PM09J.

d. Ultraviolet Detector(s) resulting in the actuation of the CO2 suppression system.

QUESTION: 066 (1.00)

Which of the following prints would show the Flow Control Loop for the 0VC03CA Make-Up Fan?

a. 3040 series of prints

b. 3041 series of prints

c. 4030 series of prints

d. 4031 series of prints

REACTOR OPERATOR Page 45 of 50 QUESTION: 067 (1.00)

The Unit Supervisor requests that you open the 2FW009A-D Feedwater Isolation valves.

Before you open these valves which combination of Purge and Flow Permissives must be met?

a. 2/3 FW temperature RTDs at SG inlet is > 255°F for > 8 minutes Feedwater flow through 2FW046 between 30 - 120 gpm for > 8 minutes Total Feedwater flow supplied to specific SG > 1130 gpm

b. 2/3 FW temperature RTDs at 2/3 locations is > 255°F for > 8 minutes Feedwater flow through 2FW046 between 130 - 220 gpm Total Feedwater flow supplied to specific SG > 1310 gpm

c. 2/3 FW temperature RTDs at 3/3 locations is > 255°F for > 8 minutes Feedwater flow through 2FW046 between 80 - 120 gpm for > 8 minutes Total Feedwater flow supplied to specific SG > 1130 gpm

d. 2/3 FW temperature RTDs at 3/3 locations is > 255°F for > 8 minutes Feedwater flow through 2FW046 between 130 - 220 gpm for > 8 minutes Total Feedwater flow supplied to specific SG > 1310 gpm

REACTOR OPERATOR Page 46 of 50 QUESTION: 068 (1.00)

Unit 1 is at 100% power. You are the Unit 1 Assist NSO. The following plant conditions exist:

- Unit 1 is at 100%.

- 2 gpm seat leakage from the Letdown Header Relief Valve to the PRT.

- ACB 1411, Bus 141 to 143 Crosstie breaker, control power fuses are blown

- Pressurizer level is 70%

Based on the conditions listed above what Technical Specification action requirement do you need to inform the Unit Supervisor to enter?

a. Tech Spec 3.4.9, Pressurizer

b. Tech Spec 3.8.1, AC Sources-Operating

c. Tech Spec 3.4.13, RCS Operational LEAKAGE

d. Tech Spec 3.8.9, Distribution Systems-Operating QUESTION: 069 (1.00)

Unit 1 is performing 1BGP 100-2, Plant Startup. The Reactor Operator is withdrawing Shutdown Bank A.

What will be the indicated rod speed and what will be the actual rod speed as Shutdown Bank A steps out?

Indicated speed Actual speed

a. 0 steps/minute 72 steps/minute

b. 48 steps/minute 48 steps/minute

c. 8 steps/minute 64 steps/minute

d. 64 steps/minute 64 steps/minute

REACTOR OPERATOR Page 47 of 50 QUESTION: 070 (1.00)

Assuming that the Rod Control Bank fully withdrawn position is 231 steps with bank overlap at 116

steps,

1. Where will the Bank Overlap Unit Control Rod Bank C Start Position thumbwheel setpoint be set, and,

2. What is the reason for Control Rod Bank Overlap?

START POSITION REASON

a. 225 Increases differential rod worth.

Maintains a more uniform flux distribution.

b. 225 Increases Shutdown Margin.

Maintains a uniform differential rod worth.

c. 232 Maintains a more uniform differential rod worth.

Maintains a more uniform flux distribution.

d. 232 Maintains a more uniform flux distribution.

Reduces potential Axial Flux Offset Power Peaks QUESTION: 071 (1.00)

A Steam Generator Tube Rupture has occurred on Unit 1.

- 1BEP-3, Steam Generator Tube Rupture procedure is being performed.

- Ruptured Steam Generator pressure and RCS pressure has been equalized to minimize RCS leakage.

- The Unit Supervisor has just ordered the performance of BOP MS-11, Operation with Steam Generator Tube Leakage.

The performance of BOP MS-11 will . . .

a. help chemistry identify contaminated secondary systems.

b. minimize secondary system contamination.

c. re-align radwaste demineralizers in anticipation of liquid radwaste processing.

d. align supplemental radiation monitors on the MS system.

REACTOR OPERATOR Page 48 of 50 QUESTION: 072 (1.00)

A Unit 2 Containment Purge has been initiated in accordance with BOP VQ-5, Containment Purge System Operation. 2VQ05C, Unit 2 Containment Mini Purge Exhaust Fan has just automatically tripped.

Which of the following was responsible for causing 2VQ05C to trip?

a. Containment Spray was Manually Actuated.

b. 2AR021J Containment Area Radiation Monitor went into High Alarm.

c. 2PR01J Containment Purge Effluent Radiation Monitor went into High Alarm.

d. Containment Exhaust Smoke Detector 2XY-VQ102 went into High Alarm.

QUESTION: 073 (1.00) 1BEP-0, Reactor Trip or Safety Injection Unit 1, has just been entered for indications of a Large Break LOCA. Which of the following personnel may be designated to monitor the Status Trees and when is the monitoring required to be initiated?

Personnel Start Monitoring

a. Duty STA Immediately upon entry into 1BEP-0

b. Non-Licensed Operator When directed in 1BEP-0 trained on Status trees

c. Unit 2 SRO Immediately after step 4 of 1BEP-0

d. Unit 2 Assist RO Immediately after transition out of BEP-0

REACTOR OPERATOR Page 49 of 50 QUESTION: 074 (1.00)

Unit 1 has experienced a Steam Generator Tube Rupture on the 1C Steam Generator. The Turbine had to be locally tripped but all other systems functioned normally, and the following plant conditions are observed:

- 1A SG NR Level is 4% and rising

- 1B SG NR Level is 3% and rising

- 1C SG NR Level is 8% and rising

- 1D SG NR Level is 3% and rising

- Pressurizer Pressure is 1735 psig and lowering

- Pressurizer Level is 18% and lowering

- 1AF013A-H are ALL OPEN The Crew has performed the actions of 1BEP-0, Reactor Trip or Safety Injection and has transitioned into 1BEP-3, Steam Generator Tube Rupture. The Crew is at step 4 to check ruptured Steam Generator level. Based on the above indications the crew should . . .

a. isolate ALL flow TO and FROM the 1C Steam Generator.

b. isolate flow FROM the 1C Steam Generator Blowdown valves and 1C MSIV & 1C MSIV Bypass valves, but allow AF flow to continue to ALL Steam Generators.

c. isolate all AF flow TO the 1C Steam Generator but do NOT isolate flow FROM 1C Steam Generator Blowdown valves, MSIV, and PORV.

d. NOT isolate any flow path on the 1C Steam Generator until AFTER a cooldown is initiated.

