ML040690107

Initial RO Examination 02/05/2004

ML040690107
Person / Time
Site:	Kewaunee
Issue date:	02/05/2004
From:	NRC/RGN-III
To:	Nuclear Management Co
References
50-305/OL04-301
Download: ML040690107 (116)

Text

U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name: MASTER RO Date: February 05, 2004 Facility/Unit: KEWAUNEE / U1 Region:

III Reactor Type: Westinghouse 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 six hours after the examination starts.

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

Applicants Signature Results Examination Value 98.0 Points Applicants Score

__________ Points Applicants Grade

__________ Percent NRC Official Use Only

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 and/or operating test) being administered. The applicants may be briefed individually or as a group. Facility licensees are encouraged to distribute a copy of this appendix to every examinee before the examinations begin. 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 asking them before starting that part of the test.

3.

SRO applicants will be tested at the level of responsibility of the senior licensed shift position (i.e., shift supervisor, senior shift supervisor, or whatever the title of the position may be).

4.

You must pass every part of the examination to receive a license or to continue performing license duties. Applicants for an SRO-upgrade license may require remedial training in order to continue their RO duties if the examination reveals deficiencies in the required knowledge and abilities.

5.

The NRC examiner is not allowed to reveal the results of any part of the examination until they have been reviewed and approved by NRC management. Grades provided by the facility licensee are preliminary until approved by the NRC. You will be informed of the official examination results about 30 days after all the examinations are complete.

PART B - WRITTEN EXAMINATION GUIDELINES 1.

[Read Verbatim] After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.

2.

To pass the examination, you must achieve an overall grade of 80.00 percent or greater, with a 70.00 percent or better 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 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 RO examination is six hours, the 25-question, SRO-only exam is three hours, the combined RO/SRO exam is eight hours, and SRO exam limited to fuel handling is four hours; extensions will be considered under extenuating circumstances.

4.

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

5.

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

6.

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

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

7.

If you have any questions concerning the intent or the initial conditions of a question, do not hesitate asking them before answering the question. Ask questions of the NRC examiner or the designated facility instructor only. When answering a question, do not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question. For example, you should not assume that any alarm has activated unless the question so states or the alarm is expected to activate as a result of the conditions that are stated in the question. Finally, answer all questions based on actual plant operation, procedures, and references. If you believe that the answer would be different based on simulator operation or training references, you should answer the question based on the actual plant.

8.

Restroom trips are permitted, but only one applicant at a time will be allowed to leave.

Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating.

9.

When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examination.

10.

After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.

11.

Do you have any questions?

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REACTOR OPERATOR Page 5 QUESTION: 001 (1.00)

All AC power has been lost and equipment is being placed in PULLOUT per ECA-0.0, LOSS OF ALL AC POWER. Which pump will be kept available, and why?

a.

One RHR pump, to provide RCS inventory makeup.

b.

One SI pump, to provide RCS inventory makeup.

c.

One service water pump, to provide Diesel Generator cooling.

d.

One charging pump, to provide RCP seal cooling.

REACTOR OPERATOR Page 6 QUESTION: 002 (1.00)

The controller output which automatically positions the Main Feedwater Regulating Valves (FW-7A & B) to maintain programmed level uses steam generator narrow range water level AND which of the following?

a.

The setpoint established by the operator on the control station b.

Steam flow and feedwater flow c.

Steam flow, feedwater flow and Turbine impulse pressure d.

Turbine impulse pressure

REACTOR OPERATOR Page 7 QUESTION: 003 (1.00)

Containment temperature has increased from 100°F to 160°F due to a containment cooling malfunction. If the plant is stable at 100% power and there are negligible RCS or containment pressure changes, which one of the following describes the effect of the increase in containment temperature on the pressurizer level indicated by the pressurizer level control channels?

a.

Indicated level will be HIGHER than actual level because the reference leg fluid density decreases.

b.

Indicated level will be LOWER than actual level because the reference leg fluid density decreases.

c.

Indicated level will be HIGHER than actual level because the elevated containment temperature causes increased flashing in the reference leg.

d.

Indicated level will be LOWER than actual level because of the elevated containment temperature causes increased flashing in the reference leg.

REACTOR OPERATOR Page 8 QUESTION: 004 (1.00)

Both Main Feedwater pumps are running when the "A" Main Feedwater pump trips. Which one of the following conditions prevents the start AND continued operation of the "A" Main Feedwater pump?

a.

Lube oil pressure is 10 psig.

b.

Suction pressure is 210 psig.

c.

Recirculation valve is FULL open.

d.

Only one Condensate pump is running.

REACTOR OPERATOR Page 9 QUESTION: 005 (1.00)

With power to 480 volt Bus 1-43 deenergized, which pressurizer heaters would be affected?

A.

Backup group 1B ONLY B.

Backup groups 1D AND 1E ONLY C.

Backup Group 1A ONLY with transfer switch in Normal D.

Backup group 1A with transfer switch in Alternate AND control group 1C ONLY

REACTOR OPERATOR Page 10 QUESTION: 006 (1.00)

Given the following:

Station and Instrument Air System is in a normal/automatic system lineup.

The station air compressor preferred selector switch is aligned to Compressor G.

Which of the following automatically occur when the Station and Instrument Air System air header pressure is decreasing and reaches 95 psig?

a.

Air dryers are bypassed AND Station air compressor F starts.

b.

Station air compressor F starts AND SA-400, (SA Header B Supply Valve) is fully closed.

c.

SA-200, (SA Header A Supply Valve) AND SA-400, (SA Header B Supply Valve) start to close.

d.

Instrument air compressor C starts and isolates from station air header AND SA-200, (SA Header A Supply Valve) is fully closed.

REACTOR OPERATOR Page 11 QUESTION: 007 (1.00)

With the plant initially at 100% power, steady-state, which one of the following describes an effect of power range channel N-42 upper detector failing HIGH?

a.

Reactor trips on high flux.

b.

Main feed regulating valves fully open.

c.

Control rods step out to high bank rod stop.

d.

Over-temperature DT setpoint for one channel decreases.

REACTOR OPERATOR Page 12 QUESTION: 008 (1.00)

The unit is at 100% power with Instrument air pressure at 115 psig. Station Air Compressor F is tagged out for preventive maintenance. What will instrument air pressure do when Station Air Compressor G is tripped.

a.

decrease, then increase to a value above its previous pressure.

b.

remain unaffected.

c.

slowly decrease to a point where the unit will have to be tripped.

d.

decrease, then increase to a value below its previous pressure.

REACTOR OPERATOR Page 13 QUESTION: 009 (1.00)

The following plant conditions exist:

Reactor is Critical at 10-3 percent power RCS Tavg is 547°F and steady Pressurizer pressure is 2235 psig and steady Control Bank D position is 100 steps ONE Control Bank D rod drops Describe the INITIAL response of Tavg and Pressurizer pressure (PZR Press):

Tavg PZR Press a.

Remain the same Remain the same b.

Increase Remain the same c.

Remain the same Increase d.

Decrease Decrease

REACTOR OPERATOR Page 14 QUESTION: 010 (1.00)

The following conditions exist:

90% Reactor power.

Pressurizer Pressure control is in automatic Backup heaters are in "AUTO".

Actual Pressurizer Pressure is 2235 psig.

The Pressurizer Pressure Master Controller malfunctions and its SETPOINT drifts to 2100 psig over a 10 minute period. Which of the following describes the INITIAL automatic response of the Pressurizer Pressure Control System as this failure occurs?

a.

Spray valves throttle closed and variable heaters go to maximum current.

b.

Spray valves throttle open and variable heaters go to minimum current.

c.

Pressurizer PORVs PR-2A and PR-2B open, Spray valves throttle open, and Group C heaters go to minimum current.

d.

Spray valves throttle closed, pressurizer backup heater groups go to maximum current, and backup heaters come "ON."

REACTOR OPERATOR Page 15 QUESTION: 011 (1.00)

Which of the following electrical interlocks prevents release of radioactive water to the RWST during transfer to Containment Sump Recirculation during a LOCA?

a.

Close SI-300A/B, RWST Supply Valve to an RHR pump, before opening SI-350A/B, Containment Sump B Isolation Valve to an RHR pump.

b.

Close SI-208 or SI-209 SI Recirculation Isolation Valves before opening SI-350A/B and 351A/B, Containment Sump B Isolation Valves to an RHR pump.

c.

Close SI-208 or SI-209 SI Recirculation Isolation Valves before opening RHR-299A/B RHR Heat Exchanger Outlet Valve to an SI pump.

d.

Open SI-350A/B and 351A/B, Containment Sump B Isolation Valves before opening RHR-299A/B RHR Heat Exchanger Outlet Valve to an SI pump.

REACTOR OPERATOR Page 16 QUESTION: 012 (1.00)

Given the following conditions:

Component Cooling Pump "A" is running Component Cooling Pump "B" is in standby D/G "B" is out of service for maintenance A safety Injection signal is generated. Which one of the following describes the response of the Component Cooling Water Pumps?

Pump A Pump B a.

Runs continuously Sequences on b.

Stops then sequences on Remains off c.

Runs continuously Remains off d.

Stops then sequences on Sequences on

REACTOR OPERATOR Page 17 QUESTION: 013 (1.00)

Given the following conditions:

The plant is operating at 18% power The high pressure piping to RCS flow instrument FT-411 on Loop A breaks What is the status of Loop A RCS flow indicators and what is the resulting plant condition (assume NO operator action is taken)?

a.

All Loop A flow indicators will read low, and the reactor will trip on RCS loop low flow.

b.

All Loop A flow indicators will read low, and the reactor will trip on low PRZR pressure.

c.

Only FI-411 RCS flow indication will read low, and no reactor trip is generated.

d.

Only FI-411 RCS flow indication will read low, and the reactor will trip on low PRZR pressure.

REACTOR OPERATOR Page 18 QUESTION: 014 (1.00)

Given the following conditions:

The plant is at 15% power S/G B level channel LT-473 is removed from service per A-MI-87 If S/G B level channel LT-471 fails high, what would be the status of feed for the S/Gs?

a.

Both S/Gs are being fed from the motor-driven AFW Pumps only.

b.

