ML072600302

Ro/Sro August 2007 Initial Retake Examination

ML072600302
Person / Time
Site:	Perry
Issue date:	08/06/2007
From:	Division of Reactor Safety III
To:
References
50-440/07-302
Download: ML072600302 (77)

Text

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

Date: August 6, 2007 Facility/Unit: Perry Nuclear Power Plant Region: I II III x IV Reactor Type: W CE BW GE x Start Time: 0800 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.

______________________________________

Applicants Signature Results Examination Value 75 Points Applicants Score __________ Points Applicants Grade __________ Percent

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

Date: August 6, 2007 Facility/Unit: Perry Nuclear Power Plant Region: I II III x IV Reactor Type: W CE BW GE x Start Time: 0800 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.

______________________________________

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

REACTOR OPERATOR Page 3 QUESTION: 001 (1.00)

At 100% power Reactor Recirculation Pump A trips to off. IOI-0003, Power Changes, requires that power be reduced to # 2500 Mwt.

Why is power reduced to # 2500 Mwt?

a. This is the license limit for single loop operation.

b. To provide a margin to the license limit for single loop operation.

c. To provide a temporary limit until APLHGR and MCPR are modified for single loop operation.

d. To provide a temporary limit until required RPS instrumentation is reset for single loop operation.

QUESTION: 002 (1.00)

A Station Blackout (SBO) has occurred.

APRM neutron flux indication is available by meters and downscale lights on panels 1H13-P669, P670, P671 and P672.

These instruments are available because they are powered from . . .

a. ATWS Uninterruptible Power Supply

b. Class 1E Instrument Panel Power Supply

c. RPS Distribution Power Supply

d. TSC Uninterruptible Power Supply

REACTOR OPERATOR Page 4 QUESTION: 003 (1.00)

RHR Pump A is running when a loss of 125 VDC breaker control power occurs.

Which one of the following describes the operational impact that the loss of DC control power has on RHR Pump A circuit breaker?

a. The breaker will trip on a fault and can be tripped from the Control Room.

b. The breaker will trip on a fault but cannot be tripped from the Control Room.

c. The breaker will not trip on a fault but can be tripped from the Control Room.

d. The breaker will not trip on a fault and cannot be tripped from the Control Room.

QUESTION: 004 (1.00)

Why does a Main Generator Lockout Relay 86 device trip also directly cause a Main Turbine trip?

a. Prevent stator overheating

b. Provide overspeed protection

c. Prevent last stage bucket erosion

d. Provide reverse power protection

REACTOR OPERATOR Page 5 QUESTION: 005 (1.00)

The plant is operating at 10% power in MODE 2. The main turbine is rolling at 1800 rpm. The Reactor scrams and the operator notes the following after the scram announcement:

- Main Turbine is tripped

- Reactor Pressure is 1000 psig and lowering

- Reactor Level peaked at 220" and lowering

- Condenser Vacuum is 21" HgA and degrading The only operator action was Mode Switch to Shutdown.

Which one of the following conditions caused the reactor scram?

a. main turbine trip signal

b. high reactor pressure signal

c. high reactor water level signal

d. MSIV closure signal QUESTION: 006 (1.00)

RHR B was operating in Suppression Pool Cooling when the Control Room was evacuated due to a fire. The Unit Supervisor directs RHR B to remain in Suppression Pool Cooling in preparation for SRV cycling.

Which of the following is correct for aligning RHR B in Suppression Pool Cooling and why?

a. Operate the Division 2 ECC/ESW control switches in order to isolate the Control Room.

b. Operate the Division 2 ECC/ESW control switches in order to place components in required position.

c. Do not operate the Division 2 ECC/ESW control switches since this could disrupt system operation.

d. Do not operate the Division 2 ECC/ESW control switches since Control Room isolation is not required.

REACTOR OPERATOR Page 6 QUESTION: 007 (1.00)

The following conditions exist:

- Plant is in Mode 3

- Reactor Pressure is 50 psig and lowering

- RHR A is operating in Shutdown Cooling

- Reactor Recirculation Pump A and B are operating in slow speed One of the two operating NCC Pumps trips and the following alarms are received:

- RCIRC A and B Seal CLR Flow LO

- RCIRC A and B Upper BRG Flow LO

- RCIRC A and B Motor CLR Flow LO The Reactor Recirculation Pumps:

a. are required to be shutdown immediately.

b. are required to be shutdown when the motor winding temperature alarm is received at 240°F.

c. may be run indefinitely provided that CRD seal injection is maintained.

d. may be run until continuous motor winding temperature is > 248°F.

REACTOR OPERATOR Page 7 QUESTION: 008 (1.00)

Instrument Air Header Pressure is 85 psig and slowly lowering. The Unit Supervisor is operating per ONI-P52, "Loss of Service and/or Instrument Air." An Operator is performing air leak isolation per attachment 1.

The following plant conditions exist:

- Initial plant lineup has all of the A train filters and dryers in service.

- 2P52-F210 IA SHUTOFF TO UNIT 1 & COMMON LOADS is open

- 1P52-F210 IA SHUTOFF TO UNIT 1 & COMMON LOADS is closed

- 1P52-F810A IA AFTERFILTER A OUTLET TO STAINLESS SYSTEM is closed

- Unit 1 Instrument Air Pressure is 90 psig and increasing The instrument air leak is in which header?

Reference Provided - ONI-P52 Attachment 1-AIR LEAK ISOLATION

a. Parallel Air Header

b. Unit 1 Instrument Air Header

c. Unit 2 Instrument Air Header

d. CC/DGB Instrument Air Header

REACTOR OPERATOR Page 8 QUESTION: 009 (1.00)

The following plant conditions exist:

- The plant is in Cold Shutdown.

- Both Reactor Recirculation Pumps are shutdown.

- RHR Loop A' is in the Shutdown Cooling mode.

Which one of the following describes the importance of maintaining reactor water level greater than 245" if Shutdown Cooling is lost?

Maintaining reactor water level greater than 245" will . . .

a. prevent a low reactor water level scram signal when a Reactor Recirculation Pump is started.

b. prevent reactor coolant thermal stratification by ensuring natural circulation flow is maintained.

c. provide an adequate margin to "time to boil" point while starting the opposite loop of Shutdown Cooling.

d. provide an adequate vessel inventory for alternate methods of decay heat removal that utilize feed and bleed evolutions.

REACTOR OPERATOR Page 9 QUESTION: 010 (1.00)

The plant is in Mode 5 with fuel handling operations in progress. The following plant conditions exist:

- All Control Rods are fully inserted

- 1/2 of Core Reload is complete

- RHR Shutdown Cooling is secured to shift from RHR A to RHR B Loop

- Upper Pool Level is 22'9" above the RPV flange

- Upper Pool Temperatures is 65°F

- SRM Count rates are: - A -- 7 cps - B -- 5 cps - C -- 3 cps - D 9 cps Which one of the following actions is required to be performed based on the above conditions?

Suspend Fuel Movement ____

a. until Upper Pool temperature is greater that or equal to 68°F.

b. until an RHR loop is in Shutdown Cooling.

c. in SRM quadrant C, until SRM C is Operable.

d. until Upper Pool level is greater than or equal to 23' above the RPV flange.

REACTOR OPERATOR Page 10 QUESTION: 011 (1.00)

The Technical Specification limitation on the Drywell to Primary Containment differential pressure is (1) , and the bases of the positive upper limit is to ensure (2) .

a. (1) $ -0.1 and # 1.0 psid (2) that vent clearing does not occur during normal operation.

b. (1) $ -0.1 and # 1.0 psid (2) the design value of drywell pressure is not exceeded during a LOCA.

c. (1) $ -0.5 and # 2.0 psid (2) that vent clearing does not occur during normal operation.

d. (1) $ -0.5 and # 2.0 psid (2) the design value of drywell pressure is not exceeded during a LOCA.

REACTOR OPERATOR Page 11 QUESTION: 012 (1.00)

An automatic reactor scram occurred and all control rods fully inserted. The operator observes the following plant parameters:

- Reactor pressure increased to 1105 psig.

- Reactor pressure then decreased to 915 psig.

- Reactor pressure is currently 935 psig and increasing.

- Condenser Vacuum is 20.5" HgA and degrading.

Which one of the following describes the current method of reactor pressure control, including the bases for this method?