REACTOR OPERATOR Page 50 of 50 QUESTION: 075 (1.00)

The following time line of events occurred on Unit 1:

- 1000 PZR Level started lowering

- 1001 SJAE / GS Exhauster Radiation Monitor 1PR27J went into high alarm

- 1005 Unit Supervisor orders a Reactor Trip and Manual Safety Injection

- 1010 Shift Manager Classified the Event as an Alert (FA1) due to 1A SGTR

- 1015 Crew enters 1BEP-3, Steam Generator Tube Rupture In order to meet the notification requirements for NARS, the INITIAL notification to the State and Local Agencies must be made NO LATER THAN . . .

a. 1015

b. 1016

c. 1020

d. 1025

SENIOR REACTOR OPERATOR Page 1 of 17 QUESTION: 076 (1.00)

The following Unit 1 plant conditions exist:

- A LOCA has occurred

- Command and Control has been transferred to the EOF

- The crew has transitioned to 1BFR-C.1, Response to Inadequate Core Cooling

- Containment pressure is 4 psig and stable

- CETC indicate 1250°F and rising

- SG levels are as follows:

1A 1B 1C 1D 0% NR 20% NR 0% NR 15% NR

- RCP #1 seal Ps are as follows:

1A 1B 1C 1D

1. 1 seal P (psid) 250 125 275 225 The crew is at step 17 in 1BFR-C.1 to check if RCPs should be started. The Unit RO recommends starting ONLY the 1D RCP to provide cooling to the core.

Which of the following is the correct response to the RO recommendation:

a. Direct the RO to start ONLY the 1D RCP.

b. Obtain authorization from the STA to start ONLY the 1D RCP.

c. Direct the RO to start the 1B and 1D RCP.

d. Obtain authorization from the EOF to start all RCPs.

SENIOR REACTOR OPERATOR Page 2 of 17 QUESTION: 077 (1.00)

Unit Two experienced a Large Break LOCA and the operators are currently performing 2BEP ES-1.3, Transfer to Cold Leg Recirculation. Valve 2RH8716A, 2A RH Discharge Crosstie, will not close from the control room. An operator must be dispatched to the unit 2 penetration area to locally close the valve. An emergency dose of 10 Rem may be allowed if the individual volunteers and it is approved by the . . .

a. Station Emergency Director

b. Site Vice President

c. Health Physics Supervisor

d. Unit Supervisor QUESTION: 078 (1.00)

Unit 1 is operating at 100% power, all systems normally aligned and available. Reactor Trip Breaker (RTA/B) testing in progress with the following configuration:

- Reactor Trip Bypass Breaker A (BYA) and Reactor Trip Breaker B (RTB) are both RACKED IN and CLOSED

- Reactor Trip Bypass Breaker B (BYB) and Reactor Trip Breaker A (RTA) are both RACKED OUT and OPEN The EO attempts to RACK IN and CLOSE Reactor Trip Bypass Breaker B (BYB) instead of Reactor Trip Breaker A (RTA). After manual closure, Reactor Trip Bypass Breaker B OPENs, and NO other breakers reposition.

What is the status of the reactor and what actions will the SRO direct?

a. tripped, direct the EO to locally open both the RTB and BYA breakers.

b. tripped, insert a manual reactor trip in accordance with 1BEP-0, Reactor Trip or Safety Injection Unit 1.

c. NOT tripped, IMMEDIATELY enter 1BFR-S.1, Response to Nuclear Power Generation / ATWS Unit 1.

d. NOT tripped, manually trip the reactor in accordance with 1BEP-0, Reactor Trip or Safety Injection Unit 1.

SENIOR REACTOR OPERATOR Page 3 of 17 QUESTION: 079 (1.00)

Unit 2 has tripped due to a trip of the main generator and a subsequent Loss of Offsite Power.

After the initial event the following adverse conditions occurred:

- The 2A Steam Generator has faulted and is at 200 psig.

- The transient caused by the faulted Steam Generator caused tube ruptures in ALL SGs (2A, 2B, 2C, and 2D SGs).

The Operations crew is at step 5 of 2BEP-3, SGTR, check ruptured SGs pressure.

Based upon the above conditions, which one of the following would be a CORRECT way to cool down the RCS?

a. Transition to 2BCA-3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery, and reduce temperature by releasing steam from the 2B, 2C, and 2D Steam Generators using the SG PORVs.

b. Transition to 2BCA-3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery, and reduce temperature by adjusting feed flow to the 2B, 2C, and 2D Steam Generators while maintaining SG PORVs closed.

c. Remain in 2BEP-3, Steam Generator Tube Rupture, and manually dump steam at the maximum rate from the 2B, 2C, and 2D Steam Generators using the SG PORVs.

d. Remain in 2BEP-3, Steam Generator Tube Rupture, and dump steam to the condenser from the 2B, 2C, and 2D Steam Generators.

SENIOR REACTOR OPERATOR Page 4 of 17 QUESTION: 080 (1.00)

Given the following plant conditions on Unit 1:

- A Steam Generator Tube Rupture occurred.

- Secondary coolant specific activity is 1x10-5 Ci/gm dose equivalent I-131.

- The Fire and Oil Sump Discharge Rad Monitor, 0RE-PR005, was declared INOPERABLE.

- The Fire and Oil Sump Discharge Rad Monitor Normal/Bypass Switch is in the BYPASS Position.

Initially, liquid releases from the Fire and Oil Sump are allowed to continue if grab samples are analyzed for radioactivity every (1) hours; if the Fire and Oil Sump Discharge Rad Monitor is NOT returned to service in (2) days, releases must be terminated from this flowpath.

NOTE: TRM 3.11.a is attached.

(1) (2)

a. 12 14

b. 12 30

c. 24 14

d. 24 30

SENIOR REACTOR OPERATOR Page 5 of 17 QUESTION: 081 (1.00)

The following Unit 1 plant conditions exist:

- Unit 1 has experienced a Reactor Trip and Loss of Offsite Power.

- The crew is implementing 1BEP ES-0.1, Reactor Trip Response and 1BOA ELEC-4, Loss of Offsite Power.

- A safety injection then occurs.

The crew is expected to

a. perform 1BOA ELEC-4 in parallel with 1BEP ES-0.1.

b. suspend performance of 1BOA ELEC-4 and implement 1BEP-0, Reactor Trip or Safety Injection.

c. suspend performance of 1BOA ELEC-4 and transition to 1BCA 0.0, Loss of All AC.

d. perform 1BOA ELEC-4 in parallel with 1BEP-0, Reactor Trip or Safety Injection.

QUESTION: 082 (1.00)

The amount of radionuclides stored in any ONSITE outdoor liquid storage tank that is NOT bermed is limited to that amount that . . .

a. would prevent exceeding regulatory limits to ground water if the tanks contents were uncontrollably released.

b. would NOT result in a radiation worker exceeding TEDE exposure if working 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week in close proximity to the tank.

c. can be directed through the plant radioactive waste system and released to the environment without exceeding National Pollution Discharge Elimination System (NPDES) limits.

d. will decay within 31 days to a regulatory limited level that can be safely released without further treatment.