Both S/Gs are being fed from the turbine-driven AFW Pump only.

c.

S/G A is being fed from the motor-driven AFW Pump. S/G B has no feed flow.

d.

Feed to S/G A remains normal. Feed to S/G B lowers due to throttling close of FW-7B, S/G B Main Feed valve.

REACTOR OPERATOR Page 19 QUESTION: 015 (1.00)

Which one of the following describes the operation of LD-13, Letdown Line Pressure Relief Valve?

a.

Relieves at 200 psig to the VCT b.

Relieves at 150 psig to the PRT c.

Relieves at 150 psig to the VCT d.

Relieves at 200 psig to the PRT.

REACTOR OPERATOR Page 20 QUESTION: 016 (1.00)

The plant is in Hot shutdown, with the following conditions:

Tave = 547°F with both RXCPs operating A Main Feedwater Pump is running Steam Generator A Narrow Range Level = 20%

Steam Generator B Narrow Range Level = 15%

Condensate Pump A is Hold Carded in Pullout.

Condensate Pump B is running All three Auxiliary Feedwater Pumps are in Off and in Pullout All the support conditions for the Auxiliary Feedwater Pumps are met At the direction of the CRS, the BOP places all three Auxiliary Feedwater Pump control switches into the AUTO position.

Without further operator actions, what is the status of the Auxiliary Feedwater Pumps?

MDAFW Pumps TDAFW a.

Running Running b.

Running Off c.

Off Running d.

Off Off

REACTOR OPERATOR Page 21 QUESTION: 017 (1.00)

SP-54-086, "Turbine Stop and Governor Valve Operability," directs depressing the CLOSE SV-1 pushbutton. Which of the following statements identifies an expected response of the turbine control valves (CV-1 through CV-4) and turbine stop valve (SV-1) to this action?

a.

SV-1 closes, then CV-1 and CV-3 close.

b.

SV-1 closes, then CV-1 and CV-2 close.

c.

CV-1 and CV-2 close, then SV-1 closes.

d.

CV-1 and CV-3 close, then SV-1 closes.

REACTOR OPERATOR Page 22 QUESTION: 018 (1.00)

The following conditions exist:

A LOCA has occurred.

The crew is trying to reduce ECCS flow.

All equipment is operating properly.

Wide Range RCS pressure is 800 psig.

65°F of subcooling is required to stop one of the ECCS pumps.

What is the MAXIMUM Core Exit Thermocouple temperature at which the pump is stopped?

a.

430°F.

b.

455°F.

c.

480°F.

d.

505°F.

REACTOR OPERATOR Page 23 QUESTION: 019 (1.00)

Given the following:

The reactor is critical at 1% power.

SI Accumulator A water volume was just found to be 1270 cubic ft (48%).

SI Accumulator B water volume was found to be 1220 cubic ft (20%) at the same time.

What action is required due to these conditions, if any?

a.

No action is needed due to these conditions.

b.

SI Accumulator A water volume must be restored to limits.

c.

SI Accumulator B water volume must be restored to limits.

d.

Action must be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to go to HOT SHUTDOWN.

REACTOR OPERATOR Page 24 QUESTION: 020 (1.00)

Given the following conditions:

Reactor Vessel Head AND Upper Internals have been removed.

Residual Heat Removal boron concentration is 2458 ppm.

The reactor has been shutdown for 7 days.

Spent Fuel Pool Pump "A" is operating per N-SFP-21, Spent Fuel Pool Cooling and Cleanup System. Spent Fuel Pool Pump "B" is inoperable.

Residual Heat Removal Pump "A" is operating. Residual Heat Removal Pump "B" is inoperable.

Refueling Cavity level is greater than 23 feet above the vessel flange.

Based on the plant status given, determine what condition must be resolved to meet requirements for a fuel shuffle within the reactor that is NOT a full core offload.

a.

Residual Heat Removal boron concentration is too low.

b.

Spent Fuel Pool Cooling Pump "B" is required to be operable.

c.

The reactor has not been shutdown long enough.

d.

Residual Heat Removal Pump "B" is required to be operable.

REACTOR OPERATOR Page 25 QUESTION: 021 (1.00)

Which of the following is the purpose of NAD-02.07, Kewaunee Refueling Operations?

a.

Provides step-by-step instructions to be used by fuel handlers during core offload.

b.

Provides administrative instructions for reactor engineers to follow when developing the fuel shuffle sequence to prevent inadvertent criticality during core reload.

c.

Describes the organization and responsibilities for reactor vessel disassembly, reassembly, and fuel handling operations.

d.

Provides step-by-step instructions on completing a Fuel Assembly Handling Deviation Report.

REACTOR OPERATOR Page 26 QUESTION: 022 (1.00)

A point source in containment is reading 300 mRem/hr at a distance of two (2) feet. Two options are available to complete a mandatory work assignment near this radiation source:

Option 1 - ONE operator can perform the assignment in fifty (50) minutes working at a distance of three (3) feet from the source.

Option 2 - THREE operators, using special extension tooling, can perform the assignment in sixty (60) minutes at a distance of six (6) feet from the source Which is the preferred option based on ALARA and the corresponding total exposure?

a.

Option 1, with a total exposure of 0.100 Person-REM b.

Option 1, with a total exposure of 0.111 Person-REM c.

Option 2, with a total exposure of 0.100 Person-REM d.

Option 2, with a total exposure of 0.111 Person-REM

REACTOR OPERATOR Page 27 QUESTION: 023 (1.00)

Which of the following is listed in E-0, "Reactor Trip or Safety Injection," step 1 CONTINGENCY ACTIONS but is NOT in FR-S.1, "Response to a Nuclear Power Generation/ATWS," step 1 CONTINGENCY ACTIONS?

a.

Source range counts b.

Intermediate range power c.

Reactor Trip Breakers d.

Bypass Breakers

REACTOR OPERATOR Page 28 QUESTION: 024 (1.00)

Which of the following is the basis for maintaining SG Narrow Range levels between 4% and 50% during procedure ECA-0.0, "LOSS OF ALL AC POWER?"

a.

Ensures the capability to cooldown the reactor only after AC power is restored.

b.

Ensures heat transfer capability exists to remove heat from the RCS.

c.

Narrow Range level is the only reliable indication of SG inventory available after a loss of all AC power.

d.

Provides capability to monitor the SGs for a Steam Generator Tube Rupture.

REACTOR OPERATOR Page 29 QUESTION: 025 (1.00)

Which ONE of the following describes the mitigating strategies contained in ECA-1.1, "Loss of Emergency Coolant Recirculation?"

a.

Minimizing the depletion of the RWST, Maximizing Subcooling, determination of minimum containment spray requirements.

b.

Maximizing Subcooling, minimizing the depletion of the RWST, and depressurization of the RCS to minimize break flow.

c.

Minimizing the depletion of the RWST, determination of minimum containment spray requirements, and depressurization of the RCS to minimize break flow.

d.

Maximizing subcooling, determination of minimum containment spray, and depressurization of the RCS to minimize break flow.

REACTOR OPERATOR Page 30 QUESTION: 026 (1.00)

Given the following CRDM coil rod motion sequence:

1. Stationary Gripper Energized at Low Voltage

2. Stationary Gripper Energizes at High Voltage

3. Movable Gripper Energizes

4. Stationary Gripper De-energizes The next step in the sequence is __________ and the rod is being moved ____________

a.

Lift Coil De-energizes, Inward b.

Lift Coil Energizes, Outward c.

Stationary Gripper Energizes at Low Voltage, Inward d.

Stationary Gripper Energizes at High Voltage, Outward

REACTOR OPERATOR Page 31 QUESTION: 027 (1.00)

With the plant in the operating mode, the reactor operator receives ROD CONTROL SYSTEM ABNORMAL annunciator and SER point 1692, "Rod Control System Non-Urgent Failure."

Instrument and Control technicians narrow the problem to the Rod Control Logic Cabinet. What may have caused this condition to occur?

a.

Slave cycler cycles without a GO pulse b.

A redundant power supply has been lost c.

Oscillator fails to generate signals when called for d.

There is a loose printed circuit card in the logic circuitry

REACTOR OPERATOR Page 32 QUESTION: 028 (1.00)

Which of the following conditions results in a Main Steam Isolation?

a.

Containment pressure of 15 psig.

b.

Steamline flow of 4X106 lb/hr AND an "SI" signal.

c.

Steamline flow of 5X106 lb/hr AND Tavg of 530°F.

d.

Steamline flow of 0.8X106 lb/hr, Tavg of 530°F, AND an "SI" signal.

REACTOR OPERATOR Page 33 QUESTION: 029 (1.00)

The unit is at 100% power with Charging pump A operating in automatic. Charging Line Flow Control valve, CVC-7, is throttled to 70%. CVC-7 then fails open. Charging pump As speed will....

a.

NOT change during this event b.

decrease at first, then returns to its initial speed c.

increases at first, then remains above its initial speed d.

decreases at first, then remains below its initial speed

REACTOR OPERATOR Page 34 QUESTION: 030 (1.00)

A plant startup is in progress with power at 8 percent. Intermediate Range drawer N-36 Level Trip switch is in the BYPASS position. What is the plant response to removal of N-36 control power fuses and the reason for the plant response?

a.

A trip will occur because the Level Trip Switch bypass function will be removed.

b.

A trip will not occur because the Level Trip Switch is in the bypass position and the bypass function is not affected at any power level.

c.

A trip will not occur because the Level Trip Switch is in the bypass position and power is less than P-10.

d.

A trip will occur because the Level Trip Switch bypass function is operable only above P-10.

REACTOR OPERATOR Page 35 QUESTION: 031 (1.00)

While operating at 100% power, Graphics Display 4 on PPCS shows the B7 CET temperature to be 592°F.

Which of the following sequence of actions identifies how the operator would display the B7 CET temperature value (Channel 20 on Train B) at the ICCMS panels?

Depress the CET ID/CET TEMP pushbutton to illuminate the...

a.

CET TEMP lamp, then depress the AVG/HOT pushbutton to display the B7 CET temperature.

b.

CET TEMP lamp, then depress the REF1/REF2 pushbutton to display the B7 CET temperature.

c.