Reactor pressure is being controlled by the . . .

a. Low-Low Set SRV(s) to reduce the number of valves cycling thus prolonging valve life.

b. Low-Low Set SRV(s) to allow the RPS system to be reset following a high reactor pressure scram.

c. Main Turbine Bypass Valve(s) to minimize the loss of reactor coolant inventory through the SRVs.

d. Main Turbine Bypass Valves to minimize the heat addition to the Suppression Pool through the SRVs.

REACTOR OPERATOR Page 12 QUESTION: 013 (1.00)

The following plant conditions exist following a scram from 100% power.

- All rods in

- Reactor Pressure 649 psig and slowly lowering

- Reactor Level is 48" and slowly lowering

- Containment and Drywell Pressure 2.0 psig and slowly increasing

- Suppression Pool Level 17.6' and slowly increasing

- Suppression Pool Temperature 96°F and slowly increasing

- Loss of all high pressure injection systems

- All low pressure ECCS systems are operating on minimum flow.

- RFBPs operating on minimum flow It is required to operate RHR A and B ______.

a. in Containment Spray

b. in Suppression Pool Cooling

c. lined up for injection in preparation for maintaining adequate core cooling

d. with one loop in Containment Spray and the other in Suppression Pool Cooling QUESTION: 014 (1.00)

A Reactor scram has occurred with a leak from the scram discharge volume. Containment pressure and temperature are increasing. Which one of the following containment conditions requires all available containment cooling fans operated?

a. Pressure 1.5 psig

b. Pressure 2.25 psig

c. Temperature 90°F

d. Temperature 100°F

REACTOR OPERATOR Page 13 QUESTION: 015 (1.00)

Given the following plant conditions following a LOCA:

- RPV pressure 900 psig

- Drywell temperature 300°F

- Containment temperature 180°F Of the following, which one is the lowest Wide Range indicated level that could be used to determine RPV Level?

Reference provided - PEI-SPI Supplement Figure 2a Wide Range Level

a. 35"

b. 23"

c. 15"

d. 8" QUESTION: 016 (1.00)

Plant Conditions are as follows:

- Reactor Power 0%, with 2 rods at position 12

- Reactor pressure 900 psig

- Reactor water level 210"

- Suppression Pool temperature 100°F

- Suppression Pool level 14.0 feet

- Drywell pressure 2.5 psig

- Containment pressure 2.0 psig What action is required to be performed?

a. Spray Containment

b. Emergency Depressurize

c. Commence Controlled Cooldown

d. Anticipate Emergency Depressurization

REACTOR OPERATOR Page 14 QUESTION: 017 (1.00)

The plant is operating at 20% power when a scram due to a loss of feedwater occurs. All plant equipment responds normally to the scram. No SRVs open. HPCS and RCIC automatically initiate and restore reactor level. Which one of the following is the expected configuration of the Reactor Recirculation Pump breakers due to this event?

a. CB-1 Closed CB-2 Closed CB-3 Closed CB-4 Closed CB-5 Open

b. CB-1 Open CB-2 Open CB-3 Closed CB-4 Closed CB-5 Open

c. CB-1 Closed CB-2 Closed CB-3 Open CB-4 Open CB-5 Open

d. CB-1 Open CB-2 Open CB-3 Open CB-4 Open CB-5 Open QUESTION: 018 (1.00)

Plant conditions as follows after a scram:

- Reactor Power 10% to 15%

- Main Turbine Tripped

- Main Steam Bypass Valves failed closed

- SRVs cycling on Low-low set

- Only Operator action taken is Mode Switch to Shutdown What are the expected conditions of the Feedwater and Reactor Recirculation Systems 30 seconds after the SRVs began cycling?

a. Feedwater pumps maintaining level in Auto, Reactor Recirculation Pumps in Slow

b. Feedwater pumps maintaining level in Auto, Reactor Recirculation Pumps Off

c. Feedwater pumps in Manual/Minimum, Reactor Recirculation Pumps in Slow

d. Feedwater pumps in Manual/Minimum, Reactor Recirculation Pumps Off

REACTOR OPERATOR Page 15 QUESTION: 019 (1.00)

PEI-D17, Radioactive Release Control directs isolation of all primary systems that are discharging into areas outside one or more of the following: Annulus, Auxiliary Building, Intermediate Building, and Steam Tunnel, except for systems required to assure adequate core cooling or shutdown the Reactor.

Per the PEI Bases, these systems are specifically exempted from isolation because:

a. isolation of these systems requires an emergency depressurization.

b. additional radiological consequences from these systems is unlikely.

c. they are required to support alternate reactor depressurization methods.

d. isolation may ultimately result in a much larger uncontrolled radiological release.

QUESTION: 020 (1.00)

The plant is operating at 100% power with Control Room HVAC Train A in normal and Control Room HVAC Train B in standby. When the following plant conditions occur:

- CONT RM EMERG RCIRC A CHAR FLTR TEMP HIGH

- SAS reports smoke detected in duct of Control Room HVAC Train A

- M26-R032A indicates 260°F and increasing Based on these indications the operator would ____

a. confirm auto initiation of the deluge system on high temperature.

b. confirm auto initiation of the deluge system on smoke in HVAC Train A.

c. manually initiate deluge by locally opening the deluge valve.

d. manually initiate deluge by arming and depressing the deluge pushbutton.

REACTOR OPERATOR Page 16 QUESTION: 021 (1.00)

The plant is operating at 50% power with RHR A in standby. The blue indicating light above the LPCI Injection Valve, 1E12-F042A is off.

Following a small break LOCA the following plant conditions exist:

- Drywell Pressure 1.8 psig and increasing

- Containment Pressure 1.0 psig and increasing

- Reactor Pressure 800 psig and lowering Based on these conditions, which of the following describes the status of the LPCI Injection Valve, 1E12-F042A?

a. Closed; can be opened by taking the switch to open.

b. Closed; will open when reactor pressure lowers to 600 psig.

c. Open; the pressure permissive is met.

d. Open; a LOCA signal bypasses the pressure permissive.

QUESTION: 022 (1.00)

The plant is operating at 40%.

Condensing chamber reference leg failures have caused RPS Level channels A and C and Feedwater Narrow Range channels A and C to fail high.

Which one of the following describes the immediate system response to these failures?

a. Feedwater pumps operating and a RPS half scram.

b. Feedwater pumps operating and a RPS full scram.

c. Feedwater pumps tripped and a RPS half scram.

d. Feedwater pumps tripped and a RPS full scram.

REACTOR OPERATOR Page 17 QUESTION: 023 (1.00)

The plant is operating at 100% power when a trip of Containment Vessel Chiller A occurred.

- Containment temperature and pressure are slowly increasing.

- Drywell temperature and pressure are steady.

- Alarm CONTAINMENT TEMP A(B) HIGH has been received on panel P601.

- No PEI Entry Conditions exist.

Which one of the following conditions will occur if Containment temperature and pressure continue to increase with no operator action taken?

a. Drywell Vacuum Breakers will open.

b. Containment Vacuum Breakers will open.

c. Indicated Suppression Pool level will increase.

d. Indicated Containment Upper Pool level will decrease.

QUESTION: 024 (1.00)

A plant startup is in progress after completion of RFO-11. Plant conditions are as follows:

- Mode 2

- APRM Power 3%

- IRMs on Range 8 To protect the reactor from an inadvertent reactivity addition due to a control rod withdrawal accident the primary scram signal is ____.

a. SRM High-High Flux

b. IRM Neutron Flux-High

c. APRM Neutron Flux-High

d. APRM Neutron Flux-High Setdown

REACTOR OPERATOR Page 18 QUESTION: 025 (1.00)

Fuel element failure is indicated by increasing plant radiation levels.

MAIN STEAM LINE RADIATION HIGH alarm is received for all Main Steam Line Radiation Monitors.

MAIN STEAM LINE RADIATION HI HI/INOP alarm is received for Main Steam Line Radiation Monitors A and B.