SENIOR REACTOR OPERATOR Page 6 of 17 QUESTION: 083 (1.00)

Unit 2 was in the process of moving fuel in the Reactor Building with all Refueling LCOs met when an earthquake occurred. The following conditions now exist:

- A fuel assembly was dropped in the refueling cavity.

- There has been a rapid drop in refueling cavity level.

- CNMT FH Incident Monitors 1RT-AR011 and 1RT-AR012 are reading greater than the high alarm.

- The CNMT Equipment Hatch is open.

Based upon the above, which one of the following conditions would result in exceeding offsite accident condition dose exposure limits?

a. If the refueling cavity water level drops to 24 feet above the reactor vessel flange.

b. The CNMT Equipment Hatch is NOT closed.

c. Both FHB Ventilation System trains do NOT operate.

d. The Containment Purge valves do NOT meet their 10 CFR 50, Appendix J, leakage criteria.

QUESTION: 084 (1.00)

Which of the following plant conditions would result in the Containment being INOPERABLE or NOT in the required status for the associated plant mode or plant configuration (assume that all other required systems/components for the associated plant mode are OPERABLE):

a. With Unit 1 in Mode 1, the interlock mechanism for the containment personnel air lock doors is found to be INOPERABLE.

b. With Unit 1 in Mode 1, the containment purge valve leakage rate is determined to be in excess of local leak rate requirements, but overall containment leak rate is within limits.

c. With Unit 1 in Mid-Loop operations, both doors of the containment personnel air lock are open.

d. With Unit 1 in Mode 6 and movement of recently irradiated fuel assemblies in progress, both doors of the containment emergency air lock are open.

SENIOR REACTOR OPERATOR Page 7 of 17 QUESTION: 085 (1.00)

Following a large break (LB) LOCA, a Red Path on the Integrity CSF is received. The operators enter 1BFR-P.1, Response to Imminent Pressurized Thermal Shock Conditions, but transition back to 1BEP-1, Loss of Reactor or Secondary Coolant, after verifying RCS pressure and RHR flow.

Which of the following is the reason 1BFR-P.1 directs a - RETURN TO procedure and step in effect - during a LB LOCA?

a. The cooldown due to the LB LOCA will be of short duration. Once the vessel refill is complete, the downcomer will heat up relieving the Vessel Thermal stresses.

b. Due to backflow caused by RHR injection, the Cold Leg Temperatures are NOT a true indication of vessel cooldown rate and Pressurized Thermal Shock is NOT a concern.

c. The actions in 1BEP-1, Loss of Reactor or Secondary Coolant, will address mitigating the PTS concern.

d. Following a LB LOCA, the RCS CANNOT repressurize, therefore vessel integrity is NOT a concern.

SENIOR REACTOR OPERATOR Page 8 of 17 QUESTION: 086 (1.00)

With Unit 2 operating at 100% power, the NSO has provided the following information concerning the RWST to the Operations crew:

- RWST temperature is 38°F.

- RWST borated water level is 90%.

- RWST boron concentration is 2260 ppm.

Based upon the above RWST parameters, what is the required action to take in accordance with LCO 3.5.4, Refueling Water Storage Tank, and, what is the bases for returning the parameter back to within its required Technical Specification values?

Restore the RWST to OPERABLE status due to low . . .

a. water level to assure that following a LOCA the resulting sump pH will be maintained in an acceptable range.

b. RWST temperature. Returning the RWST temperature to within limits prevents ice blockage.

c. water level to ensure a sufficient supply of water is available post LOCA to support Containment Spray System pump operation.

d. boron concentration to assure that following a LOCA the resulting sump pH will be maintained in an acceptable range.

SENIOR REACTOR OPERATOR Page 9 of 17 QUESTION: 087 (1.00)

Unit 2 has experienced a total loss of Feedwater. The following conditions now exist:

- The Operations crew is performing 2BEP-0, Reactor Trip or Safety Injection.

- The Main Turbine did NOT trip automatically and had to be tripped locally.

- Containment Pressure is 0.5 psig.

- Both Unit 2 AF pumps have tripped and cannot be restored.

- PZR pressure is 2250 psig.

- SG wide range levels read as follows:

2A 2B 2C 2D 13% 10% 18% 12%

Considering ONLY the above information, which one of the following would be a CORRECT action to respond to this event?

a. Transition to 2BFR-H.1, Response to Loss of Secondary Heat Sink, then establish RCS Bleed and Feed by actuating SI and bleeding through ONE PORV to the PRT.

b. Transition to 2BFR-H.1, Response to Loss of Secondary Heat Sink, then establish RCS Bleed and Feed by actuating SI and bleeding through TWO PORVs to the PRT.

c. Transition to 2BEP-1, Loss of Reactor or Secondary Coolant, then transition to 2BEP-2, Faulted Steam Generator Isolation.

d. Transition to 2BEP-1, Loss of Reactor or Secondary Coolant, Check PZR PORVs and Isolation Valves to ensure at least ONE is operable/available.

SENIOR REACTOR OPERATOR Page 10 of 17 QUESTION: 088 (1.00)

Concerning LCO 3.4.9, Pressurizer, which one of the following events is the basis for Pressurizer OPERABILITY in MODES 1, 2, and 3?

a. Steam Line Break.

b. Small Break LOCA.

c. Steam Generator Tube Rupture.

d. Loss of Offsite Power.

QUESTION: 089 (1.00)

Unit 2 is operating at 100% power. The 2A DG has been INOPERABLE for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> due to planned maintenance. The Unit Supervisor has just declared the 2B containment spray pump as INOPERABLE due to a motor failure.

AT THIS TIME, and based upon the selections below, what is/are REQUIRED Technical Specification action(s) for this condition? (NOTE: TS LCOs 3.6.6 and 3.8.1 are attached.)

1. Restore containment spray train B to OPERABLE status within 7 days.

2. Enter LCO 3.0.3 Immediately.

3. 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 />.

a. 1 ONLY.

b. 1 AND 2 ONLY.

c. 2 ONLY.

d. 3 ONLY

SENIOR REACTOR OPERATOR Page 11 of 17 QUESTION: 090 (1.00)

Unit 1 is responding to a large-break LOCA. You have just transitioned to1BEP ES-1.3, Transfer to Cold Leg Recirculation, because of low water level in the RWST. The following then occur in succession:

- SI reset is completed.

- The STA announces that a RED path exists due to containment pressure.

- You are unable to open the CC to RH heat exchanger valve 1CC9412A.

- Containment floor water level is 10 inches.

- Step 2 of 1BEP ES-1.3 checks CNMT floor water level - AT LEAST 8 INCHES (13 INCHES ADVERSE CNMT) with the RNO - GO TO 1BCA-1.1, Loss of Emergency Coolant Recirculation.