CET ID lamp, then depress the REF1/REF2 pushbutton until "B7" is displayed, then depress the CET ID/CET TEMP again to display the B7 CET temperature.

d.

CET ID lamp, then depress the AVG/HOT pushbutton until "20" is displayed, then depress the CET ID/CET TEMP again to display the B7 CET temperature.

REACTOR OPERATOR Page 36 QUESTION: 032 (1.00)

Given the following conditions:

Four CFCUs were running prior to the event A Safety Injection signal was just received Containment Pressure has increased to 3.5 psig and is currently stable What condition would you expect the CFCUs to be in at this time?

a.

All CFCUs would be running with SW return isolation valves throttled to maintain temperature b.

All CFCUs would be running with CFCU Emergency Discharge dampers open and SW return isolation valves throttled to maintain temperature c.

All CFCUs would be running with SW return isolation valves fully open d.

All CFCUs would be running with CFCU Emergency Discharge dampers and SW return isolation valves fully open

REACTOR OPERATOR Page 37 QUESTION: 033 (1.00)

Given the following conditions:

The plant has tripped and a Safety Injection signal has been generated.

Engineered Safeguard Features Actuation System Train A relays have failed to operate ONLY the Service Water System valves.

Train B relays have operated properly.

At the completion of the SI sequence, what is the status of Service Water flow to Component Cooling Water (CCW) Heat Exchanger A, if NO operator action is taken?

a.

There will be NO Service Water flow through the CCW heat exchanger.

b.

SW flow will be at a set constant value lower than its post-accident expected value.

c.

SW flow through the CCW heat exchanger will be at its post-accident expected value.

d.

SW flow will be lower than its post-accident expected value and controlled by the CCW outlet header temperature.

REACTOR OPERATOR Page 38 QUESTION: 034 (1.00)

Given the following conditions:

The plant is at 55% power and steady.

The "A" Main FW pump is in pullout.

The "B" Main FW pump is running with two condensate pumps.

Annunciator 47062-A, "S/G A Program Level Deviation" is LIT.

All 3 S/G A level indicators are 38% and steady.

Which of the following describes actions to be taken by the operator based on these conditions?

a.

START "A" Main FW pump per N-FW-05B, Feedwater System Normal Operation AND MONITOR "A" S/G level automatic control for proper operation.

b.

REDUCE load to < 50% AND MONITOR "A" S/G level automatic control for proper operation.

c.

Place Feedwater Flow Control Valve FW-7A to MANUAL AND GO to A-MI-87, Bistable Tripping for Failed Reactor Protection or Safeguards Instrumentation.

d.

GO to A-FW-05A, Abnormal Feedwater System Operation, AND DETERMINE if manual feedwater control is required.

REACTOR OPERATOR Page 39 QUESTION: 035 (1.00)

Given the following plant conditions:

The reactor is at 100% power All systems are in a normal lineup Based on these conditions, which one of the following correctly states the power supply to the Reactor Coolant Pumps?

RXCP A RXCP B a.

MAT MAT b.

RAT MAT c.

MAT RAT d.

RAT RAT

REACTOR OPERATOR Page 40 QUESTION: 036 (1.00)

In procedure FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, the step to check if SI can be terminated is based on which of the following parameter(s):

a.

RCS Subcooling ONLY.

b.

RCS Subcooling and RCS Cold Leg Temperatures.

c.

RCS Pressure ONLY.

d.

RCS Pressure and RCS Cold Leg Temperatures.

REACTOR OPERATOR Page 41 QUESTION: 037 (1.00)

ECA-2.1, Uncontrolled Depressurization of Both Steam Generators, has been entered from E-2, Faulted Steam Generator Isolation. Containment pressure is 1.0 psig. Each steam generator is being fed at 100 gpm producing an RCS cooldown rate of 120°F/hr. Steam generator water levels are as follows:

Steam Generator A narrow range levels - 2%

Steam Generator B narrow range levels - 8%

Based on the conditions above, what is the appropriate initial operator action?

a.

Decrease feed flow to each steam generator to 60 gpm.

b.

Decrease feed flow to "A" steam generator ONLY to 60 gpm.

c.

Increase feed flow as required to maintain "A" steam generator narrow range level greater than or equal to 4%.

d.

Control feed flow as required to maintain RCS hot leg temperatures stable or decreasing.

REACTOR OPERATOR Page 42 QUESTION: 038 (1.00)

The following plant conditions exist:

A safety injection has actuated A transition has been made to ES-1.1, SI Termination No charging pump is running CC flow to the RXCP thermal barrier HX has been lost since the SI actuation What action is initially taken associated with RXCP seal cooling and what is the reason for the action?

a.

RXCP seal injection is isolated before starting a charging pump, to avoid taking time to reestablish seal cooling since RXCP seals are already heated up.

b.

A charging pump is started and then CC flow is established to the RXCP thermal barriers, to prevent steam binding of the CC system.

c.

CC flow is established to the RXCP thermal barriers and then a charging pump is started, to prevent RXCP shaft warping.

d.

CC flow is established to the RXCP thermal barriers, to prevent thermal shock to the RXCP seals.

REACTOR OPERATOR Page 43 QUESTION: 039 (1.00)

Given the following plant conditions:

Reactor trip occurred with subsequent loss of RXCPs.

RCS Pressure is 800 psig Operators have implemented ES-0.2, "Natural Circulation Cooldown" to go to Cold Shutdown.

A cooldown rate of 25°F/hour has been established.

RCS depressurization has been initiated PZR level - Unexpected large variations are occurring RVLIS RXCPs OFF Indication - 90%

The Shift Manager has determined that cooldown and depressurization shall proceed as quickly as possible. Which ONE of the following describes the appropriate actions?

a.

Pressurize the RCS to collapse the voids, continue the cooldown and remain in ES-0.2, Natural Circulation Cooldown.

b.

Raise the cooldown rate to collapse the voids and remain in ES-0.2, Natural Circulation Cooldown.

c.

Pressurize the RCS to collapse the voids, continue the cooldown and transition to ES-0.3, "Natural Circulation Cooldown With Steam Voids in Vessel."

d.

Raise the cooldown rate to collapse the voids and transition to ES-0.3, "Natural Circulation Cooldown With Steam Voids in Vessel.

REACTOR OPERATOR Page 44 QUESTION: 040 (1.00)

Given the following plant conditions:

- The unit is at 80% power.

- Control Bank D rod K7 is stuck at 220 steps (not the most reactive rod).

- All other Control Bank "D" rods are at 220 steps.

- Tavg = 567°F.

How would the shutdown margin calculation performed prior to the condition discovered above be affected?

a. No effect on shutdown margin.

b. Shutdown margin would be more conservative.

c. Shutdown margin would be less conservative.

d. The effect on shutdown margin can not be determined.

REACTOR OPERATOR Page 45 QUESTION: 041 (1.00)

Which of the following indications will be inaccurate during the performance of ES-0.2, "Natural Circulation Cooldown", with RXCPs secured?

a.

RCS core exit TCs.

b.

RCS wide range hot leg temperature.

c.

RCS T-average indication.

d.

RCS wide range cold leg temperature.

REACTOR OPERATOR Page 46 QUESTION: 042 (1.00)

The following plant conditions exist:

Reactor power is 80%

Rod Control is in MANUAL All other controls are in AUTO An inadvertent Emergency Boration was performed for two minutes. Considering steady-state to steady-state conditions, which of the following parameters will NOT change?

a.

Reactor Power b.

RCS Tavg c.

Przr Level d.

S/G Pressure

REACTOR OPERATOR Page 47 QUESTION: 043 (1.00)

Given the following conditions:

The plant is at 50% power Condensate pump A is in pullout Condensate pump B is running If condenser hot well level subsequently decreases to 10%, which one of the actions below is now required?

a.

Startup condensate pump A per N-CD-03, Condensate System.

b.

Address abnormal condensate conditions per A-CD-03, Condensate System Abnormal Operation.

c.

OPEN MU-3B, Condenser Emergency Make-up valve per N-CD-03, Condensate System.

d.

Respond to the reactor trip per E-0, Reactor Trip or Safety Injection

REACTOR OPERATOR Page 48 QUESTION: 044 (1.00)

Due to decreasing condenser vacuum, the operator performs the actions of E-AR-09, Loss of Condenser Vacuum. At one point, condenser pressure increases to 5 inches Hg absolute, and the operator must locally place the hogging jet in service.

In addition to opening MS-400, Steam Supply to Hogging Jet, and throttling MS-401, Steam to Hogging Jet, to maintain between 105-115 psig on PI-11323, four additional valves, listed below, must be aligned to place the hogging jet in service:

AR-302, Gland Steam Cdsr Exhaust to Outside AR-305, Gland Steam Cdsr Exhaust to Vent AR-100, Hogging Jet Air Inlet AR-2A, First Stage Ejector Inlet Which of the following is the correct valve alignment?

AR-302 AR-305 AR-100 AR-2A a.

CLOSED OPEN CLOSED OPEN b.

OPEN CLOSED OPEN CLOSED c.

CLOSED OPEN OPEN CLOSED d.

OPEN CLOSED CLOSED OPEN

REACTOR OPERATOR Page 49 QUESTION: 045 (1.00)

The Gaseous Radioactive Waste System (WG) vent header pressure has just increased to 2 psig. Per A-GWP-32B, the AUTOMATIC ACTIONS the operator must now verify are:

a.

The start of the Waste Gas Compressors.

b.

The closure of Gas Decay Tank inlet isolation valves for the tank selected for fill.

c.

The closure of Gas Decay Tanks to Holdup Tanks valve.

d.

The opening of Gas Decay Tank inlet isolation valve for the tank selected for standby.

REACTOR OPERATOR Page 50 QUESTION: 046 (1.00)

The contents of a Gas Decay Tank is being released in accordance with the appropriate administrative controls when Gas Decay Tanks to Plant Vent, WG-36/CV-31215, closed.

Which of the following monitors could have caused this to occur?

a.

BOTH R-13 and R-14 (Aux Building Vent Monitors) b.

BOTH Aux. Area Monitors 03-06 and 03-08 (Beta Air Monitors Aux. Building) c.

ONLY R-13 (Aux Building Vent Monitor) d.