Which one of the following receives a close signal?

a. Off-Gas Discharge Isolation Valve, N64-F632

b. Reactor Water Sample Isolation Valve, B33-F019

c. Main Steam Line Isolation Valves, B21-F022A-D and B21-F028A-D

d. Mechanical Vacuum Pump Suction Valves, N62-F130A and N62-F130BExamination Outline Cross QUESTION: 026 (1.00)

The following plant conditions exist:

- ATWS

- MSIVs are closed

- Pressure control is on SRVs

- Suppression Pool Level is 21.5'

- Suppression Pool Temperature is 129°F Which one of the following is the highest pressure the reactor can reach without exceeding the Heat Capacity Limit based on the given conditions?

Reference Provided - PEI-SPI Supplement Figure 4

a. 700 psig.

b. 750 psig.

c. 900 psig.

d. 950 psig.

REACTOR OPERATOR Page 19 QUESTION: 027 (1.00)

PEI-N11, Containment Leakage Control is entered on high RCIC room temperature and sump level. A non-Licensed Operator reports from outside the RCIC room that there is only a steam leak and that fire protection deluge has initiated in the RCIC room. Control Room actions are in progress to isolate the steam leak.

Which of the following is the correct action and why? (Reference Provided - Modified PEI-N11 Flowchart)

a. Do not isolate fire protection deluge; to confine the high temperature problem to the RCIC room.

b. Do not isolate fire protection deluge; the high RCIC room temperature takes precedence over other Secondary Containment concerns.

c. Isolate fire protection deluge; to prevent from exceeding a maximum safe water level in the RCIC room.

d. Isolate fire protection deluge; to prevent from emergency depressurizing due to threatening Secondary Containment.

QUESTION: 028 (1.00)

The following plant conditions exist:

- A LOCA is in progress.

- All ECCS systems are injecting into the RPV.

Fifteen minutes later, a LOOP occurs and the Division 1 Diesel Generator fails to start. ONI-R10, Loss of AC Power, is entered.

Prior to restoring the Division 1 Diesel Generator, an automatic start of LPCI Pump A is prevented due to the: ________________________.

a. loss of NPSH.

b. loss of pump seal cooling.

c. potential for water hammer.

d. potential for Diesel Generator overload.

REACTOR OPERATOR Page 20 QUESTION: 029 (1.00)

The plant is in Mode 3 with RHR Loop A in Shutdown Cooling, when a trip of RHR Pump A occurs. Efforts are being made to place RHR Pump B into Shutdown Cooling.

Reactor Pressure is currently 85 psig and ERIS indicates a constant heatup rate of 30°F/hr.

Predict the maximum amount of time available in order to place RHR Pump B into Shutdown Cooling and terminate the heatup.

Reference provided -- Steam Tables

a. 60 minutes

b. 68 minutes

c. 78 minutes

d. 100 minutes QUESTION: 030 (1.00)

Plant conditions are as follows:

- Mode 3, forced cooldown in progress

- Reactor Pressure 400 psig

- Reactor Level 185" An inadvertent initiation of Low Pressure Core Spray (LPCS) occurs.

Which of the following actions is required and predict if injection occurred?

a. Shut the LPCS Injection Valve; LPCS injection occurred.

b. Shut the LPCS Injection Valve; LPCS injection did not occur.

c. Stop the LPCS Pump; LPCS injection occurred.

d. Stop the LPCS Pump; LPCS injection did not occur.

REACTOR OPERATOR Page 21 QUESTION: 031 (1.00)

Due to a valve mis-positioning error, both HPCS Suction Valves (E22-F001 and E22-F015) are closed.

Which one of the following is the expected response of the HPCS System, upon receipt of a HPCS Auto Initiation Signal?

a. The HPCS Pump will not start since no clear suction path is available.

b. The HPCS Pump will start and the HPCS CST Suction Valve will automatically open.

c. The HPCS Pump will start but neither one of the suction valves will automatically open.

d. The HPCS Pump will start and the HPCS Suppression Pool Suction Valve will automatically open.

QUESTION: 032 (1.00)

Pull to criticality is in progress during a plant startup following a refuel outage.

The Standby Liquid Control Storage Tank heaters are removed from service in preparation for Electrical Maintenance to work on a heater ground. Current tank temperature is 80°F and slowly lowering. Chemistry reports that the boron solution concentration is 2.83 weight percent.

Which one of the following is an acceptable SLC System Storage Tank net volume and temperature?

Reference provided - Technical Specification page 3.1-23

a. 65°F, 4580 gallons

b. 67°F, 4750 gallons

c. 71°F, 4690 gallons

d. 74°F, 4490 gallons

REACTOR OPERATOR Page 22 QUESTION: 033 (1.00)

The plant is operating at 100% power with the Reactor Protection System MG SET TRANSFER switch in NORM.

The following occurs:

1. Numerous Control Room Alarms are received

2. Half scram is indicated

3. MSIV position indication is lost for the Inboard MSIVs

4. Inboard BOP isolation has occurred This is an indication of a loss of power from Bus:

a. F1B08

b. F1C08

c. F1C12

d. F1D12 QUESTION: 034 (1.00)

While decreasing reactor power, Intermediate Range Monitor (IRM) Channel A is indicating 30/125 of scale on range 6. The operator inadvertently ranges IRM A to range 5.

What is the result of the operator error?

Power Indication System Effect

a. 9.5/125 None

b. 9.5/125 Rod Block

c. 95/125 None

d. 95/125 Rod Block

REACTOR OPERATOR Page 23 QUESTION: 035 (1.00)

The following plant conditions exist:

- The reactor is critical.

- Reactor power is on Range 4 of the Intermediate Range Monitors.

- Source Range (SRM) detectors are being withdrawn from the core.

Subsequently, the high voltage power supply to SRM D detector fails low.

Which one of the following describes the response of the Source Range Monitoring System?

Assume no operator actions have been performed.

An SRM control rod block signal is . . .

a. not generated; SRM D detector withdrawal from the core stops.

b. not generated; SRM D detector withdrawal from the core continues.

c. generated; SRM D detector withdrawal from the core stops.

d. generated; SRM D detector withdrawal from the core continues.

REACTOR OPERATOR Page 24 QUESTION: 036 (1.00)

The plant is operating at 100% power with the following LPRMs bypassed for APRM H:

- 5A-08-17

- 3B-32-41

- 4C-40-33

- 5C-24-17

- 3D-48-41 LPRM 1C-24-49 fails downscale, Reactor Engineering recommends bypassing the failed LPRM.

When the LPRM is bypassed APRM H is _____.

Reference provided - SOI-C51(APRM) Attachment 1

a. operable, one additional LPRM failure will generate an INOP Trip.

b. operable, the LPRM Downscale alarm has cleared.

c. inoperable, with an INOP Trip signal in.

d. inoperable, the LPRM Downscale alarm has cleared.

REACTOR OPERATOR Page 25 QUESTION: 037 (1.00)

The Reactor Core Isolation Cooling System (RCIC) has been started in CST to CST mode per SOI-E51, Reactor Core Isolation Cooling. The following conditions exist:

- RCIC Flow Controller, 1E51-R600 is in Auto, set at 700 gpm

- RCIC Flow is 700 gpm

- RCIC discharge pressure 1075 psig

- Reactor Power 100%

What happens to RCIC speed and discharge pressure if 1E22-F022, RCIC First Test Valve To CST is throttled open slightly?

a. RCIC speed and discharge pressure both higher

b. RCIC speed and discharge pressure both lower

c. RCIC speed lower and discharge pressure higher

d. RCIC speed higher and discharge pressure lower

REACTOR OPERATOR Page 26 QUESTION: 038 (1.00)

The reactor has scrammed from 100% power due to a loss of offsite power. The following conditions exist:

- All emergency diesel generators started and are supplying their respective EH Bus.

- All low pressure ECCS pumps are in Standby

- Reactor pressure is cycling on SRV operation

- Reactor is shutdown

- RCIC has isolated

- HPCS Pump has tripped

- Reactor level is 186.5", decreasing at 10"/min

- Drywell pressure is 1.50 psig, increasing at 0.25psig/min Which one of the following describes the response of the Automatic Depressurization System (ADS), if plant conditions remain as stated, no operator action is taken and all equipment responds as expected?

a. ADS will automatically initiate in 2 minutes and 36 seconds.

b. ADS will automatically initiate in 7 minutes and 24 seconds.

c. ADS will automatically initiate in 17 minutes.

d. ADS will automatically initiate in 18 minutes and 45 seconds.