Based on the conditions listed above, you should

a. complete ALL steps in 1BEP ES-1.3 and then go to 1BFR-Z.1, Response to High Containment Pressure.

b. align both RHR trains for recirculation and then go to 1BFR-Z.1, Response to High Containment Pressure.

c. immediately go to 1BFR-Z.1, Response to High Containment Pressure.

d. go to 1BCA-1.1, Loss of Emergency Coolant Recirculation.

SENIOR REACTOR OPERATOR Page 12 of 17 QUESTION: 091 (1.00)

The operators are performing 1BCA-0.1, Loss of All AC Recovery Without SI Required, step 3 (loading equipment on the ESF bus). The STA has just reported that CETCs are reading 720°F.

Which of the following is the correct response to the above conditions?

a. Immediately exit 1BCA-0.1 and proceed to 1BCA-0.2, Loss of All AC Recovery With SI Required.

b. Immediately exit 1BCA-0.1 and proceed to 1BFR-C.2, Response to Degraded Core Cooling.

c. Continue in 1BCA-0.1 until directed at step 5 to proceed to 1BCA-0.2, Loss of All AC Recovery With SI Required.

d. Continue in 1BCA-0.1 until directed to proceed to 1BFR-C.2, Response to Degraded Core Cooling.

QUESTION: 092 (1.00)

Given the following conditions:

- The 1A CV pump is in operation

- The 1A CV pump is stronger than the 1B CV pump

- 1CV121, Charging Pump Flow Control Valve, is in MANUAL What would be the consequences of swapping CV pumps and which Tech Spec(s) could be impacted? Assume 1CV121 remains in manual and total seal injection flow is 20 gpm following the swap.

a. Seal Injection flow would rise while charging flow would lower. Tech Specs 3.5.2, ECCS - Operating and 3.5.5, Seal Injection Flow could be impacted.

b. Pressurizer Level and Seal Injection Flow would both rise, and Tech Spec 3.5.2, ECCS - Operating could be impacted.

c. Letdown Temperature would rise and Tech Spec 3.5.5, Seal Injection Flow could be impacted.

d. Pressurizer Pressure would lower and no Tech Specs would be impacted.

SENIOR REACTOR OPERATOR Page 13 of 17 QUESTION: 093 (1.00)

The Heat Sink Critical Safety Function monitors (1) while it protects the radiological barriers of the (2) .

(1) Monitors (2) Protects

a. Wide range S/G Lvl, AF Flow, Fuel Clad & RCS Press boundary S/G Pressure

b. Narrow range S/G Lvl, AF Flow, Fuel Clad & Cnmt S/G Pressure

c. Wide range S/G Lvl, Cnmt & RCS Press boundary RCS Cold leg Temp, AF Flow

d. Narrow range S/G Lvl, AF Flow, Fuel Clad & RCS Press boundary S/G Pressure QUESTION: 094 (1.00)

Which one of the following is the MINIMUM requirement for the total number of on-shift Fire Brigade members?

a. Five (5), including the Field Supervisor as Fire Chief.

b. Six (6), including the Field Supervisor as Fire Chief.

c. Five (5), NOT including the Field Supervisor as Fire Chief.

d. Four (4), NOT including the Shift Technical Advisor as Fire Chief.

SENIOR REACTOR OPERATOR Page 14 of 17 QUESTION: 095 (1.00)

The Station has experienced a large break Loss of Coolant Accident on Unit 2. The Shift Manager has assumed the duties of the Shift Emergency Director and is in Command and Control.

Which of the following is a list of the Shift Emergency Directors Non-delegable responsibilities?

a. Classification of the Emergency Notification of the Site Vice President Notification of the State and Federal Agencies Site Assembly / Accountability

b. Classification of the Emergency Authorization for Emergency Dose Exposure Notification of the State and Federal Agencies Determination of Protective Action Recommendations to the State

c. Classification of the Emergency Authorization for Emergency Dose Exposure Site Assembly / Accountability Determination of Protective Action Recommendations to the State

d. Classification of the Emergency Notification of the Site Vice President Notification of State and Federal Agencies Determination of Protective Actions for Plant Personnel

SENIOR REACTOR OPERATOR Page 15 of 17 QUESTION: 096 (1.00)

Which of the following conditions would require a 50.59 Evaluation to be performed?

a. The Emergency Planning (EP) Coordinator is changing the Protective Action Recommendations (PARS) in the Emergency Plan.

b. Mechanical Maintenance will be performing a like-for-like replacement of the 1B SI pump rotating element.

c. System Engineering will be adversely changing the overall isolation time response of several containment isolation valves in the Containment Chilled Water System.

d. The Station Fire Chief is making changes to the Station Fire Plan that is going to change the temperature at which sprinkler heads actuate in the Turbine Lube Oil Area.

QUESTION: 097 (1.00)

An MMD FLS and a Plant Engineer are reviewing a troubleshooting plan with the U-1 US. Per MA-AA-716-004, Conduct of Troubleshooting, the Operations Department will review , Troubleshooting Log, to ensure . . .

a. the Engineering Department has performed a Risk Assessment of the troubleshooting activities.

b. that equipment calibrations are within current calibration frequency.

c. the Plant Engineering review of troubleshooting results is assigned to the proper system engineer.

d. that adequate bounds on the troubleshooting activities have been established to limit plant impact.

SENIOR REACTOR OPERATOR Page 16 of 17 QUESTION: 098 (1.00)

Given the following plant conditions:

- U-1 was at 100% power.

- The main turbine tripped due to an EHC leak.

- NEITHER Reactor Trip Breaker opened automatically or manually from the Main Control Room.

The SRO directs the use of 1BFR S.1, Response to Nuclear Power Generation/ATWS, in order to . . .

a. mitigate the severe RCS pressure transient.

b. mitigate the severe RCS temperature rise.

c. provide adequate boration to shutdown the reactor within 10 minutes.

d. mitigate a possible loss of feedwater after Safety Injection is actuated to trip the reactor.

QUESTION: 099 (1.00)

The following plant conditions exist:

- Workers are making repairs to the 1B Centrifugal Charging Pump.

- A Job Specific RWP was issued at the start of the job.

Which one of the following DOES NOT require the Job Specific RWP to be placed into a HOLD status?

a. Maintenance personnel notice a significant increase in radiological conditions.

b. The Maintenance First Line Supervisor wants to use different tools than those already at the pump to disassemble the pump.

c. Radiation Protection personnel observe Maintenance personnel not complying with RWP requirements.

d. Operations will be conducting an LLRT in the area which could affect radiological conditions in the 1B CV Pump Room.

SENIOR REACTOR OPERATOR Page 17 of 17 QUESTION: 100 (1.00)

The following plant conditions exist:

- A loss of all AC Power occurred 60 minutes ago.

While implementing 1BCA 0.0, Loss of All AC Power, the operators were performing Step 55, SELECT PROPER RECOVERY PROCEDURE. The following conditions existed:

- RCS subcooling is ACCEPTABLE.