ONLY R-14 (Aux Building Vent Monitor)

REACTOR OPERATOR Page 51 QUESTION: 047 (1.00)

Which of the following describes the detector types used in Area Radiation Monitors at Kewaunee?

a.

ONLY GM tubes and ion chambers b.

ONLY ion chambers c.

ONLY GM tubes d.

Ion chambers, GM tubes AND scintillation detectors

REACTOR OPERATOR Page 52 QUESTION: 048 (1.00)

Procedure A-RM-45, Abnormal Radiation Monitoring, requires the operator to determine if Post Accident Recirc must be started due to a failure of R-23, Control Room Vent Monitor, by monitoring specific process and area monitors for increasing radiation levels.

Besides R-1, Control Room Area Monitor, what additional AREA Radiation monitor(s), if any, must be monitored and Post Accident Recirc manually started on increasing levels of radiation?

a.

BOTH R-10, New Fuel Pit Area Monitor AND R-5, Fuel Handling Area Monitor.

b.

NO additional Area Radiation Monitors.

c.

ONLY R-10, New Fuel Pit Area Monitor.

d.

ONLY R-5, Fuel Handling Area Monitor.

REACTOR OPERATOR Page 53 QUESTION: 049 (1.00)

ES-0.2, Natural Circulation Cooldown, is being implemented. The following conditions exist:

RCS cold leg Temp is at 380°F.

RCS Pressure is at 1450 psig.

All CRDM fans are off and CANNOT be started.

18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> waiting period has begun.

What is the basis for the 18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> waiting period?

a.

Prevent damage to the CRDM coils due to overheating.

b.

Ensure heat is being removed from the Steam Generator to prevent void formation in the U-tubes.

c.

Minimize void formations in the Reactor Vessel head during subsequent RCS depressurization to place RHR in service.

d.

Allow sufficient flow to the upper head region to make the upper head fluid temperature equal to the cold leg fluid temperature.

REACTOR OPERATOR Page 54 QUESTION: 050 (1.00)

The following plant conditions exist:

PZR pressure control channel selector switch is in the 4-3 position PZR pressure yellow channel (PT-449 IV) has just failed low What is the effect of these conditions on the RCS?

a.

All PZR heaters will come ON; PZR PORVs PR-2A and PR-2B would not be available to open on a subsequent high RCS pressure condition.

b.

PZR heaters are unaffected; ONLY PZR PORV PR-2A would not be available to open on a subsequent high RCS pressure condition.

c.

All PZR heaters will come ON; PZR PORV PR-2A would not be available to open on a subsequent high RCS pressure condition.

d.

PZR heaters are unaffected; PZR PORVs PR-2A and PR-2B would not be available to open on a subsequent high RCS pressure condition.

REACTOR OPERATOR Page 55 QUESTION: 051 (1.00)

Following an ATWS and an SI actuation from 100% power, the Reactor Trip Breakers remain closed. What effect will this have on the plant?

a.

Automatic Turbine Trip will not occur.

b.

Automatic steam line isolation will be blocked.

c.

Automatic SI re-actuation CANNOT be blocked.

d.

Feedwater isolation due to SI actuation will be blocked.

REACTOR OPERATOR Page 56 QUESTION: 052 (1.00)

Which of the following parameters is monitored to determine the need to minimize DC loads while performing ECA-0.0, LOSS OF ALL AC POWER?

a.

Battery amps b.

Battery amp-hours c.

Battery volts d.

Battery specific gravity

REACTOR OPERATOR Page 57 QUESTION: 053 (1.00)

Given the following:

The unit is stable at 15% power.

A failure of Instrument Bus I, BRA-113, occurs.

All systems and control loops are in their NORMAL position.

Which of the following action(s) are required, if any, to restore pressurizer (PRZR) pressure and level conditions resulting from this failure?

a.

Place PRZR spray valves in MANUAL, Position PRZR Pressure Control Switch to another position, AND Then place PRZR spray valves back in AUTO.

b.

Place Charging Pump Speed control to MANUAL, Position PRZR Level Control Switch to another position, AND Then place Charging Pump Speed control back to AUTO.

c.

Position PRZR Level Control Switch to another position, AND Then restore normal letdown and PRZR heaters.

d.

No actions are required to restore PRZR pressure and level.

REACTOR OPERATOR Page 58 QUESTION: 054 (1.00)

Which one of the following would cause annunciator 47041-P, ROD BOTTOM ROD DROP, to alarm?

a.

Control Bank B demand is 38 and a Control Bank B IRPI reads 18.

b.

Control Bank A demand is 18 and a Control Bank A IRPI reads 32.

c.

Shutdown Bank B demand is 32 and a Shutdown Bank B IRPI reads 18.

d.

Shutdown Bank A demand is 18 and a Shutdown Bank A IRPI reads 32.

Removed From Exam Because All Answers Are Correct.

REACTOR OPERATOR Page 59 QUESTION: 055 (1.00)

Given the following:

There has been a fire necessitating the evacuation of the control room.

E-0-06, Fire in Alternate Fire Zone has been entered from E-FP-08, Emergency Operating Procedure - Fire.

Which of the following indications is available at the Dedicated Shutdown Panel?

a.

S/G 1A Narrow Range Level b.

Charging Flow c.

RWST Level d.

Reactor Coolant Loop B Hot Leg Temp

REACTOR OPERATOR Page 60 QUESTION: 056 (1.00)

Given the following conditions:

A large break LOCA occurs.

Containment pressure is observed to be 25 psig.

Containment Spray has NOT initiated.

Manual actuation of Containment Spray has been unsuccessful.

All other ESF actuations and components have functioned normally.

What actions need to be taken to manually initiate Containment Spray for Train A?

Manual start of ICS Pump A and...

(1)

ICS-5A/MV-32066 and ICS-6A/MV32067, Ctmt Spray Pump A Discharge Isolation valves (2)

ICS-201/CV-31272 and ICS-202/CV31273, ICS Recirculation RWST valves (3)

CI-1001A/CV31393 and CI-1001B/CV-31394, Caustic Additive to Ctmt Spray valves a.

(1) check auto open (2) check closed (3) check auto open b.

(1) manual open (2) check closed (3) check auto open c.

(1) manual open (2) check closed (3) manual open d.

(1) manual open (2) manual close (3) manual open

REACTOR OPERATOR Page 61 QUESTION: 057 (1.00)

Given the following conditions:

Radiation Monitor R-11, Containment Atmosphere, is in HIGH alarm.

All other plant conditions are normal Which of the following lists valves in the Reactor Building Ventilation System to be verified automatically CLOSED by the operator?

a.

ONLY the following valves:

-TAV-12, Cntmt Purge/Vent Supply

-RBV-2, Cntmt Purge/Vent Supply B

-RBV-5, Cntmt Purge/Vent Exhaust

-RBV-3, Cntmt Purge/Vent Exhaust B

-SA-7003B, Hydrogen Dilution to Cnmt

-LOCA-2B, Post LOCA H2 Cntmt Vent Isol B.

b.

ONLY the following valves:

-TAV-12, Cntmt Purge/Vent Supply

-RBV-5, Cntmt Purge/Vent Exhaust

-SA-7003B, Hydrogen Dilution to Cnmt

-LOCA-2B, Post LOCA H2 Cntmt Vent Isol B.

c.

ONLY the following valves:

-TAV-12, Cntmt Purge/Vent Supply

-RBV-2, Cntmt Purge/Vent Supply B

-RBV-5, Cntmt Purge/Vent Exhaust

-RBV-3, Cntmt Purge/Vent Exhaust B.

d.

ONLY the following valves

-RBV-2, Cntmt Purge/Vent Supply B

-RBV-3, Cntmt Purge/Vent Exhaust B

-SA-7003B, Hydrogen Dilution to Cnmt

-LOCA-2B, Post LOCA H2 Cntmt Vent Isol B.

REACTOR OPERATOR Page 62 QUESTION: 058 (1.00)

ECA-1.1, Loss of Emergency Coolant Recirculation, determines the required number of operating ICS pumps based on which of the following?

a.

Containment pressure, containment temperature, and sump level.

b.

Containment pressure, operating CFCUs, and sump level.

c.

Containment temperature, operating CFCUs, and RWST level.

d.

Containment pressure, operating CFCUs, and RWST level.

REACTOR OPERATOR Page 63 QUESTION: 059 (1.00)

The plant is at 100% power when control room operators receive SPENT FUEL POOL ABNORMAL Annunciator. The SFP level is decreasing AND is lower than the SFP canal level.

The CRS instructs you to initiate makeup to the SFP. Per procedure A-SFP-21, Abnormal SFP Cooling and Cleanup System Operation, you would use...

a.

the RWST b.

Service Water c.

the Reactor Makeup Control System d.

Reactor Makeup Water through manual makeup valve

REACTOR OPERATOR Page 64 QUESTION: 060 (1.00)

The Unit is at 40% power in a power ascension to full power. All systems are aligned in their normal lineups for the current power level except Turbine EHC control is in MANUAL-IMP OUT.

The operator depresses the CV raise pushbutton for 2 seconds to continue the load ascension.

Which of the following is the response of the main feedwater regulating valves to this action?

The Main Feedwater Regulating Valves will initially throttle...

a.

CLOSED due to swell, and then throttle OPEN when level drops below 44%.

b.

OPEN due to the steam flow/feed flow mismatch, and then regulate to control level at 44%

c.

CLOSED due to the steam flow/feed flow mismatch, and then throttle OPEN when level drops below 44%.

d.

OPEN due to shrink, and then regulate to control level at 44%.

REACTOR OPERATOR Page 65 QUESTION: 061 (1.00)

SP-05B-284, "Turbine Driven AFW Pump Full Flow Test - IST," was in progress. The turbine driven auxiliary feedwater pump (TDAFP) was started and had been running for 2 minutes.

Alarm window 47062-N, "T/D AFW Pump Abnormal" then annunciates and the NAO reports the TDAFP auxiliary lube oil pump is continuously stopping and starting with lube oil pressure fluctuating between 8 and 17 psig.

Which ONE of the following correctly explains the above conditions?

a.

Conditions are normal, no operator action is required, the test should continue.

b.

Conditions are normal, the NAO should be directed to locally shutdown the auxiliary lube oil pump, the test should continue.

c.