REACTOR OPERATOR Page 27 QUESTION: 039 (1.00)

The plant is in Mode 4 with RPV temperature being maintained 80°F to 110°F by RHR B in Shutdown Cooling. The ATC Operator has signed in two I&C SVIs:

- RPV Level 3 on Narrow Range Level Channel A

- RPV Level 3 on Narrow Range Level Channel D.

During performance of these SVIs, a RPV Level 3 trip signal is input concurrently into their respective channels.

What is(are) the consequence(s) of this action?

a. 1E12-F008 and 1E12-F053B close

b. 1E12-F009 and 1E12-F053B close

c. only 1E12-F008 closes

d. No Isolation QUESTION: 040 (1.00)

With the plant operating at 100% power, a Main Steam Isolation Valve (MSIV) isolation signal is received due to Main Steam Line (MSL) A Flow High. A check of panel 1H13-P691 indicates that a trip is indicated on all four MSL A Flow Channels.

The Reactor Operator checks 1H13-P601 to confirm proper system response.

Which MSIVs and which MSL Drain Valves, if any, will the operator find closed?

a. No MSIVs and No MSL Drain Valves

b. Inboard MSIVs and Inboard MSL Drain Valve

c. Outboard MSIVs and Outboard MSL Drain Valve

d. All MSIVs and All MSL Drain Valves

REACTOR OPERATOR Page 28 QUESTION: 041 (1.00)

With the plant operating at 40% power a high drywell pressure due to an air leak caused a reactor scram. Instrument Air (P52) was isolated to Containment.

How many of the SRVs have a continuous supply of air available for long-term pressure control?

a. None

b. 8

c. 9

d. 19 QUESTION: 042 (1.00)

The plant was operating at 100% power when ADS SRV B21-F041B inadvertently opened. The following conditions exist at this time:

- Reactor Power 100%

- Suppression Pool Temperature 111°F

- Suppression Pool Level 18.7'

- B21-F041B solenoid light energized Containment has been evacuated but no other operator actions have been performed.

Which of the following actions must the operators perform next?

a. Scram the reactor and place the mode switch in Shutdown.

b. Reduce reactor power to #90% with Recirc flow.

c. Place both keylock switches for the SRV in OFF.

d. Pull the A solenoid fuses for the SRV.

REACTOR OPERATOR Page 29 QUESTION: 043 (1.00)

With the plant operating at 75% power the following plant feedwater conditions exist:

- Three element in control in Auto

- C34-N004A transmitter is bypassed for testing

- C34-N004B indicates 199"

- C34-N004C indicates 196" Level instrument C34-N004C then fails downscale.

Level instrument (1) was initially controlling RPV Level. After the instrument failure, feedwater control is in (2) .

(1) (2)

a. C34-N004B manual

b. C34-N004B single element in Auto

c. C34-N004C manual

d. C34-N004C single element in Auto QUESTION: 044 (1.00)

Which of the following is the process filter flow path for Annulus Exhaust Gas Treatment System?

a. HEPA Charcoal Roughing Filter

b. Roughing Filter Charcoal HEPA

c. Roughing Filter HEPA Charcoal HEPA

d. Roughing Filter HEPA Charcoal Roughing Filter

REACTOR OPERATOR Page 30 QUESTION: 045 (1.00)

The plant is at 100% power with the following plant conditions:

- EH11 and EH13 supplied from Normal Preferred Source

- EH12 supplied from Alternate Preferred Source

- Control Rod Drive Pump B in service, A in standby

- Service Water Pumps A and B in service, C in standby and D OOS.

The following alarm is received; BUS EH12 VOLTAGE DEGRADATION. Bus EH12 volts indicate 3700 VAC.

The EH12 Bus undervoltage actions will occur in (1) . In response to these actions the operator must (2) .

(1) (2)

a. 12 seconds perform CRD Pump Trip Recovery

b. 12 seconds confirm the Auto start of Service Water Pump B

c. 4 minutes perform CRD Pump Trip Recovery

d. 4 minutes confirm the Auto start of Service Water Pump B

REACTOR OPERATOR Page 31 QUESTION: 046 (1.00)

The Main Generator is in the process of being paralleled to the grid per IOI-0003, Power Changes. The SYNC SELECT SWITCH is in the S610-PY-TIE position.

The following indications are observed on panel H13-P680:

- MAIN TRANSFORMER (incoming) S11-R013 346 KV

- PY-EL-LINE (running) N41-R120 344 KV

- Synchroscope is rotating slow in the counter-clockwise direction.

Before the S610-PY-TIE breaker can be closed, the operator must (1) the Auto Voltage Regulator to match voltage and must (2) the Load Set until the Synchroscope is moving slowly in the clockwise direction.

(1) (2)

a. lower decrease

b. lower increase

c. raise decrease

d. raise increase QUESTION: 047 (1.00)

A plant worker inadvertently opens the DIV 1 ATWS UPS supply breaker on bus ED1A06.

What is the impact of this event on the Division 1 ATWS UPS?

The Static Transfer Switch (1) transferred to the alternate (2) source.

(1) (2)

a. automatically AC

b. automatically DC

c. must be manually AC

d. must be manually DC

REACTOR OPERATOR Page 32 QUESTION: 048 (1.00)

A Station Blackout is in progress. ONI-SPI D1, Maintaining System Availability directs that the Telephone Battery Room door be opened within two hours.

Which of the following describes the location and the specific reason given for performing this action?

a. Control Complex 638': Dissipate Heat

b. Control Complex 638': Prevent Hydrogen build up

c. Service Building 640': Dissipate Heat

d. Service Building 640': Prevent Hydrogen build up QUESTION: 049 (1.00)

The Division 1 Diesel Generator is operating in parallel with the grid for surveillance testing. A Loss of Offsite Power occurs. Division 2 and 3 Diesel Generators energize EH12 and EH13.

The following plant conditions exist:

- Reactor Scram All Rods In

- Reactor Level is lowering rapidly

- HPCS and RCIC failed to start on lowering Reactor Level

- Reactor Pressure being controlled on SRVs Subsequently, the following alarm is received, DG TRIP CRANKCASE PRESS HIGH for Division 1 DG. A plant operator reports that crankcase pressure is high.

Which of the following is correct regarding Division 1 DG for the above condition and what action, if any, is required by the operator?

a. Crankcase fans are operating and the operator shall shutdown the DG.

b. Crankcase fans are not operating and the operator shall shutdown the DG.

c. Crankcase fans are operating and the operator shall not shutdown the DG.

d. Crankcase fans are not operating and the operator shall not shutdown the DG.

REACTOR OPERATOR Page 33 QUESTION: 050 (1.00)

Following the paralleling of the Division 1 Diesel Generator with its respective bus, Diesel Generator parameters are as follows:

1. 4200 Volts

2. 100 KVAR

3. 200 KW If the Operator places the generator voltage regulator to the RAISE position and the indicated KVARs decrease, the diesel generator's present power factor is (1) , and in order to establish and/or maintain the proper power factor, the Operator must (2) .

(1) (2)

a. lagging; continue to increase the generator's voltage regulator output

b. lagging; maintain the generator's present voltage regulator output

c. leading; continue to increase the generator's voltage regulator output

d. leading; maintain the generator's present voltage regulator output after the engine comes to a complete stop.

QUESTION: 051 (1.00)

A complete loss of instrument air occurs. Which of the following describes the expected valve response for the listed air operated valves?

(1) Motor Feed Pump Flow Control Valves (2) Hotwell Make-up and Dump Valves (1) (2)

a. Fail As Is Fail Closed

b. Fail As Is Fail As Is

c. Fail Open Fail Closed

d. Fail Open Fail As Is

REACTOR OPERATOR Page 34 QUESTION: 052 (1.00)

SOI-P43, Nuclear Closed Cooling System requires NCC HX OUT TEMP to be maintained between 70°F and 89°F.

The 70°F temperature is based on (1) and the 89°F temperature is based on (2) .

(1) (2)

a. MSIV Closure Event Reactor Recirculation Pumps

b. P47 Chillers MSIV Closure Event

c. Reactor Recirculation Pumps P47 Chillers

d. Reactor Recirculation Pumps MSIV Closure Event QUESTION: 053 (1.00)

Which of the following Power / Flow combinations will enable the OPRM scram function?