- PZR level is at 22% with containment pressure at 8 psig.

- No SI pumps are running.

The recovery procedure to be entered, and the reason for selection, is . . .

a. 1BCA 0.2, Loss of All AC Power Recovery With SI Required, because PZR level is greater than 20% with an ADVERSE containment.

b. 1BCA 0.1, Loss of All AC Power Recovery Without SI Required, because SI equipment has not actuated.

c. 1BCA 0.1, Loss of All AC Power Recovery Without SI Required, because PZR level is greater than 12% with an ADVERSE containment.

d. 1BCA-0.2, Loss of All AC Power Recovery With SI Required, because PZR level is less than or equal to 28% with an ADVERSE containment.

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

REFERENCE Page 1 of 14 ANSWER: 001 (1.00) ANSWER: 005 (1.00)

c. d.

REFERENCE:

1BEP-0, Reactor Trip or Safety Injection Horse Notes RCP-1, RCP Seal Package, I1-EP-XL-01, 1BEP-0, Reactor Trip or Safety Rev. 2 Injection LP: Reactor Coolant and Reactor Coolant New Pumps, Rev. 2, Section Higher III, System Description, A.2.a.9)c); B.2.,3.,5.

Proposed references to be provided to Section IV, System Operation, B.1.d, B.2.e, applicants during examination: None B.2.j 000007K301 ..(KA's) BOA RCP-1, Reactor Coolant Pump Seal Failure, Rev. 102 BOP RC-1 ANSWER: 002 (1.00) New

a. Higher

REFERENCE:

000015/17K ..(KAs)

Horse Notes RY-2, PZR Pressure Control, Rev. 2; RY-1, Pressurizer, Rev. 3; ANSWER: 006 (1.00)

Lesson Plan, Pressurizer (RY), Rev. 6, b.

Attachment B;

REFERENCE:

BAR 1-12-B2, PZR PORV OR SAF VLV Horse Notes TH-1, Rx Theory-1, Rev. 0; TH-OPEN, Rev. 4; 2, Rx Theory-1, Rev. 0.

1BOA INST-2, Rev. 103 New New Higher Higher 000029K105 ..(KAs) 000008A101 ..(KA's)

ANSWER: 007 (1.00)

ANSWER: 003 (1.00) c.

b.

REFERENCE:

1BEP ES-3.1, Post-SGTR Cooldown Using Horse Notes CV-1, CVCS, Rev. 2; RCP-1, Backfill Unit 1, Rev. 101 RCP Seal Package, Rev. 2 ILT Simulator Lesson Plan BEP-3 Series, New SGTR, Rev. 5,Section V, Higher Post-SGTR Cooldown Using Backfill, 000009A213 ..(KA's) CAUTION, p. 59 of 123.

New Higher ANSWER: 004 (1.00) 000038K304 ..(KAs) d.

REFERENCE:

Horse Notes CV-1, CVCS, Rev. 2 ANSWER: 008 (1.00)

LP: Chemical and Volume Control System, d.

Rev. 8,Section III

REFERENCE:

System Description, A.2.e and .f  ???

Higher New New Fundamental 000022 2.1.30 ..(KAs) Proposed references to be provided to applicants duringexamination: None 000040A107 ..(KAs)

REFERENCE Page 2 of 14 ANSWER: 009 (1.00) ANSWER: 013 (1.00) DELETED

d. a.

REFERENCE:

BAR 1-18-C3, EMER TRIP HDR PRESS BOP DC-7, 125 VDC ESF LOW TRIP, Rev. 1 Crosstie/Restoration 1BFR-S.1, Response to Nuclear Power New Generation/ATWS Unit 1, Rev. Higher 102; LP Auxiliary Feedwater System, Rev 5; 000058A101 ..(KAs)

ILT BFR S Series Subcriticality, Rev.3.

New ANSWER: 014 (1.00)

Higher b.

000054A207 ..(KAs)

REFERENCE:

I1-SX-XL-01, Essential Service Water System (Training Plan)

ANSWER: 010 (1.00) New

a. Fundamental

REFERENCE:

Proposed references to be provided to 1BCA-0.0, 0.1, 0.2, Loss of All AC Power, applicants during examination: None Unit 1, Rev. 105, page 000062A201 ..(KAs) 4 of 97 New Higher ANSWER: 015 (1.00) 2.4.50 000055 ..(KAs) a.

REFERENCE:

BOA SEC-4, Loss of Instrument Air ANSWER: 011 (1.00) BEP-0, Reactor Trip or Safety Injection

c. New

REFERENCE:

Higher 1BCA-0.1, Loss of All AC Power Recovery Proposed references to be provided to without SI Required applicants during examination: None New 000065 2.4.2 ..(KAs)

Fundamental 000056K101 ..(KAs)

ANSWER: 016 (1.00) d.

ANSWER: 012 (1.00)

REFERENCE:

b. 1BEP-0, Reactor Trip or Safety Injection

REFERENCE:

1BCA-1.2, LOCA Outside Containment

??? I1-CV-XL-01, Chemical and Volume Control New System (Lesson Plan)

Fundamental 6E-1-4030CV21 000057K301 ..(KAs) 6E-1-4030CV28 New Higher

REFERENCE Page 3 of 14 ANSWER: 017 (1.00) ANSWER: 021 (1.00)

a. b.

REFERENCE:

1BCA-1.1, Loss of Emergency Coolant 1BEP ES-0.1, Reactor Trip Response Recirculation 1BOA PRI-2, Emergency Boration New New Fundamental Higher Proposed references to be provided to Proposed references to be provided to applicants examination: None applicants examination: None 000024K101 ..(KAs)

ANSWER: 018 (1.00)

d. ANSWER: 022 (1.00)

REFERENCE:

d.

1BFR-H.1, Response to Loss of Secondary

REFERENCE:

Heat Sink Task 8, Respond to a PZR level control FR-H.1, Background malfunction.

New Higher Fundamental NEW Proposed references to be provided to 000028K202 ..(KAs) applicants examination: None ANSWER: 023 (1.00)

ANSWER: 019 (1.00) d.

c.

REFERENCE:

TS 3.3.1 and bases for Conditions F, G, H, I 1BAR 1-10-D5, Bank D Rod Stop C-11 and Table 3.3.1-1; 1BOA ROD-3, Dropped or Misaligned Rod LP - Gamma-Metric Source and Intermediate New Range Nuclear Higher Instrumentation, Rev. 2,Section III.A.d. and Proposed references to be provided to e.

applicants examination: None 1B0A INST-1, Nuclear Instrumentation 000003A204 ..(KAs) Malfunction Unit 1, Rev.

104, Attachment C, SR Channel Failure.

New ANSWER: 020 (1.00) Higher

c. 000032K301 ..(KAs)

REFERENCE:

TS LCO 3.1.4, Rod Group Alignment Limits 1BOA ROD-3, Dropped or Misaligned Rod New Higher Proposed references to be provided to applicants examination: None 2.1.7 000005 ..(KAs)

REFERENCE Page 4 of 14 ANSWER: 024 (1.00) DELETED ANSWER: 027 (1.00)

c. c.