The shaft driven pump has malfunctioned, the test should be terminated.

d.

The Auxiliary Lube Oil Pump has malfunctioned, the test should be terminated.

REACTOR OPERATOR Page 66 QUESTION: 062 (1.00)

Given the following conditions:

The plant is at 100% power S/G blowdown is in service in Mode II Condenser air removal is aligned for normal operation The NCO positions the R-19 keyswitch to the OFF position Which of the following describes the effect of the operator actions?

a.

Blowdown flowpath switches to Mode I alignment.

b.

Blowdown flowpath switches to the Primary Sampling System.

c.

Condenser Air Ejector discharge AR-6 (CV-31168) remains in the duct position.

d.

Condenser Air Ejector discharge AR-6 (CV-31168) switches to its ATM position.

REACTOR OPERATOR Page 67 QUESTION: 063 (1.00)

During a liquid radwaste discharge from the Waste Condensate Tanks to the Auxiliary building standpipe, control room operators receive a Waste Disposal Panel Trouble Alarm and dispatch an operator. The operator reports from the Waste Disposal Panel (53702) that LIQUID WASTE MONITOR R-18 HIGH RADIATION" has alarmed. Radiation Monitor R-18 is verified to be alarming, but automatic action(s) do NOT occur. What automatic operation of A-RM-45, Abnormal Radiation Monitoring System must now be performed manually?

a.

Manually close WD-19, Waste Liquid Discharge Isolation Valve.

b.

Manually close WD-17, Waste Condensate Pumps Discharge Valve.

c.

Manually close WD-22, Waste Condensate Pumps to Auxiliary Building Standpipe.

d.

Stop Waste Condensate Pump 1A.

REACTOR OPERATOR Page 68 QUESTION: 064 (1.00)

When the RHR system is placed in the shutdown cooling mode of operation, component cooling is __(1)__ aligned to the associated RHR heat exchanger prior to RHR pump start AND component cooling flows through the __(2)__ side of the RHR heat exchanger.

(1)

(2) a.

Automatically Tube b.

Manually Tube c.

Automatically Shell d.

Manually Shell

REACTOR OPERATOR Page 69 QUESTION: 065 (1.00)

A LOCA has occurred. Post-LOCA containment hydrogen concentration is 7%. What method is available to address hydrogen control in the containment?

a.

dilute the containment atmosphere.

b.

place the Hydrogen Recombiner in service.

c.

vent containment through the Shield Building Ventilation System.

d.

spray containment using the containment spray pumps.

Removed From Exam Because There Are No Correct Answers.

REACTOR OPERATOR Page 70 QUESTION: 066 (1.00)

During refueling operations, an irradiated fuel assembly is dropped in the reactor vessel. A fuel handler reports to the control room that gas bubbles are emanating from the dropped assembly.

Shortly afterwards, R-11 alarms on high radiation. The control room operator enters E-FH-53A, "Dropped or Damaged Fuel Assembly" procedure and ____(1)_____. Controls for the R11 alarm ____(2)_____.

a.

(1) verifies that the Auxiliary Building Special Vent system starts (2) automatically stops upward movement of the manipulator hoist b.

(1) verifies that the Containment Vent Isolation occurred (2) do NOT affect the fuel handling system c.

(1) actuates the containment evacuation alarm (2) automatically stops movement of the manipulator trolley and bridge d.

(1) orders the affected area evacuated (2) automatically stops upward movement of the manipulator hoist

REACTOR OPERATOR Page 71 QUESTION: 067 (1.00)

In the event that access to an area with radiation levels in excess of 1000 mrem/hour CANNOT be prevented using a locked door, Technical Specification 6.13 requires the area to be roped off and conspicuously posted.

Which one of the following lists the additional measure that fulfills the requirements of Technical Specifications for the entrance to this area?

a.

Install an audible alarm.

b.

Setup a control point.

c.

Install a flashing light.

d.

Setup a dose rate indicating device.

REACTOR OPERATOR Page 72 QUESTION: 068 (1.00)

Air Compressor A is operating when cooling water to the compressor is inadvertently isolated.

The air compressor will trip...

a.

due to low jacket water pressure.

b.

when the limit for oil temperature is exceeded.

c.

when the limit for air outlet temperature is exceeded.

d.

due to seal leakage resulting in low air discharge pressure.

REACTOR OPERATOR Page 73 QUESTION: 069 (1.00)

Given the following:

RCS Average Temperature = 547°F.

The reactor is critical at approximately 3% power.

The "B" Diesel Generator is inoperable.

The NORMAL power supply for pressurizer heater control group "A" was taken out of service to repair a breaker fault.

Which of the following describes the Technical Specification operability and required actions for the pressurizer heaters, if any?

a.

Technical Specifications requirements are MET and no action is required.

b.

Technical Specifications requirements are NOT met, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action is required to go to at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

Technical Specifications requirements are NOT met, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action is required to go to at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d.

Technical Specifications requirements are NOT met, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action is required to go to at least COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

REACTOR OPERATOR Page 74 QUESTION: 070 (1.00)

Which of the following uses Safeguard 125 VDC power as the NORMAL power supply?

a.

Bus 4 Circuit Breaker Control b.

7.5 KVA Inverter BRA-111 c.

Reactor Trip Breaker shunt trip coil d.

Non-interruptible Bus BRD-115

REACTOR OPERATOR Page 75 QUESTION: 071 (1.00)

Power is lost to BRB-104. Which component(s) associated with the 1B EDG will be affected by this condition?

A.

Field flash circuit AND jacket water pumps ONLY B.

Field flash circuit AND fuel oil priming pump ONLY C.

Jacket water pumps AND immersion heaters ONLY D.

Fuel oil priming pump AND starting air compressors ONLY

REACTOR OPERATOR Page 76 QUESTION: 072 (1.00)

Which of the following correctly describes the effect of a failure (HIGH) of R-15, Air Ejector Exhaust Monitor during a release?

(1)

Air Ejector Discharge Vent. AR-6/CV-31168 positions to DUCT (2)

S/G Blowdown Isolation valves CLOSE (3)

S/G Sample Isolation valves CLOSE (4)

Humidification Steam Inlet CV HS-17-1/CV31770 CLOSES a.

ONLY (1), (2) AND (3) occur b.

ONLY (2) AND (3) occur c.

(1), (2), (3) AND (4) occur d.

ONLY (2), (3) AND (4) occur

REACTOR OPERATOR Page 77 QUESTION: 073 (1.00)

Which of the following describes the CW condition(s) that would provide an interlock to PREVENT starting a CW pump?

(I)

Seal Water Flow < 2 gpm (II)

"Forebay Level Low Low" (566 or 42%)

(III)

Thrust Bearing Cooler Flow < 4 gpm a.

ONLY (II) b.

ONLY (I) and (III) c.

ONLY (II) and (III) d.

(I), (II) and (III)

REACTOR OPERATOR Page 78 QUESTION: 074 (1.00)

A malfunction of ONE of the "A" Diesel Generator Room CO2 temperature switches occurs, causing it to fail HIGH. Which of the following describes the response of the CO2 system to the "A" Diesel Generator Room?

a.

The CO2 actuation sequence will not begin until a second switch actuation occurs.

b.

The CO2 actuation sequence will sound a local horn, but will not discharge.

c.

The CO2 actuation sequence will sound a local horn and then discharge.

d.

The CO2 actuation sequence will start a local, flashing red light, sound a local horn and then discharge.

REACTOR OPERATOR Page 79 QUESTION: 075 (1.00)

Given the following conditions:

The plant is at 100% power.

All lineups/switch positions are in their NORMAL position.

Pressurizer Level Channel LT-426 (Channel I) fails LOW What is the status of the following BEFORE any operator actions are taken?

Letdown Flow Indication "Pressurizer Level Low" Annunciator a.

Normal LIT b.

Normal Not LIT c.

Zero LIT d.

Zero Not Lit

REACTOR OPERATOR Page 80 QUESTION: 076 (1.00)

Given the following conditions:

Steam Generator NR Levels are 88%

MSIVs are CLOSED Per Procedure FR-H.2, "Response to Steam Generator Overpressure," which of the methods given below has PRIORITY for decreasing S/G pressure?

a.

Dump steam using SG PORVs b.

Isolate AFW to the S/Gs c.

Dump steam using Steam Supply to Turbine-Driven AFW Pump d.

Dump steam using Main Steam Isolation Bypass Valves

REACTOR OPERATOR Page 81 QUESTION: 077 (1.00)

In addressing a PRZR relief valve (PORV) that is stuck open, the associated block valve must be closed. Which of the following indication(s) can be used to identify which PORV is stuck open?

(I)

PR-2A(B) indicating lights on the Mechanical Console C (II)

Acoustic monitor indicating lights on the Mechanical Console C (III)

Outlet temperatures for each PORV a.

ONLY (I) b.

ONLY (I) OR (III) c.

ONLY (II) OR (III) d.

(I), (II) OR (III)

REACTOR OPERATOR Page 82 QUESTION: 078 (1.00)

A LOCA has occurred and a controlled RCS cooldown and depressurization per ES-1.2, "Post LOCA Cooldown and Depressurization" is in progress. ALL ECCS equipment is OPERABLE.

RCS Pressure and Temperature is 1500 psig / 480°F. After SI pump A is secured as part of the RCS cooldown and depressurization, the following alarms occur:

47022-D, "CONTAINMENT HIGH PRESSURE SI" 47024-A, "ACCUMULATOR A PRESSURE HIGH/LOW" 47024-B, "ACCUMULATOR A LEVEL HIGH/LOW" What action(s) must be taken, if any, based on these conditions:

a.

Trip both RCPs.

b.

Restart SI pump A.

c.

Trip both RCPs AND Restart SI pump A.

d.

No action required.

REACTOR OPERATOR Page 83 QUESTION: 079 (1.00)

Which of the following systems is considered to be the most likely location for a rupture or break outside containment, and therefore is the system of primary concern during ECA-1.2, "LOCA Outside Containment?"

a.

Safety Injection b.

Residual Heat Removal c.

Component Cooling d.

Chemical and Volume Control

REACTOR OPERATOR Page 84 QUESTION: 080 (1.00)

Given the following conditions:

A loss of normal feedwater flow has occurred.