Power Core Flow

a. 20% 30%

b. 50% 65%

c. 60% 55%

d. 75% 90%

REACTOR OPERATOR Page 35 QUESTION: 054 (1.00)

A post scram reactor startup is in progress with the following plant conditions:

- Reactor Pressure 350 psig

- Reactor Level 200"

- Reactor Power Range 6 on IRMs Control Rod 10-47 did not move when given a withdraw signal from it's current notch position of

12. Drive water differential pressure had been adjusted to 450 psid.

The operator's next action should be to ____.

a. individually scram rod 10-47, then disarm it electrically and hydraulically.

b. insert rod 10-47 to position 00, then disarm it electrically and hydraulically.

c. raise drive water differential to 500 psid and attempt double clutching to withdraw rod 10-47.

d. raise drive water differential to 500 psid and apply a withdrawal signal to withdraw rod 10-47.

QUESTION: 055 (1.00)

The plant is operating at 100% power when the supply breaker Bus L11 trips and Bus L11 is de-energized.

Which of the following would be directly affected as a result of the loss of Bus L11?

a. Reactor Recirculation Pump A

b. Reactor Recirculation Pump B

c. Circulating Water Pump C

d. Motor Feed Pump

REACTOR OPERATOR Page 36 QUESTION: 056 (1.00)

Given the following initial conditions on Reactor Recirculation Flow Control Valve Hydraulic Power Unit (HPU) A:

- Subloop 1 LEAD, READY, OPERATIONAL AND PRESSURIZED lights on

- Subloop 2 READY light on A plant operator reports rising oil temperature on HPU A. A Control Room operator checks panel 1H13-P614 and the OIL WARM light illuminates.

With this condition what is the status of HPU A Subloops and Reactor Recirculation Flow Control Valve A?

a. Subloop 1 and 2 in MAINTENANCE and Flow Control Valve motion is inhibited.

b. Subloop 1 and 2 in MAINTENANCE and Flow Control Valve motion is not inhibited.

c. Subloop 1 in MAINTENANCE and Subloop 2 in LEAD and Flow Control Valve motion is inhibited.

d. Subloop 1 in MAINTENANCE and Subloop 2 in LEAD and Flow Control Valve motion is not inhibited.

REACTOR OPERATOR Page 37 QUESTION: 057 (1.00)

The plant has experienced a LOCA and the following plant conditions exist:

- Reactor Level minus 50"

- Hydrogen Igniters energized

- Containment Hydrogen Concentration 6.5%

- Drywell Hydrogen Concentration 10%

Which of the following is the primary hydrogen production mechanism and what action is required at the above hydrogen concentrations?

a. Zirc-Water Reaction Stop Hydrogen Igniters

b. Zirc-Water Reaction Stop Hydrogen Recombiners

c. Steel Oxidation Reaction Stop Hydrogen Igniters

d. Steel Oxidation Reaction Stop Hydrogen Recombiners QUESTION: 058 (1.00)

Given the following plant conditions:

- A LOCA has occurred

- Reactor Level 1 at 10:00

- Drywell Pressure was 1.7 psig at 10:03

- Containment Pressure was 8.0 psig at 10:05 Based on the above conditions, when did/(will) Containment Spray Mode automatically initiate?

a. Containment Spray initiated at 10:03.

b. Containment Spray initiated at 10:05.

c. Containment Spray will initiate at 10:10.

d. Containment Spray will initiate at 10:15.

REACTOR OPERATOR Page 38 QUESTION: 059 (1.00)

The Main Generator is at 150 Mwe and plant power is being held at this level until SVI-B21-T2005, SRV Exercise test is completed.

When SRV 1B21-F051A is tested, Main Steam Line A Flow indicator on 1H13-P680 will (1) and Generator Load will (2) . (NOTE: SRV 1B21-F051A is located on A Main Steam Line.)

(1) (2)

a. decrease decrease

b. decrease remain as is

c. increase decrease

d. increase remain as is

REACTOR OPERATOR Page 39 QUESTION: 060 (1.00)

Given the following initial plant conditions:

- Reactor Power 100%

- Reactor Pressure 1025 psig

- N32/C85 Throttle Pressure 970 psig

- Pressure Setpoint 940 psig

- Max. Combined Flow Set 130%

- Load Limit Set 104%

- Load Set 108%

- B regulator in Test

- Bypass Jack in Control What is the response of the Steam Bypass and Pressure Regulating System with a slight increase in Reactor Pressure and an increase in N32/C85 Throttle Pressure to 972 psig with no Operator action? Reference provided - EHC Control System Block Diagram Control Valves will receive a (1) open signal and Bypass Valves a (2) open signal.

(1) (2)

a. 104% -1%

b. 104% 2%

c. 107% -1%

d. 107% 2%

REACTOR OPERATOR Page 40 QUESTION: 061 (1.00)

The plant is operating at 100%, when the following occurs:

- H2 SEAL/STATOR CLG TRBL alarm is received

- Neither Stator Water Cooling Pump is operating Predict the initial plant response to this condition with no operator action?

a. Reactor Power will lower and Turbine Bypass Valves will open.

b. Reactor Power will lower and Turbine Bypass Valves will remain closed.

c. Reactor Power will remain at 100% and Turbine Bypass Valves will open.

d. Reactor Power will remain at 100% and Turbine Bypass Valves will remain closed.

QUESTION: 062 (1.00)

The plant is operating at 100%, when the following occurs:

- HEATER 2C LEVEL HIGH is received

- Heater level 2C continues to rise Predict the plant response to this condition with no operator action? (1) flow will isolate to Heater 2C and the normal drain(s) from Heater (2) will close.

(1) (2)

a. Condensate 3B

b. Condensate 3A and 3B

c. Steam 3B

d. Steam 3A and 3B

REACTOR OPERATOR Page 41 QUESTION: 063 (1.00)

Following a reactor scram the following conditions exist:

- Operating in PEI-B13 non-ATWS on Level 3

- RFPTs A and B are operating at their low speed stop

- RPV Pressure is at 930 psig and lowering RFPTs A and B speed is approximately (1) RPM and they will commence feeding to the reactor at a reactor pressure of approximately (2) psig if no operator action is taken.

(1) (2)

a. 1100 800

b. 1100 900

c. 3300 800

d. 3300 900 QUESTION: 064 (1.00)

A small reactor water leak has occurred in the Reactor Water Cleanup Pump Valve Room on Auxiliary Building 599 elevation. The leak has resulted in the following Auxiliary Building Airborne Radiation Monitor (1D17-K700) alarms:

- Particulate Channel (1D17-K708) Alert

- Iodine Channel (1D17-K707) Alert

- Gas Channel (1D17-K706) High Auxiliary Building Ventilation Supply Fans have (1) and PEI-N11, Containment Leakage Control entry is (2) .

(1) (2)

a. tripped required

b. tripped not required

c. not tripped required

d. not tripped not required

REACTOR OPERATOR Page 42 QUESTION: 065 (1.00)

Which one of the following signals will generate a Diesel Fire Pump Trip?

a. Overspeed

b. High Water Temperature

c. Low Lube Oil Pressure

d. Low Oil Reservoir Level QUESTION: 066 (1.00)

Assume that you receive your license on September 1, 2007, but because of vacation and required training you do not start standing watches (RO or SRO as applicable) until Friday September 28, 2007 and are scheduled to stand watch through Wednesday October 3, 2007.

Your shifts are scheduled for twelve hours each day. Select the statement below that describes your license status on October 1, 2007.

a. Your license is considered active and you can assume the watch on October 1, 2007. If you stand watches through Wednesday, you will not need to stand any more watches until the January-March quarter to maintain proficiency.

b. Your license is considered active and you can assume the watch on October 1, 2007. If you stand watches through Wednesday, you will need to stand at least two additional watches before January 1, 2008 to maintain proficiency.

c. Your license will be considered inactive and you cannot assume the watch on October 1, 2007. You must complete a minimum of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift functions, under the direction of a licensed RO or SRO as applicable, in the position to which you are assigned in order to regain active status.

d. Your license will be considered inactive and you cannot assume the watch on October 1, 2007. You may regain active status by completing your Monday through Wednesday shifts, under the direction of a licensed RO or SRO as applicable, in the position to which you are assigned.