REFERENCE:

Horse Notes, NI-3, Intermediate Range, Rev. I1-OA-XL-38, BOA SEC-3, Loss of 3 Condenser Vacuum System Description, Gamma-Metric Source 1BOA SEC-3, Loss of Condenser Vacuum and Intermediate Range New Nuclear Instrumentation, Rev. 2 Fundamental 1B0A INST-1, Nuclear Instrumentation Proposed references to be provided to Malfunction Unit 1, applicants examination: none Attachment B, IR Channel Failure. 000051K301 ..(KAs)

Bank Higher 000033A103 ..(KAs) ANSWER: 028 (1.00) c.

REFERENCE:

ANSWER: 025 (1.00) I1-RC-XL-02, Reactor Coolant Pump

b. New

REFERENCE:

Higher Horse Notes, AR/PR-1,FM-11,23,80, Rev. 1 Proposed references to be provided to Lesson Plan, Radiation Monitors, Rev. 5, applicants examination: None Appendix B, Item 3 and 003K501 ..(KAs)

Section II, System Description, Section H.2.

BOP AR/PR-11T1, Rad Monitor Interlock Function Table, Rev. 7; ANSWER: 029 (1.00) 1BAR RM11-4-1ARAAJ, Rev. 8 c.

BAR RM11-4-0AR55J, Rev. 2.

REFERENCE:

New 1BCA-0.0, Loss of All AC Power Fundamental ECA 0.0, Bases for Loss of All AC Power 000061A203 ..(KAs) New Higher Proposed references to be provided to ANSWER: 026 (1.00) applicants examination: None

d. 003K602 ..(KAs)

REFERENCE:

TS LCO 3.4.13 RCS Operational Leakage 1BOA SEC-8, Steam Generator Tube Leak ANSWER: 030 (1.00)

New c.

Higher

REFERENCE:

2.2.22 000037 ..(KAs) BOP CV-1a(b), Startup of the CV System; Rev 21; Precaution 6 New Higher Proposed references to be provided to applicants examination: None 004K509 ..(KAs)

REFERENCE Page 5 of 14 ANSWER: 031 (1.00) ANSWER: 035 (1.00)

b. a.

REFERENCE:

USAR Chapter 9; Table 9.3-5 LCO 3.4.12 and Bases, LTOP System CVCS Lesson Plan I1-RH-XL-01, Residual Heat Removal New System Higher New 004K610 ..(KAs) Higher Proposed references to be provided to applicants examination:

ANSWER: 032 (1.00) TRM LTOP PORV Setpoint Curve

d. 006A302 ..(KAs)

REFERENCE:

RH-1, RHR Cooldown (Big Notes)

New ANSWER: 036 (1.00)

Fundamental b.

005A102 ..(KAs)

REFERENCE:

Bank ANSWER: 033 (1.00) Higher

d. 007A404 ..(KAs)

REFERENCE:

I1-RH-XL-01, Residual Heat Removal System ANSWER: 037 (1.00) 2BEP ES-1.3, Transfer to Cold Leg b.

Recirculation

REFERENCE:

RH-1, RHR Cooldown (Big Notes) New New Higher Higher 2.1.20 008 ..(KAs)

Proposed references to be provided to applicants examination: None 005A204 ..(KAs) ANSWER: 038 (1.00) b.

REFERENCE:

ANSWER: 034 (1.00) I1-RY-XL-01, Pressurizer (RY)

c. RY-2, PZR Pressure Control (Big Notes)

REFERENCE:

New 1BEP-0, Reactor Trip or Safety Injection Higher New Proposed references to be provided to Higher applicants examination: None Proposed references to be provided to 010K105 ..(KAs) applicants examination: None 006A201 ..(KAs)

REFERENCE Page 6 of 14 ANSWER: 039 (1.00) Proposed references to be provided to

a. applicants examination: None

REFERENCE:

026A101 ..(KAs)

I1-RP-XL-01, SSPS BAR 1-4-B3, SOLID STATE PROT CAB GENERAL WARNING SSPS-1, SSPS Block Diagram (Big Notes)

New Fundamental Proposed references to be provided to applicants examination: None 012K201 ..(KAs)

ANSWER: 040 (1.00) b.

REFERENCE:

Figure 2BCA 1.1-1, Required ECCS Flow vs.

Time from Trip I1-DG-XL-01, Diesel Generator & Aux System 2BEP-01, Loss of Reactor or Secondary Coolant New Higher Proposed references to be provided to applicants examination: Figure 2BCA 1.1-1 013K301 ..(KAs)

ANSWER: 041 (1.00) c.

REFERENCE:

BAR 2-2-D7, CNMT PEN CLG FLOW HIGH LOW I1-PC-XL-01, Primary Containment New Fundamental Proposed references to be provided to applicants examination: None 022K401 ..(KAs)

ANSWER: 042 (1.00) b.

REFERENCE:

2BEP-1, Loss of Reactor or Secondary Coolant New Higher

REFERENCE Page 7 of 14 ANSWER: 043 (1.00) DELETED ANSWER: 047 (1.00)

d. d.

REFERENCE:

Main Steam System, I1-MS-XL-01 BAR 1-21-E8, 125V DC BATT CHGR 111 New TROUBLE Higher I1-DC-XL-01, 125 VDC Power Systems Proposed references to be provided to (Training Plan) applicants examination: None New 039K508 ..(KAs) Higher Proposed references to be provided to applicants examination: None ANSWER: 044 (1.00) 063A402 ..(KAs) c.

REFERENCE:

I1-FW-XL-01, SG Water Level Control ANSWER: 048 (1.00)

System b.

I1-CD-XL-01, Cond/FW System

REFERENCE:

Modified Higher New Proposed references to be provided to Higher applicants examination: None Proposed references to be provided to 059A205 ..(KAs) applicants examination: None 064 2.1.30 ..(KAs)

ANSWER: 045 (1.00)

b. ANSWER: 049 (1.00)

REFERENCE:

b.

1BEP-0, Reactor Trip or Safety Injection

REFERENCE:

I1-AF-XL-01, Auxiliary Feedwater System BOP VC-1, Startup of Control Room HVAC (AF) I1-AR-XL-01, Radiation Monitors 060LSYS, Remote Shutdown Panel I1-VC-XL-01, Control Room HVAC System New New Higher Fundamental Proposed references to be provided to 073K101 ..(KAs) applicants examination: None 061K601 ..(KAs)

ANSWER: 050 (1.00) d.