The actions of FR-S.1 "Response to Nuclear Power Generation/ATWS" must be performed due to a failure of the plant to trip Which of the following describes the proper sequence of steps to be taken with a failure of the reactor to trip, AFTER beginning to manually insert the Control Rods?

(I)

- Locally Open Reactor Trip Breakers (II)

- Open Bus 33 and Bus 43 supply breakers (III)

- TRIP Rod Drive MG Set Motor & Generator Circuit Breaker Control Switches a.

(I), (II), and THEN (III).

b.

(II), (I), and THEN (III).

c.

(II), (III), and THEN (I).

d.

(III), (II), and THEN (I).

REACTOR OPERATOR Page 85 QUESTION: 081 (1.00)

Given the following conditions:

Reactor power is 100%

VCT level transmitter LT-112 (24015) fails high (100%)

Which of the following describes what occurs if NO operator action is taken?

VCT level decreases __________.

a.

because auto makeup capacity is not able to maintain VCT level with letdown diverted b.

with NO auto makeup capability causing charging suction to shift to the RWST c.

until charging pumps lose suction and start to cavitate d.

until auto makeup starts and maintains VCT level

REACTOR OPERATOR Page 86 QUESTION: 082 (1.00)

Given the following conditions:

The plant is at 255°F, cooling down to Cold Shutdown with RHR Train A.

RHR Train B is out of service for testing.

Annunciator 47024-H, CC SURGE TANK LEVEL HIGH/LOW is LIT.

CC Surge Tank Level is 53% and INCREASING.

R-17, Component Cooling Liquid Rad Monitor, is in HIGH ALARM.

VCT level is DECREASING.

All other indications are NORMAL.

Which of the following is the location of the leak?

a.

RHR system.

b.

SFP heat exchanger.

c.

Seal Water heat exchanger.

d.

SW system.

REACTOR OPERATOR Page 87 QUESTION: 083 (1.00)

Complete the following statement:

Source Range neutron detectors operate in the ____(1)_____ region, so decreasing the detector voltage beyond calibration limits would result in a _____(2)_____ indicated power level.

a.

(1) Ionization, (2) higher b.

(1) Proportional, (2) higher c.

(1) Ionization, (2) lower d.

(1) Proportional, (2) lower

REACTOR OPERATOR Page 88 QUESTION: 084 (1.00)

The following conditions exist:

A reactor startup has been completed per N-CRD-49B, "Reactor Startup."

The Source Range trip is blocked.

The N35 Intermediate Range channel is failed LOW with the level trip bypassed.

The N36 Intermediate Range channel is reading ERRATICALLY.

Source Range counts have just reached 106 CPS What is the expected indication on the intermediate range nuclear instruments for this condition?

a.

10-3 % Power (IR) b.

10-2 % Power (IR) c.

10-1 % Power (IR) d.

100 or 1% Power (IR)

REACTOR OPERATOR Page 89 QUESTION: 085 (1.00)

The plant is at 100% power. TLA-15, RMS ABOVE NORMAL is in alarm due to increasing radiation level on R-19, S/G Blowdown Liquid Monitor. What action(s) must be taken based on these conditions?

a.

IF the radiation level on R-19, S/G Blowdown Liquid Monitor increases to HIGH alarm, THEN go to E-0-14, "Steam Generator Tube Leak."

b.

Go to E-0-14, "Steam Generator Tube Leak" and perform Operator immediate actions.

c.

Go to A-RM-45, "Abnormal Radiation Monitoring System" and verify the automatic actions occur as listed for R-19, S/G Blowdown Liquid Monitor.

d.

Per A-RM-45, "Abnormal Radiation Monitoring System" determine primary-to-secondary leak rate using "R-19 to Leakage Rate Conversion Graph."

REACTOR OPERATOR Page 90 QUESTION: 086 (1.00)

Which of the following describes the reason for tripping both RXCPs, if required, per step 1 of E-3, "Steam Generator Tube Rupture?"

a.

To minimize the potential for RCP damage when an RCS depressurization is initiated.

b.

To minimize the heat input when a controlled RCS cooldown is initiated.

c.

To prevent the automatic opening of a pressurizer PORV.

d.

To prevent unnecessary RCS water depletion.

REACTOR OPERATOR Page 91 QUESTION: 087 (1.00)

The following plant conditions exist:

FR-H.1, Response to Loss of Secondary Heat Sink is in progress.

The CST is unavailable.

Yarway wide range S/G levels are at 20%.

RCS pressure is at 2200 psig.

Containment pressure is 1 psig.

Which of the following heat removal methods is available, if any, before RCS bleed and feed is required AND what is the preferred sequence for establishing flow to at least one S/G?

(1)

Depressurize SG and establish Condensate flow (2)

Establish AFW flow using Service Water (3)

Establish Main Feedwater flow a.

(2), (3), (1) b.

(3), (2), (1) c.

(3), (1), (2) d.

No S/G heat removal method is available; RCS bleed and feed is required immediately.

REACTOR OPERATOR Page 92 QUESTION: 088 (1.00)

Which of the following places the plant in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Limiting Condition of Operation per Technical Specifications?

a.

BRA-101, Station Battery A, fuse blows.

b.

BRA-108, Battery Charger A, damaged due to fire.

c.

BRA-102, DC DIstribution Train A, damaged bus bar.

d.

BRA-111, Instrument Bus 1 Inverter, damaged rectifier.

REACTOR OPERATOR Page 93 QUESTION: 089 (1.00)

The following plant conditions exist:

An accidental gaseous release has occurred.

The derived air concentration (DAC) of this release is 4 DAC.

Which of the following is the expected exposure to the whole body of a worker breathing air in this area for 30 minutes?

a.

2 mrem b.

5 mrem c.

8 mrem d.

10 mrem

REACTOR OPERATOR Page 94 QUESTION: 090 (1.00)

The plant is operating at 100% power. Annunciator 47033-35, TLA-15, RMS ABOVE NORMAL, alarms due to rising count rate on R-42, S/G A N16 Monitor.

Plant conditions:

Pressurizer level:

47%, stable.

Pressurizer pressure:

2235 psig.

Which of the following describes the action or actions required for this situation?

a.

Enter E-0-14, Steam Generator Tube Leak.

b.

Manually trip the reactor and enter E-0, Reactor Trip or Safety Injection.

c.

Contact Health Physics to assist in identifying the radiation source.

d.

Evacuate the reactor building.

REACTOR OPERATOR Page 95 QUESTION: 091 (1.00)

Given the following conditions:

A LOCA has occurred The crew is performing a cooldown per ES-1.2 " Post LOCA Cooldown and Depressurization" Two Containment Cooling Fan Coil Units are running Containment pressure is stable at 2.2 psig A transition to FR-Z.3 "Response to High Containment Radiation Level" is made on a YELLOW path condition Why does FR-Z.3 start idle Containment Cooling Fan Coil Units?

a.

To remove radioactive particulates during condensation of water vapor.

b.

To remove radioactive gases during condensation of water vapor.

c.

To support Containment Purge and Vent Subsystem Exhaust filtration.

d.

To support Containment Purge and Vent Subsystem Purge filtration.

REACTOR OPERATOR Page 96 QUESTION: 092 (1.00)

Given the following conditions:

A fire has occurred on site.

E-0-06, "Fire in Alternate Fire Zone" is being implemented.

Complete the following statement:

During implementation of E-0-06, only ____(1)____ equipment is being controlled from the Dedicated Shutdown Panel and offsite power is considered to be ____(2)____.

____(1)____

____(2)____

a.

Train A available b.

Train A NOT available c.

Train B available d.

Train B NOT available

REACTOR OPERATOR Page 97 QUESTION: 093 (1.00)

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

0100: RC-413, Pressurizer Liquid Sampling Isolation valve is determined to be INOPERABLE.

0200: RC-412, Pressurizer Liquid Sampling Isolation valve is determined to be INOPERABLE.

What log entry or entries need to be made to track these inoperable valves?

a.

An entry for each valve in the Shift Managers Log AND in the Shift Managers LCO Tracking Log at the time they became INOPERABLE.

b.

An entry for each valve in the Control Room Log AND in the Shift Managers LCO Tracking Log at the time the valves were discovered to be INOPERABLE.

c.

An entry in the Shift Managers Log AND the Control Room Log for each valve at the time they became INOPERABLE, AND an entry in the Control Room Shift Turnover Checklist at shift turnover.

d.

One log entry for both valves in the Control Room Shift Turnover Checklist at shift turnover AND an entry for each valve in the Periodic Daily Log at the time each valve became INOPERABLE.

REACTOR OPERATOR Page 98 QUESTION: 094 (1.00)

Given the following:

A LOCA has occurred.

Containment Pressure is 6 psig.

Core Exit Thermocouples are at 600°F.

RCS pressure is 200 psig.

RHR is in its at-power lineup.

FR-C.3, Response to Saturated Core Cooling, is being implemented.

What flows to the RCS must be verified per FR-C.3, Response to Saturated Core Cooling?

a.

Charging pump flow ONLY.

b.

RHR and SI pump flows ONLY.

c.

SI pump flow ONLY.

d.

Charging and SI pump flows ONLY.

REACTOR OPERATOR Page 99 QUESTION: 095 (1.00)

The following conditions exist:

A runback from 80% to 60% power occurred 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago.

Chemistry samples of the RCS indicate high dose-equivalent I-131.

Why is it desirable to increase letdown flow through the CVC mixed bed demineralizers to 80 gpm under these conditions?

a.

To reduce RCS activity.

b.

To control RCS pH.

c.

To reduce RCS corrosion products.

d.

To control RCS boron concentration.

REACTOR OPERATOR Page 100 QUESTION: 096 (1.00)

While performing ECA-1.1, Loss of Emergency Coolant Recirculation, the "RHR Pump A Supply to ICS Pump A", valve RHR-400A could not be operated from the control room. The steps contingency action states "Locally open valve". On which elevation of the auxiliary building is this valve located?

a.

572 b.

586 c.

606 d.

626

REACTOR OPERATOR Page 101 QUESTION: 097 (1.00)

In order to establish a Containment Purge in HOT SHUTDOWN, which of the following is required?