REACTOR OPERATOR Page 43 QUESTION: 067 (1.00)

In which of the following areas may the Reactor Operator At The Controls initiate corrective actions or verify receipt of an annunciator, in the event of an emergency affecting the safe operation of the plant.

Reference provide - Modified NOP-OP-1002 attachment 3 Perry Control Room

a. Only area 1

b. Only areas 1 and 2

c. Only areas 1, 2 and 3

d. areas 1, 2, 3, and 4 QUESTION: 068 (1.00)

During a plant cooldown from 1% power at normal operating pressure to 120°F and 0 psig, the Narrow Range Level instrument is selected for digital display on 1H13-P680.

ICS is not available so the Reactor Operator maintains 196" indicated on the digital display during the entire cooldown.

What will actual RPV level be when the final plant conditions are reached?

Reference provided - PDB-C0005, RPV Level Comparison Graphs

a. 185"

b. 190"

c. 196"

d. 205"

REACTOR OPERATOR Page 44 QUESTION: 069 (1.00)

The is an individual assigned responsibility for issuing Clearances and keeping Control Room personnel informed of all plant configuration changes prior to establishing or removing a Clearance.

a. Clearance Authority

b. Clearance Holder

c. Operating Representative

d. Work Document Holder QUESTION: 070 (1.00)

In addition to the Refueling Supervisor and the Platform Operator, which of the following personnel is required to be on the refueling bridge during refueling?

a. Health Physics Technician

b. Reactor Engineer

c. Refuel Floor Supervisor

d. Spotter QUESTION: 071 (1.00)

Which of the following is the lowest radiation exposure that would allow the Shift Manager to waive the IV/CV of a component?

a. 9 mrem

b. 11 mrem

c. 16 mrem

d. 21 mrem

REACTOR OPERATOR Page 45 QUESTION: 072 (1.00)

Which one of the following conditions requires the Control Room Operator to verify that a liquid radwaste discharge has automatically terminated?

a. Discharge Tunnel Service Water low flow.

b. Emergency Service Water Pump B low flow.

c. HPCS ESW Pump Discharge low pressure.

d. Service Water Pump Discharge Header low pressure.

QUESTION: 073 (1.00)

The plant scrammed from 100% power following an earthquake. The following plant conditions exist:

- Control Rod 30-31 at position 2

- Control Rod 18-31 at position 4

- Drywell Pressure 1.5 psig

- MSIVs closed on high Steam Tunnel Temperature

- Suppression Pool Temperature 94°F

- No valid RPV Level indication The plant should be operating in which of the following Plant Emergency Instructions (PEI)?

a. only PEI-B13 RPV Control (non-ATWS), PEI-B13 RPV Flooding, and PEI-T23 Containment Control.

b. only PEI-B13 RPV Control (ATWS), PEI-B13 RPV Flooding, and PEI-N11 Containment Leakage Control

c. PEI-B13 RPV Control (non-ATWS), PEI-B13 RPV Flooding, PEI-T23 Containment Control and PEI-M51/56 Hydrogen Control

d. PEI-B13 RPV Control (ATWS), PEI-B13 RPV Flooding, PEI-N11 Containment Leakage Control, and PEI-M51/56 Hydrogen Control

REACTOR OPERATOR Page 46 QUESTION: 074 (1.00)

Given the following conditions:

- PEI-B13, RPV Control (Non-ATWS), was entered due to low RPV water level

- 10 minutes later, while still in PEI-B13, Drywell Pressure rises to 1.7 psig Which one of the following describes the required shift crew actions?

a. Continue on in PEI-B13 and enter all legs of PEI-T23.

b. Re-enter PEI-B13 and enter all legs of PEI-T23.

c. Continue on in PEI-B13 and enter only the Drywell and Containment Pressure Control leg of PEI-T23.

d. Re-enter PEI-B13 and enter only the Drywell and Containment Pressure Control leg of PEI-T23.

QUESTION: 075 (1.00)

A plant startup is in progress, Reactor Recirculation Pump shift to fast preparations have started. The following occurs:

- ANN PWR SUPPLY FAIL is illuminated

- Alarms that were locked in have deactivated The Control Room actions would be to dispatch an operator to (1) and (2) plant startup.

(1) (2)

a. D-1-A continue

b. D-1-A suspend

c. D-1-B continue

d. D-1-B suspend

SENIOR REACTOR OPERATOR Page 47 QUESTION: 076 (1.00)

While operating at 100% with all rods withdrawn. The following sequence of events occurs as noted in the Plant Narrative Log:

11:15 C11 - CRDH Pump A trips. Operator sent to investigate.

11:20 C11 - RO Attempted to start CRDH Pump B. Pump failed to start.

11:25 C11 - Accumulator fault on rod 10-31. Operator sent to investigate.

11:27 C11 - Accumulator fault on rod 30-31. Operator sent to investigate.

11:35 C11 - Operator investigating C11 Accumulator faults reports back that rod 10-31 Accumulator is 1500 psig and rod 30-31 Accumulator is 1480 psig.

Based on these log entries, when must the Unit Supervisor direct the Reactor Mode Switch be placed in Shutdown?

a. 11:35

b. 11:45

c. 11:47

d. 11:55

SENIOR REACTOR OPERATOR Page 48 QUESTION: 077 (1.00)

A plant startup to full power is being performed. The Reactor Engineer reports that due to a failure of the feedwater flow inputs to the Process Computer, the calculations on the Periodic Log were incorrect. He has entered the proper values of substitute data and printed the valid Periodic Log.

Reference provided -- Modified Valid Periodic Log Based on the information contained on the valid Periodic Log, which one of the following is required?

a. Restore MCPR to within the limit and shutdown the reactor.

b. Restore MCPR to within the limit or reduce power to < 23.8%.

c. Restore MFLCPR to within the limit or reduce power to < 23.8%.

d. Restore loadline to less than the MEOD Boundary and continue plant operation.

QUESTION: 078 (1.00)

Which one of the following examples of configuration changes is required to be controlled per PAP-1402, Temporary Modification Control?

a. Additional fire suppression equipment is connected and staged per PAP-1910, Fire Protection Program for compensatory actions for 14 days.

b. Test equipment is installed to determine RPS actuation during the MSIV closure scram functional surveillance test.

c. Test equipment is installed on the operating Control Complex Chiller to bypass the low NCC flow trip for 7 days.

d. RCIC is isolated for maintenance; drain and vent pipe caps are removed for system draining.

SENIOR REACTOR OPERATOR Page 49 QUESTION: 079 (1.00)

During RFO-11, work in RWCU Heat Exchanger Room was in progress. During this work a failure of telemetry dosimetry occurred on a worker.

Radiation Protection determined that the worker received the following doses:

- 4 Rem TEDE to the whole body

- 5 Rem to the eyes

- 100 Rem shallow dose to his right knee What NRC communication(s) is (are) required for this event per PAP-1604, Reports Management?

Reference Provided PAP-1604 Reports Management

a. Only an Immediate Notification

b. Only a 24 Hour Notification

c. Immediate Notification and a 30 day written report

d. 24 Hour Notification and a 30 day written report QUESTION: 080 (1.00)

Venting of the Containment using PEI-SPI 7.3, FPCC Containment Venting has been initiated due to exceeding Primary Containment Limit (PCL). Which one of the following correctly describes the condition that must be met before venting of the Containment can be terminated?

Venting is continued only until containment pressure has been reduced to minimize the amount of radioactivity released while assuring containment integrity.

a. below the Primary Containment Limit (PCL)

b. below the Pressure Suppression Pressure (PSP) limit

c. below 2.25 psig

d. to atmospheric pressure

SENIOR REACTOR OPERATOR Page 50 QUESTION: 081 (1.00)

A reactor scram and station blackout has occurred. Reactor pressure is being maintained by manually operating SRVs. Reactor level is slowly lowering. All control rods are fully inserted.

RCIC and HPCS have failed and the operators are in the process of lining up Fast Fire Water at this time. No other injection systems are available.

Considering only Reactor level, which of the following statements describes the requirement for Emergency Depressurization if Reactor level continues to lower, based on the current status of injection systems?

Reference provided Modified PEI-B13 RPV Control (Non-ATWS)

a. Emergency Depressurization may be performed anytime while Reactor level is between 0" and -42.5".

b. Emergency Depressurization must not be performed until Reactor level reaches

-42.5".

c. Emergency Depressurization may be performed anytime while Reactor level is between 0" and -25".

d. Emergency Depressurization must not be performed until Reactor level reaches

-25".