ANSWER: 046 (1.00)

REFERENCE:

a. BOP CC-14, Post LOCA Alignment of the CC

REFERENCE:

System I1-AP-XL-01, AC Electrical Power System I1-CC-XL-01, Component Cooling Water New System Higher New Proposed references to be provided to Higher applicants examination: None Proposed references to be provided to 062A305 ..(KAs) applicants examination: M-66A, Sheet 1 076K204 ..(KAs)

REFERENCE Page 8 of 14 ANSWER: 051 (1.00) ANSWER: 055 (1.00)

d. d.

REFERENCE:

I1-SX-XL-01, Essential Service Water I1-VP-XL-01, Containment Ventilation and System Purge System BOP CC-8, Isolation of CC between Units 1 VP-3, Containment Cooling (Big Notes) and 2 New BOP CC-14, Post LOCA Alignment of the CC Higher System Proposed references to be provided to 1BOA PRI-7, Essential Service Water applicants examination: None Malfunction 103A101 ..(KAs)

New Higher 076K303 ..(KAs) ANSWER: 056 (1.00) d.

REFERENCE:

ANSWER: 052 (1.00) I1-CV-XL-02, Reactor Makeup Control

a. System

REFERENCE:

I1-RD-XL-01, Rod Control System SA-2, SAC and IA Dryer (Big Notes) New I1-SA-XL-01, Service Air and Instrument Air Higher New Proposed references to be provided to Higher applicants examination: None 078K303 ..(KAs) 001A105 ..(KAs)

ANSWER: 053 (1.00) ANSWER: 057 (1.00)

c. c.

REFERENCE:

SA-2, SAC and IA Dryer (Big Notes)

I1-SA-XL-01, Service Air and Instrument Air Bank 1BOA ELEC-1, Loss of DC Bus Higher New Proposed references to be provided to Higher applicants examination: None Proposed references to be provided to 002A201 ..(KAs) applicants examination: None 078K401 ..(KAs)

ANSWER: 058 (1.00) a.

ANSWER: 054 (1.00)

REFERENCE:

d. I1-RY-XL-01, Pressurizer (RY)

REFERENCE:

RY-3, PZR Level Control (Big Notes)

BAP 1450-8, Primary Containment New Equipment/Emergency Hatch Higher Personnel Airlock Doors Operation Proposed references to be provided to New applicants examination: None Higher 011A303 ..(KAs)

Proposed references to be provided to applicants examination: None 103K404 ..(KAs)

REFERENCE Page 9 of 14 ANSWER: 059 (1.00) ANSWER: 063 (1.00)

b. c.

REFERENCE:

DRPI Lesson Plan BCP 400-TWX01 Figure 29-8 DRPI Display Panel Controls New Figure 29-9 DRPI Accuracy Fundamental Big Note RD-6, Digital Rod Position 068K401 ..(KAs)

Indication New Higher ANSWER: 064 (1.00)

Proposed references to be provided to d.

applicants examination: None

REFERENCE:

014A401 ..(KAs) BOP GW-5 New Fundamental ANSWER: 060 (1.00) 071K504 ..(KAs) c.

REFERENCE:

Technical Specification 3.3.1, RTS ANSWER: 065 (1.00)

Instrumentation c.

Big Note NI-1 & NI-2 Power Range Detector

REFERENCE:

New I1-FP-XL-01, Fire Protection System Higher New Proposed references to be provided to Higher applicants examination: None Proposed references to be provided to 2.2.24 015 ..(KAs) applicants examination: None 086K604 ..(KAs)

ANSWER: 061 (1.00)

c. ANSWER: 066 (1.00)

REFERENCE:

d.

Big Note RY-1 Pressurizer

REFERENCE:

TS 3.4.12 0-4031VC04 New New Fundamental Fundamental Proposed references to be provided to Proposed references to be provided to applicants examination: None applicants examination: None 016K108 ..(KAs) 2.1.24 ..(KAs)

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

a. d.

REFERENCE:

Lesson Plan Main Turbine Controls and Lesson Plan 25r05, Condensate and Protection Feedwater New New Higher Fundamental Proposed references to be provided to Proposed references to be provided to applicants examination: None applicants examination: None 055K301 ..(KAs) 2.1.32 ..(KAs)

REFERENCE Page 10 of 14 ANSWER: 068 (1.00) ANSWER: 072 (1.00)

a. a.

REFERENCE:

2-4030VQ013 Electrical Print New 2-4030VQ01 Electrical Print Fundamental Big Note VP-2, Cnmt Purge Proposed references to be provided to New applicants examination: None Higher 2.1.33 ..(KAs) Proposed references to be provided to applicants examination: None 2.3.11 ..(KAs)

ANSWER: 069 (1.00) d.

REFERENCE:

ANSWER: 073 (1.00) 1BGP 100-2, Plant Startup d.

Big Note RD-2, Reactor Control Unit

REFERENCE:

New BAP 1310-10, revision 10, HU-AA-104-101, Higher Procedure Use and Proposed references to be provided to Adherence, Byron Addendum applicants examination: None New 2.2.1 ..(KAs) Higher 2.4.12 ..(KAs)

ANSWER: 070 (1.00)

c. ANSWER: 074 (1.00)

REFERENCE:

b.

Rod Control Lesson Plan

REFERENCE:

BOP RD-7 1BEP-3, Steam Geneator Tube Rupture Big Note RD-2 & RD-3 WOG Background Document for E-3 Steam New Generator Tube Rupture Higher New Proposed references to be provided to Higher applicants examination: None Proposed references to be provided to 2.2.33 ..(KAs) applicants examination: None 2.4.24 ..(KAs)

ANSWER: 071 (1.00)

b. ANSWER: 075 (1.00)

REFERENCE:

d.

1BEP-3, Steam Generator Tube Rupture

REFERENCE:

BOP MS-11, Operation with Steam EP-AA-114, Notifications Generator Tube Leakage Bank New Fundamental Fundamental Proposed references to be provided to Proposed references to be provided to applicants examination: None applicants examination: None 2.4.39 ..(KAs) 2.3.10 ..(KAs)

REFERENCE Page 11 of 14 ANSWER: 076 (1.00) ANSWER: 079 (1.00)

a. a.

REFERENCE:

1BFR-C.1, Response to Inadequate Core 2BCA-3.1, SGTR with Loss of Reactor Cooling Coolant ? Subcooled Recovery FR-C.1 Background Information for WOG Desired Emergency Response 2BEP-3, Steam Generator Tube Rupture Guideline New BAP 1310-10, revision 10, HU-AA-104-101, Higher Procedure Use and Proposed references to be provided to Adherence, Byron Addendum applicants examination: None EP-AA-112-100-F-01, Shift Emergency 000038A208 ..(KAs)

Director Checklist New Fundamental ANSWER: 080 (1.00)

Proposed references to be provided to d.

applicants examination: None

REFERENCE:

000015/017 ..(KAs) TRM 3.11.a, Radioactive Liquid Effluent Monitoring Instrumentation ANSWER: 077 (1.00) BAR RM11-1-0PR05J

a. BRP 5820-13, Response to High Radiation

REFERENCE:

Monitor Alarms EP-AA-113, revision 8, Personnel Protective Bank Actions Higher Bank Proposed references to be provided to Fundamental applicants examination: TRM 3.11.a Proposed references to be provided to 000038 2.3.6 ..(KAs) applicants examination: None 000011 2.3.10 ..(KAs)

ANSWER: 081 (1.00) b.