1.

Notify NRC prior to opening 36" RBV valves.

2.

Obtain a Gaseous Waste Discharge Permit.

3.

Verify Annunciator 47051-B, "Containment Vent High Radiation Disabled" is CLEAR.

a.

ONLY 1 and 2.

b.

ONLY 1 and 3.

c.

ONLY 2 and 3.

d.

1, 2 and 3.

REACTOR OPERATOR Page 102 QUESTION: 098 (1.00)

For a Steam Line Break of a given size and location, which of the following initial conditions results in the smallest reactivity rate of change immediately after the break?

CORE BURNUP (MWD/MTU)

RCS Tavg a.

9000 450°F b.

9000 547°F c.

5000 450°F d.

5000 547°F

REACTOR OPERATOR Page 103 QUESTION: 099 (1.00)

Given the following:

Reactor power is stabilized at the eight-fold power level.

The Eight-Fold Critical Rod Position is determined to be 65 steps on Control Bank C.

Which action is required in this situation?

a.

Emergency Borate 300 gallons.

b.

SHUT DOWN the reactor per N-CRD-49C, "Reactor Shutdown" c.

Get permission from Station Nuclear Engineer to continue with the startup.

d.

Verify the Eight-fold Critical Rod Position is within +400pcm of ECP

REACTOR OPERATOR Page 104 QUESTION: 100 (1.00)

Given the following:

The plant is in normal 100% power operations.

Containment Fan Coil Units Emergency Discharge Dampers RBV-150 A and B both fail OPEN.

What is the major concern at this time?

a.

Damage to the Nuclear Instrumentation.

b.

Damage to the Reactor Vessel Gap.

c.

RXCP A motor stator overheating.

d.

RXCP B motor stator overheating.

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

REACTOR OPERATOR

REFERENCES:

Page 105 ANSWER: 001 C

REFERENCE:

ECA-0.0, CAUTION prior to Step 6; BYRON1 10/29/2001 Exam Bank Higher 000062K303..(KAs)

ANSWER: 034 D

REFERENCE:

A-FW-05A, Abnormal FW Operation, step 4.4.

New Higher 059000 2.2.2..(KAs)

ANSWER: 067 C

REFERENCE:

LO R01-01-LPTS4.010; Tech Spec 6.1; KNPP EQB Bank Memory 2.1.10..(KAs)

ANSWER: 002 B

REFERENCE:

LP RO2-02-LP05A, Main Feedwater; SD 05A, Feedwater New Higher 059000K408..(KAs)

ANSWER: 035 A

REFERENCE:

SD 39, 4160 V System; SD 36, RCS New Memory 003000K201..(KAs)

ANSWER: 068 C

REFERENCE:

OP A-AS-1, Abnormal Station/

Instrument Air System Operation; SD #1, Station/Instrument Air, Pg 9.

Bank Memory 078000K403..(KAs)

ANSWER: 003 A

REFERENCE:

NRC EQB; KNPP 02/21/1994 Exam; ADV-SYS-LP-36D, p. 11; ADV-SYS-LP-36D, EO-RO4.a Bank Higher 022000K302..(KAs)

ANSWER: 036 A

REFERENCE:

IPEOP Background Document for FR-P.1; LP RO4-04-LP-016, Response to Imminent Pressurized Thermal Shock Condition New Memory E08EA202..(KAs)

ANSWER: 069 B

REFERENCE:

LP RO2-01-LP-36B, PZR and PRT; Tech Spec 3.1.a.6 and its basis; SD 38, DC and Emergency AC Power New Higher 062000 2.1.12..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 106 ANSWER: 004 D

REFERENCE:

NRC EQB; Prairie Island 1 and 2 05/09/1994 Exam.

Modified Memory 056000K103..(KAs)

ANSWER: 037 A

REFERENCE:

ECA-2.1, Uncontrolled Depressurization of Both S/Gs New Higher E12EA13..(KAs)

ANSWER: 070 C

REFERENCE:

SD 38, DC and Emergency AC Electrical Distribution.

New Memory 063000K201..(KAs)

ANSWER: 005 B

REFERENCE:

E-240 Rev. AQ; Adv System LP Objective R02-05-LP36D.003 ("N/A).

New Memory 011000K202..(KAs)

ANSWER: 038 A

REFERENCE:

ES-1.1, SI Termination and Background Document; LP RO4-04-LP-005 New Memory 000026K303..(KAs)

ANSWER: 071 B

REFERENCE:

SD 38, DC and Emergency AC Power; LP RO2-03-LP-042A, D/Gs New Memory 064000K202..(KAs)

ANSWER: 006 C

REFERENCE:

KNPP SD 1, Rev 1, p 6; LP O-RO-LP-2.11.1, EO 3; NRC EQB; KNPP 1993 Exam Modified Higher 079000K101..(KAs)

ANSWER: 039 C

REFERENCE:

ES-0.2, Natural Circulation Cooldown Modified Higher 2.4.4..(KAs)

ANSWER: 072 C

REFERENCE:

SD 45, Radiation Monitors; A-RM-45, Abnormal Radiation Monitoring; E-3748, PRM Integrated Logic Diagram New Memory 073000K301..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 107 ANSWER: 007 D

REFERENCE:

NRC EQB; V. C. Summer 1 1992/05/18 Exam Bank Higher 015000K101..(KAs)

ANSWER: 040 C

REFERENCE:

SD 49 Rod Control and RPI; Tech Specs New Higher 000005K105..(KAs)

ANSWER: 073 C

REFERENCE:

SD 4, CW System; ARP 47051-N, Forebay Level Low; LP RO2-02-LP-004, CW New Memory 075000K401..(KAs)

ANSWER: 008 D

REFERENCE:

NRC EQB; North Anna 1 01/26/1996 Exam.

Bank Higher 000065K304..(KAs)

ANSWER: 041 C

REFERENCE:

ES-0.2, Rev 0, Caution before step 1, p 2.;

NRC EQB; Point Beach 04/29/1991.

Bank Memory 000015A109..(KAs)

ANSWER: 074 C

REFERENCE:

SD 8, Fire Protection System; RO2-02-LP-008, Fire Protection System; New Memory 086000K604..(KAs)

ANSWER: 009 A

REFERENCE:

NRC EQB; Prairie Island 1 and 2 09/28/1992 Exam.

Bank Higher 000003K103..(KAs)

ANSWER: 042 A

REFERENCE:

PWR Fundamentals; Braidwood 4/1/1996 Exam Bank Memory 000024K102..(KAs)

ANSWER: 075 A

REFERENCE:

SD 36, Reactor Coolant System; ARP 47043-F, PRZR Level Low; A-MI-87, B/S Tripping for Failed RP or Safeguards Inst.

New Memory 000028A206..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 108 ANSWER: 010 B

REFERENCE:

BYRON1 10/29/2001 Exam.

Modified Higher 000027K203..(KAs)

ANSWER: 043 D

REFERENCE:

LP RO2-02-LP-003, Condensate and Air Removal System; New Higher 056000A204..(KAs)

ANSWER: 076 A

REFERENCE:

LP RO4-04-LP-036; FR-H.2, Response to S/G Overpressure; IPEOP Background Document New Memory E13EK11..(KAs)

ANSWER: 011 B

REFERENCE:

ES-1.3, Transfer to Cntnmt Sump Recirc; SD 33, SI System.

New Memory 006000 2.3.11..(KAs)

ANSWER: 044 B

REFERENCE:

RO2-02-LP-003.004; O-AOP-LP-D8; E-AR-09, Loss of Condenser Vacuum New Higher 000051 2.1.30..(KAs)

ANSWER: 077 A

REFERENCE:

LP RO4-04-LP-36B ARP 47042-A, PZR PORV Open; ARP 47042-B, PZR PORV Discharge Temperature High; New Higher 000008A203..(KAs)

ANSWER: 012 A

REFERENCE:

KNPP EQB; LP Obj RO2-01-LP31.004; CCW SD 31.

Bank Higher 008000A308..(KAs)

ANSWER: 045 A

REFERENCE:

SD 32B, Gaseous Radioactive Waste Disposal; New Memory 071000A302..(KAs)

ANSWER: 078 B

REFERENCE:

ES-1.2, Post LOCA Cooldown and Depressurization, Step 16a (Contingency Actions)

New Higher 000009 2.4.45..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 109 ANSWER: 013 A

REFERENCE:

LP RO2-02-LP362; SD 36, RCS; KNPP 12/2000 Exam.

Bank Higher 003000A304..(KAs)

ANSWER: 046 A

REFERENCE:

SD 32B, Gaseous Radioactive Waste Disposal; SD 45, Radiation Monitoring New Memory 071000A409..(KAs)

ANSWER: 079 B

REFERENCE:

ECA-1.2, LOCA Outside Containment; IPEOP Background Document; Prairie Island 05/15/2000 Exam; LP RO4-04-LP-020, LOCA Outside Containment Modified Memory E04EK22..(KAs)

ANSWER: 014 A

REFERENCE:

SD 05A, Feedwater System; SD 05B, Auxiliary Feedwater System; KNPP EQB.

Bank Higher 061000K101..(KAs)

ANSWER: 047 A

REFERENCE:

SD 45, Radiation Monitoring; RO2-01-LP045, Radiation Monitoring New Memory 072000K501..(KAs)

ANSWER: 080 B

REFERENCE:

FR-S.1, Response to Nuclear Power Generation/ATWS; IPEOP Background Document; New Memory 000029 2.4.49..(KAs)

ANSWER: 015 A

REFERENCE:

KNPP Exam 12/18/1997; SD 035, CVCS.

Bank Memory 004000A405..(KAs)

ANSWER: 048 D

REFERENCE:

SD 45, Radiation Monitoring; A-RM-45, Abnormal Radiation Monitoring New Memory 072000 2.3.10..(KAs)

ANSWER: 081 C

REFERENCE:

KNPP NRC Exam 10/24/2000 ARP 47043-L, VCT Level High/Low SD 35, CVCS Bank Memory 000022A108..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 110 ANSWER: 016 B

REFERENCE:

LP RO2-05-LP05B, Aux Feedwater; KNPP EQB Bank Higher 061000A101..(KAs)

ANSWER: 049 C

REFERENCE:

ES-0.2, Natural Circulation Cooldown; ES-0.2 Background Document; Bank Memory 002000A203..(KAs)

ANSWER: 082 A

REFERENCE:

A-CC-31, Abnormal CCW Operations; ARP 47024-H, CC Surge Tank Level High/Low; A-RHR-34, Abnormal RHR Operations; A-RM-45, Abnormal Rad Monitoring New Higher 000025A202..(KAs)

ANSWER: 017 D

REFERENCE:

SP-54-086, TSV and GV Operability Test, Pg 3 of 17.