SENIOR REACTOR OPERATOR Page 51 QUESTION: 082 (1.00)

An off-site release event is in progress.

The following information is available for the Shift Manager:

- HIGH radiation alarm has been received on the TB/HB Ventilation Gas, 1D17-K856.

- HIGH radiation alarm has been received on the TB/HB Ventilation Iodine, 1D17-K857.

- HIGH radiation alarm has been received on the TB/HB Ventilation Particulate, 1D17-K858.

- TB/HB Ventilation GAS module indicates 7.2 x 104 cpm.

- TB/HB Ventilation GAS module HIGH alarm setpoint is 3.4 x 103 cpm.

- Chemistry reports that it will take 30 minutes to obtain a TB/HB Ventilation gas sample for analysis.

- Chemistry reports that it will take 20 minutes to perform Emergency Dose Calculations needed to determine the actual radiation levels at the site boundary.

As the Emergency Coordinator, what is the required Emergency Plan classification for this event?

Reference provided Modified EPI-A1, Emergency Action Levels

a. Unusual Event

b. Alert

c. Site Area Emergency

d. General Emergency

SENIOR REACTOR OPERATOR Page 52 QUESTION: 083 (1.00)

A plant scram from 100% results in the following conditions:

- Reactor Power 35%

- Mode Switch in Shutdown The Main Turbine trips, pressure control is now on (1) and the correct pressure control band is (2) .

Reference provided -- PEI-B13 RPV Control (ATWS)

(1) (2)

a. Only the Bypass Valves 800-1000 psig

b. Only the SRVs 700-900 psig

c. Bypass Valves and SRVs 800-1000 psig

d. Bypass Valves and SRVs 700-900 psig QUESTION: 084 (1.00)

The plant is operating at 100% power with HPCS Out of Service for breaker maintenance, day 5 of the 14 day LCO. The ADS A AIR STRG TANK PRESS HI/LO alarm is received.

The Reactor Operator observes that the Safety Related Air Receiver pressures are reading 100 psig and lowering on the A receiver and 165 psig and steady on the B receiver.

Based on these conditions, which one of the following Technical Specification actions is controlling plant operation?

Reference provided -- Technical Specification 3.5.1

a. Be in Mode 2 in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, Mode 3 in 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, and Mode 4 in 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

b. Be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and Reduce reactor steam dome pressure to < 150 psig in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

c. Restore air pressure or HPCS to OPERABLE in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

d. Restore air pressure in 14 days.

SENIOR REACTOR OPERATOR Page 53 QUESTION: 085 (1.00)

The plant scrams from 100% power. The following alarms and indications are called to your attention:

- Drywell Pressure 1.7 psig and rising

- Reactor Level at 50" and slowly lowering

- Containment Pressure 2.5 psig and rising

- DW UNIDENTIFIED RATE OF CHANGE HIGH, recorder on high peg These alarms and indications establish that _____.

a. no loss of a Fission Product Barrier currently exists

b. a loss of the Fuel Clad Barrier exists

c. a loss of the Reactor Coolant System Barrier exists

d. a loss of the Containment Barrier exists QUESTION: 086 (1.00)

The Heat Capacity Limit is being challenged by high reactor pressure and high suppression pool temperature.

In order to direct Emergency Depressurization using SRVs, the Unit Supervisor must confirm Suppression Pool Level at a minimum of .

a. 5.25 feet

b. 5.75 feet

c. 7.25 feet

d. 14.25 feet

SENIOR REACTOR OPERATOR Page 54 QUESTION: 087 (1.00)

During an ATWS, which one of the following identifies the highest Suppression Pool temperature, and its corresponding bases, that requires the initiation of the Standby Liquid Control System (SLC)?

a. 110°F; to ensure that Hot Shutdown Boron Weight is injected prior to Suppression Pool temperature exceeding the Heat Capacity Limit (HCL).

b. 110°F; to ensure that SLC is initiated prior to exceeding the Technical Specification limit that requires a reactor scram.

c. 120°F; to ensure that Hot Shutdown Boron Weight is injected prior to Suppression Pool temperature exceeding the Heat Capacity Limit (HCL).

d. 120°F; to ensure that SLC is initiated prior to exceeding the Technical Specification limit that requires RPV depressurization to less than 200 psig.

QUESTION: 088 (1.00)

Following a LOCA, the following parameters are noted:

- RPV Pressure 40 psig

- Containment Temperature 150°F

- Drywell Temperature 290°F

- RPV Levels

- Narrow Range 180"

- Wide Range 185"

- Upset Range 200"

- Shutdown Range 205" Which of these level instruments can be used to determine level?

Reference provided - PEI-SPI Supplement Figures 1 and 1a, Figures 2a, 2b, and 2c

a. All of the level ranges can be used.

b. Only Narrow and Wide Range can be used.

c. Only Shutdown and Upset Range can be used.

d. None of the level ranges can be used.

SENIOR REACTOR OPERATOR Page 55 QUESTION: 089 (1.00)

With the plant operating at 100% power, the following occurs:

- OG PRE-TREAT PRCS RAD MON RAD HIGH alarm is received.

- OFF-GAS PRETREAT Radiation Monitor 1D17-K612 is above the high alarm setpoint and rising.

- BYPASS VLV SHUT OG POST-TREAT PRCS RAD A/B HI alarm is received.

- OFF-GAS POST TREATMENT Radiation Monitors 1D17-K601A and B are above the high alarm setpoint and rising.

- OFF GAS POST TREATMENT PROCESS RAD REC 1D17-R601 indicates increasing radiation levels.

The Unit Supervisor should direct monitoring of (1) . If this condition continues to degrade Off-Gas will isolate at HIGH-HIGH from (2) .

a. (1) condenser vacuum, as this condition could be a result of high off-gas flow.

(2) Radiation Monitors 1D17-K601A and B.

b. (1) main steam line radiation levels, as this condition could be a result of a fuel defect.

(2) Radiation Monitors 1D17-K601A and B.

c. (1) condenser vacuum, as this condition could be a result of high off-gas flow.

(2) Recorder 1D17-R601

d. (1) main steam line radiation levels, as this condition could be a result of a fuel defect.

(2) Recorder 1D17-R601

SENIOR REACTOR OPERATOR Page 56 QUESTION: 090 (1.00)

An ATWS has occurred. The Unit Supervisor has been maintaining a level band of 50"-100" with the Motor Feed Pump (MFP), when Emergency Depressurization is performed due to a containment problem. The following conditions exist:

- All equipment operable

- PEI-SPI 5.1, 5.2 and 5.3 complete

- PEI-SPI 6.1 and 6.2 prepared

- 6 SRVs open

- Reactor Power 2%

- Level Band 50" -- 100" As RPV pressure reaches 600 psig and decreasing the Reactor Operator informs the Unit Supervisor that level is out of band low at 25" and lowering.

Which one of the following actions may the Unit Supervisor direct to restore RPV Level to the required band?

Reference provided -- PEI-B13 RPV Control (ATWS)

a. Immediately commence feeding with the MFP to restore level in band.

b. Commence feeding with the MFP only after RPV pressure decreases to below 140 psig.

c. Commence feeding with either RHR A or RHR B, outside the shroud, only after RPV pressure decreases to below 190 psig.

d. Commence feeding with either RHR A or RHR B, outside the shroud, as soon as RPV pressure decreases to below RHR pump shutoff head.

SENIOR REACTOR OPERATOR Page 57 QUESTION: 091 (1.00)

Power ascension was in progress when an RPV Level 8 Scram occurred while shifting feed pumps. Immediately following the scram, plant conditions are as follows:

- 10 Control Rods are at a position other than 00

- APRMs are downscale

- Reactor Level is being restored to 196" from Level 8

- Motor Feed Pump running

- Pressure Control on bypass valves at 940 psig

- All HCUs have a lit green LED when the Scram Valves pushbutton is depressed Which one of the following should the Unit Supervisor direct for inserting control rods?

a. Individually scram the rods using SRI Test switches.

b. Bypass the LPSP and manually insert the control rods.

c. Remove the fuses that de-energize the scram pilot valve solenoids.

d. Bypass rod positions as required and manually insert the control rods.