ANSWER: 078 (1.00)

REFERENCE:

d. 1BOA ELEC-4, Loss of Offsite Power

REFERENCE:

New 1BEP-0, Reactor Trip or Safety Injection Unit Higher 1, Rev. 108; EF-2, Proposed references to be provided to ESF Setpoints, Rev. 0; BAR 1-10-B7, RX applicants examination: None BYP BRKR 1A RACKED IN, 000056 2.1.20 ..(KAs)

Rev. 51; BAR 1-10-B8, RX BYP BRKR 1B RACKED IN, Rev. 51; TS 3.3.1, RTS Instrumentation, Condition N.

New Higher 000029A207 ..(KAs)

REFERENCE Page 12 of 14 ANSWER: 082 (1.00) ANSWER: 086 (1.00)

a. d.

REFERENCE:

T.S. 5.5.12, Explosive Gas and Storage Tank TS LCO and Bases for 3.5.4, RWST Radioactivity New Monitoring Program section C ensuring an Higher uncontrolled release 006A210 ..(KAs) would result in concentrations below the limits of 10CFR20, appendix B, table 2, column 2. ANSWER: 087 (1.00)

New b.

Fundamental

REFERENCE:

059 2.3.1 ..(KAs) 2BEP-0, Reactor Trip or Safety Injection 2BEP-1, Loss of Reactor or Secondary Coolant ANSWER: 083 (1.00) 2BFR-H.1, Response to Loss of Secondary

c. Heat Sink

REFERENCE:

New TS LCO 3.7.13, FHB Ventilation System Higher TS LCO 3.9.4, Containment Penetrations Proposed references to be provided to TS LCO 3.9.7, Refueling Cavity Water Level applicants examination: None New 007 2.4.6 ..(KAs)

Higher Proposed references to be provided to applicants examination: None ANSWER: 088 (1.00) 000036A202 ..(KAs) d.

REFERENCE:

TS LCO and Bases 3.4.9, Pressurizer ANSWER: 084 (1.00) New

d. Fundamental

REFERENCE:

Proposed references to be provided to TS LCO 3.6.1 and Bases, Containment applicants examination: None TS LCO 3.6.2 and Bases, Containment 010A201 ..(KAs)

Airlocks TS LCO 3.9.4 and Bases, Containment Penetrations ANSWER: 089 (1.00)

New a.

Fundamental

REFERENCE:

000069A201 ..(KAs) TS LCO and Base 3.6.6, Containment Spray and Cooling System TS LCO 3.8.1, AC Sources - Operating ANSWER: 085 (1.00) New

d. Higher

REFERENCE:

Proposed references to be provided to New applicants examination:

Higher TS LCOs 3.6.6 and 3.8.1 Proposed references to be provided to 2.2.22 026 ..(KAs) applicants examination: None 2.1.32 ..(KAs)

REFERENCE Page 13 of 14 ANSWER: 090 (1.00) ANSWER: 093 (1.00)

d. d.

REFERENCE:

1BCA-1.1, Loss of Emergency Coolant ST-1, Status Trees Recirculation New 1BFR-Z.1, Response to High Containment Fundamental Pressure Proposed references to be provided to Modified applicants examination: None Higher 2.4.21 035 ..(KAs)

Proposed references to be provided to applicants examination: None 103A203 ..(KAs) ANSWER: 094 (1.00) a.

REFERENCE:

ANSWER: 091 (1.00) BAP 1210-a.

REFERENCE:

Fundamental BCA-0.1, Loss of All AC Power Recovery 2.1.4 ..(KAs) without SI Required BCA-0.2, Loss of All AC Power Recovery with SI Required ANSWER: 095 (1.00)

BFR-C.2, Response to Degraded Core b.

Cooling.

REFERENCE:

New EP-AA-11, Higher EP-AA113-F-02, Authorization for Proposed references to be provided to Emergency Exposure applicants examination: None New 001A209 ..(KAs) Fundamental Proposed references to be provided to applicants examination: None ANSWER: 092 (1.00) 2.1.20 ..(KAs) c.

REFERENCE:

Tech Spec. 3.5.2, ECCS ? Operating ANSWER: 096 (1.00)

Tech Spec 3.5.5, Seal Injection Flow c.

1BOSR 5.5.1.1, RCS Seal Injection Flow

REFERENCE:

Verification Monthly Surv. LS-AA-104-1000, 50.59 Resource Manual New LS-AA-128, RegulatoryReview of Proposed Higher Changes to the Approved Proposed references to be provided to Fire Protection Program applicants examination: None New 011A204 ..(KAs) Higher Proposed references to be provided to applicants examination: None 2.2.5 ..(KAs)

REFERENCE Page 14 of 14 ANSWER: 097 (1.00)

REFERENCE:

d. RP-MW-403-1001, Radiation Work Permit

REFERENCE:

Processing MA-AA-716-004, Conduct of Troubleshooting New sections 3.3.4 Fundamental New Proposed references to be provided to Fundamental applicants examination: None Proposed references to be provided to 2.3.7 ..(KAs) applicants examination: None 2.2.18 ..(KAs)

ANSWER: 100 (1.00) d.

ANSWER: 098 (1.00)

REFERENCE:

a. 1BCA-0.0, Loss of All AC Power

REFERENCE:

New Fundamental Higher Proposed references to be provided to Proposed references to be provided to applicants examination: None applicants examination: None 2.4.7 ..(KAs) 2.3.5 ..(KAs)

ANSWER: 099 (1.00)

b. (********** END OF EXAMINATION **********)

ANSWER KEY Page 1 of 1 001 c 021 b 041 c 061 c 081 b 002 a 022 d 042 b 062 a 082 a 003 b 023 d 043 d 063 c 083 c 004 d 024 c 044 c 064 d 084 d 005 d 025 b 045 b 065 c 085 d 006 b 026 d 046 a 066 d 086 d 007 c 027 c 047 d 067 d 087 b 008 d 028 c 048 b 068 a 088 d 009 d 029 c 049 b 069 d 089 a 010 a 030 c 050 d 070 c 090 d 011 c 031 b 051 d 071 b 091 a 012 b 032 d 052 a 072 a 092 c 013 a 033 d 053 c 073 d 093 d 014 b 034 c 054 d 074 b 094 a 015 a 035 a 055 d 075 d 095 b 016 d 036 b 056 d 076 c 096 c 017 a 037 b 057 c 077 a 097 d 018 d 038 b 058 a 078 d 098 a 019 c 039 a 059 b 079 a 099 b 020 c 040 b 060 c 080 d 100 d

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