New Memory 045A401..(KAs)

ANSWER: 050 D

REFERENCE:

RO2-05-LP-36C, Pressurizer Pressure Control New Higher 010000K103..(KAs)

ANSWER: 083 C

REFERENCE:

LP RO2-05-LP048, Excore Nuclear Instrumentation System; SD 48, Excore Nuclear Instrumentation New Memory 000032K101..(KAs)

ANSWER: 018 B

REFERENCE:

Byron 10/29/2001 Exam Bank Higher 2.1.25..(KAs)

ANSWER: 051 C

REFERENCE:

RO2-05-LP472, Reactor Protection New Higher 012000K304..(KAs)

ANSWER: 084 A

REFERENCE:

LP RO2-05-LP048, Excore Nuclear Instrumentation System; SD 48, Excore Nuclear Instrumentation New Memory 000033A201..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 111 ANSWER: 019 C

REFERENCE:

LP R02-05-LP-033; TS 3.3.a. Accumulators New Higher 2.1.11..(KAs)

ANSWER: 052 C

REFERENCE:

LP RO4-04-LP040, Loss of all AC Power; Comanche Peak 11/26/90; ECA-0.0, Loss of all AC Power Bank Memory 000055K101..(KAs)

ANSWER: 085 B

REFERENCE:

A-RM-45, Abnormal Radiation Monitoring System; E-0-14, Steam Generator Tube Leak New Memory 000037A113..(KAs)

ANSWER: 020 A

REFERENCE:

LP R02-05-LP-053; N-FH-53-CLC, Pre-Refueling Checklist; RF-01.00, KNPP Refueling Procedure; New Higher 2.2.27..(KAs)

ANSWER: 053 D

REFERENCE:

LP RO2-05-36C, Pzr Pressure Control; SD 36, RCS; A-MI-87, Bistable Tripping for Failed RP or Safeguards Inst.

New Higher 000057A102..(KAs)

ANSWER: 086 D

REFERENCE:

BKG E-3, Steam Generator Tube Rupture; IPEOP Background Document; E-3, Steam Generator Tube Rupture New Memory 000038K308..(KAs)

ANSWER: 021 C

REFERENCE:

LP R02-05-LP-053; NAD-02.07, KNPP Refueling Operations New Memory 2.2.26..(KAs)

ANSWER: 054 C

REFERENCE:

KNPP SD 49, Rod Control and RPI; KNPP EQB RO2-05-LP049.004 010; LP RO2-05-LP049, Rod Control and RPI.

Bank Higher 014000K502..(KAs)

ANSWER: 087 C

REFERENCE:

A-FW-05B, Abnormal AFW System Operation; FR-H.1, Response to Loss of Secondary Heat Sink; BKG FR-H.1, Loss of Secondary Heat Sink; IPEOP Background Document New Higher E05EK22..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 112 ANSWER: 022 C

REFERENCE:

12/11/2000 KNPP Exam.

Modified Higher 2.3.2..(KAs)

ANSWER: 055 C

REFERENCE:

Fire Protection Program Plan, Appendix D; E-0-06, Fire in Alternate Fire Zone; E-FP-08, EOP - Fire.

New Memory 016000K401..(KAs)

ANSWER: 088 C

REFERENCE:

Tech Spec and Bases; A-EDC-38, Abnormal DC Supply and Distribution System; SD 38, "DC and Emergency AC Distribution; LP RO2-03-LP 038, DC and Emergency AC Distribution New Memory 000058 2.2.22..(KAs)

ANSWER: 023 B

REFERENCE:

FR-S.1, Response to a Nuclear Power Generation/ATWS; E-0, Reactor Trip or Safety Injection.

New Higher 2.4.1..(KAs)

ANSWER: 056 C

REFERENCE:

FR-Z.1, Response to High Ctnmt Pressure, Step 3, pgs 3-4; CS Integrated Logic Diag E-1604; System Integrated Logic Diagram ICS E-2012.

Bank Memory 026000A301..(KAs)

ANSWER: 089 B

REFERENCE:

10CFR20, definitions and part 1204; SD 32B, Gaseous Radioactive Waste Disposal; New Higher 000060K102..(KAs)

ANSWER: 024 B

REFERENCE:

Byron Exam 10/20/2000; ECA-0.0, LOSS OF ALL AC POWER and Background Document.

New Higher 2.4.7..(KAs)

ANSWER: 057 A

REFERENCE:

A-RM-45, Abnormal Radiation Monitoring System.

New Memory 029000A102..(KAs)

ANSWER: 090 A

REFERENCE:

ARP 47033-35, TLA-15, RMS ABOVE NORMAL; New Memory 000061AK30..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 113 ANSWER: 025 C

REFERENCE:

FR-C.2, Response to Degraded Core Cooling.

Bank Memory E11EK32..(KAs)

ANSWER: 058 D

REFERENCE:

Byron Exam 06/29/2000.

Bank Memory 103000A101..(KAs)

ANSWER: 091 A

REFERENCE:

FR-Z.3, Response to High Containment Radiation Level; BKG FR-Z.3,Response to High Containment Radiation Level; IPEOP Background Document; KNPP NRC Exam 12/11/2000 Modified Higher E16EK33..(KAs)

ANSWER: 026 B

REFERENCE:

LP RO2-05-LP 049, Rod Control and RPI; SD-49, Section 3.3.2.

New Higher 001000K103..(KAs)

ANSWER: 059 B

REFERENCE:

KNPP Exam 02/24/1997 Modified Memory 033000A203..(KAs)

ANSWER: 092 B

REFERENCE:

E-0-06, Fire in Alternate Fire Zone; Byron Exam 06/29/2000 New Higher 000067A217..(KAs)

ANSWER: 027 B

REFERENCE:

RO2-05-LP 049, Rod Control and RPI, pg 43 and 44; ARP 47043-R, Rod Control System Abnormal New Memory 001000K611..(KAs)

ANSWER: 060 B

REFERENCE:

Beaver Valley 2 Exam 03/17/1997; Modified Higher 035000A301..(KAs)

ANSWER: 093 C

REFERENCE:

LO R02-04-LP056.007; Tech Spec 3.6.b.3.B, Containment System New Memory 000069 2.2.23..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 114 ANSWER: 028 D

REFERENCE:

KNPP SD 55, ESF Bank Memory 013000K403..(KAs)

ANSWER: 061 C

REFERENCE:

KNPP SD 05B, AFW System; ARP 47062-N Bank Higher 039000A404..(KAs)

ANSWER: 094 C

REFERENCE:

FR-C.3, Response to Saturated Core Cooling; New Higher E07EK13..(KAs)

ANSWER: 029 A

REFERENCE:

LP RO2-05-035, CVCS; SD 35, CVCS.

New Higher 004000K305..(KAs)

ANSWER: 062 C

REFERENCE:

KNPP 12/11/00 NRC Exam; SD 09, Air Removal System Bank Higher 055000K106..(KAs)

ANSWER: 095 A

REFERENCE:

A-RC-36A, High Reactor Coolant Activity; KNPP NRC Exam 12/18/1997 Bank Higher 000076K305..(KAs)

ANSWER: 030 A

REFERENCE:

LP RO2-05-048, NIS; SD 48, NIS; 11/13/1990 Millstone 3 Exam.

Bank Higher 015000K603..(KAs)

ANSWER: 063 A

REFERENCE:

A-RM-45, Abnormal Radiation Monitoring System; LP Obj RO2-01-LP045.004; Dwg XK-100-131; FD Waste Disposal System.

New Memory 068000A204..(KAs)

ANSWER: 096 B

REFERENCE:

SD 23, ICS System; Dwg A-204, A-210, OPERXK-100-18; Dwg OPERM-217; New Memory 026000 2.4.34..(KAs)

REACTOR OPERATOR

REFERENCES:

Page 115 ANSWER: 031 D

REFERENCE:

SD 50, Incore Instrumentation; Inadequate Core Cooling Monitor, Sect. 1.4, CET Operation, and 3.4.3, CET Instrumentation.

New Memory 017000A401..(KAs)

ANSWER: 064 D

REFERENCE:

SD RHR pg 16; Dwgs OPERXK-100-18, -19.

New Memory 005000K101..(KAs)

ANSWER: 097 C

REFERENCE:

SD 18, Reactor Building Ventilation; ARP 47051-B, Containment Vent High Radiation Disabled; N-RBV-18B, Reactor Bldg Vent System Cold Operation and Releases New Higher 029000 2.3.9..(KAs)

ANSWER: 032 C

REFERENCE:

SD 18, RBV; Hydrogen Control; LP RO2-04-LP 18, RBV System.

New Higher 022000A301..(KAs)

ANSWER: 065 A

REFERENCE:

N-RBV-18C, POST-LOCA Hydrogen Control.

New Memory 028000K502..(KAs)

ANSWER: 098 C

REFERENCE:

KNPP Exam 12/18/97; Reactor Data Manual Bank Higher 000040 2.2.34..(KAs)

ANSWER: 033 D

REFERENCE:

SD 2, Service Water; KNPP EQB.

Bank Higher 076000K307..(KAs)

ANSWER: 066 B

REFERENCE:

E-RH-53A, "Dropped or Damaged FA,8/17/2001; SD Rad Mon, pgs 12-13; SD FH, pgs 13,16, 28.

New Higher 034000A201..(KAs)

ANSWER: 099 B

REFERENCE:

N-CRD-49B, Reactor Startup; N-CRD-49C, Reactor Shutdown; New Higher 2.2.1..(KAs)

ANSWER: 100 D

REFERENCE:

KNPP Exam 12/11/2000; SD 18, Reactor Building Vent Bank Higher 2.1.32..(KAs)

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