SENIOR REACTOR OPERATOR Page 58 QUESTION: 092 (1.00)

The plant is in Mode 4 after shutdown for RFO-11, the following plant conditions exist:

- RHR Pump A is operating in Shutdown Cooling

- Reactor Level is 230" and stable

- Reactor Coolant Temperature is 100°F and stable

- Reactor Recirculation Pump B is operating Subsequently, an inadvertent loss of RPS Bus A occurs and a trip of RHR Pump A. It is estimated that RPS Bus A can be recovered in two hours.

What is the affect on Shutdown Cooling and what action must the Unit Supervisor direct in order to comply with Technical Specifications?

a. Only Division 1 Shutdown Cooling Valve isolation. Verify two alternate methods of decay heat removal are available within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b. Only Division 1 Shutdown Cooling Valve isolation. Monitor reactor coolant temperature and pressure once per hour.

c. Both Division 1 and 2 Shutdown Cooling Valve isolation. Verify two alternate methods of decay heat removal are available within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

d. Both Division 1 and 2 Shutdown Cooling Valve isolation. Monitor reactor coolant temperature and pressure once per hour.

SENIOR REACTOR OPERATOR Page 59 QUESTION: 093 (1.00)

The plant is operating at 100% power. The Low Pressure Core Spray Pump and Valve Operability Test (SVI-E12-T2001) was recently completed.

The LPCS Pump Min Flow Valve, E21-F011 failed to stroke open when securing the LPCS Pump. Maintenance has reported the problem is with the motor operator. The LPCS Pump Min Flow Valve can be operated manually.

The repair estimate for the LPCS Pump Min Flow Valve is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In order to comply with Technical Specifications, the LPCS Pump Min Flow Valve is required to be (1) and the Technical Specification(s) that the Unit Supervisor must enter is/are 2) .

a. (1) Shut (2) Only 3.5.1, ECCS Operating

b. (1) Shut (2) 3.5.1, ECCS Operating and 3.6.1.3, Primary Containment Isolation Valves

c. (1) Open (2) Only 3.5.1, ECCS Operating

d. (1) Open (2) 3.5.1, ECCS Operating and 3.6.1.3, Primary Containment Isolation Valves

SENIOR REACTOR OPERATOR Page 60 QUESTION: 094 (1.00)

The plant is in Mode 5, with fuel movement complete. Core verification is in progress.

The minimum number of SRMs required to be Operable is (1) , and with less than the minimum (2) would not be permitted.

(1) (2)

a. 2 anticipatory rod stroking

b. 2 LPRM detector replacement

c. 3 anticipatory rod stroking

d. 3 LPRM detector replacement QUESTION: 095 (1.00)

The plant is operating at 100% power when the output of the Flow Channel Summer in APRM Channel B fails to zero.

A (1) will be generated and a(an) (2) Limiting Condition for Operation must be written.

Reference provided - Modified SDM Figure C51 (APRM-OPRM)-11 (1) (2)

a. Rod Block only active

b. Rod Block only potential

c. Rod Block and a half-Scram active

d. Rod Block and a half-Scram potential

SENIOR REACTOR OPERATOR Page 61 QUESTION: 096 (1.00)

The plant is operating at 100% power with only APRM H INOPERABLE. I&C commences the channel functional test on APRM B. The Unit Supervisor has delayed entering Conditions and Required actions in accordance with the following note per Technical Specification 3.3.1.1, RPS Instrumentation:

NOTE 2: When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.

The following information is documented in the SVI:

- The Unit Supervisor's authorization to start prerequisites was obtained at 0900 on May 1.

- The Reactor Operator's authorization to start the test was obtained at 1000 on May 1.

- The Unit Supervisor's signature for Inoperability was obtained at 1100 on May 1.

- Due to delays in the SVI performance, I&C does not finish the surveillance within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Which of the following is the correct time of entry into Technical Specification 3.3.1.1, RPS Instrumentation Condition A?

a. 1000 on May 1

b. 1100 on May 1

c. 1600 on May 1

d. 1700 on May 1

SENIOR REACTOR OPERATOR Page 62 QUESTION: 097 (1.00)

The plant is operating at full power. HPCS was declared INOPERABLE on January 10 at 1400, for repairs on the pump breaker.

On January 13 at 1200, a Non-Licensed Operator reports that the oil level in the RCIC oil level sight glass is out of sight high.

When must the plant be placed in Hot Shutdown if neither of these issues can be corrected?

Reference provided -- Technical Specification 3.5.1 and 3.5.3

a. 0000 on January 14

b. 0100 on January 14

c. 0100 on January 15

d. 0200 on January 25

SENIOR REACTOR OPERATOR Page 63 QUESTION: 098 (1.00)

The Unit Supervisor is performing a review of RCS cooldown data from SVI-B21-T1176, RCS Heatup and Cooldown Surveillance.

Time Temperature (°F)

Start 0800 520 0830 490 0900 450 0930 385 1000 345 1030 300 1100 260 1130 210 1200 165 1230 120 1300 100 Which one of the following is the correct analysis of the cooldown and required Technical Specifications action(s)?

The cooldown rate was exceeded (1) .

At the time of LCO entry it was required to restore parameter(s) to within limits (2) .

(1) (2)

a. once immediately.

b. once within 30 minutes.

c. twice immediately.

d. twice within 30 minutes.

SENIOR REACTOR OPERATOR Page 64 QUESTION: 099 (1.00)

The plant is operating at 75% power. The following plant conditions exist:

- RHR Loop B operating in suppression pool cooling

- RHR A Waterleg Pump Motor Control Center failure

- RHR Loop A is filled and vented and on alternate keep- fill The current Technical Specification Operability for RHR A and RHR B is .

Containment Spray Suppression Pool Cooling LPCI

a. RHR A Inoperable Inoperable Inoperable RHR B Operable Operable Inoperable

b. RHR A Operable Operable Operable RHR B Operable Operable Inoperable

c. RHR A Inoperable Inoperable Inoperable RHR B Operable Operable Operable

d. RHR A Operable Operable Operable RHR B Operable Operable Operable

SENIOR REACTOR OPERATOR Page 65 QUESTION: 100 (1.00)

The plant is in Mode 5, refueling operations are in progress. A new fuel bundle is being moved from IFTS to the Reactor. A PLC failure on the Refuel Platform then occurs.

Which of the following is correct regarding use of the Refuel Platform in this condition?

a. In vessel fuel movement may continue in manual.

b. Complete the fuel move to the proper vessel location in override.

c. Place the new fuel in a designated RP-1 storage location in override.

d. No use of the Refuel Platform is permitted until the PLC is repaired.

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

SENIOR REACTOR OPERATOR Page 66 ANSWER: 001 (1.00) ANSWER: 006 (1.00)

a. c.

REFERENCE:

IOI-0003 and Technical Specification 3.4.1 IOI-11 NEW NEW FUNDAMENTAL HIGHER 295001K305 ..(KA's) 295016 2.1.32 ..(KA's)

ANSWER: 002 (1.00) ANSWER: 007 (1.00)

a. d.

REFERENCE:

ONI-SPI-H3, ONI-SPI-D2 ONI-P43, SOI-B33 NEW ARI-H13P680-0004-D8 FUNDAMENTAL BANK 295003 2.4.3 ..(KA's) FUNDAMENTAL 295018K303 ..(KA's)

ANSWER: 003 (1.00)

d. ANSWER: 008 (1.00)

REFERENCE:

a.

ONI-R42-1

REFERENCE:

BANK ONI-P52 Attachment 1 FUNDAMENTAL MODIFIED 295004K105 ..(KA's) HIGHER 295019A102 ..(KA's)

ANSWER: 004 (1.00)

b. ANSWER: 009 (1.00)

REFERENCE:

b.

SDM 41/51

REFERENCE:

BANK IOI-12 FUNDAMENTAL BANK 295005K304 ..(KA's) FUNDAMENTAL 295021K301 ..(KA's)

ANSWER: 005 (1.00)

ANSWER: 010 (1.00)

b. a.

REFERENCE:

ONI-N62 IOI-9 MODIFIED SOI-G41(FPCC)

HIGHER BANK 295006A206 ..(KA's) HIGHER 295023K102 ..(KA's)

SENIOR REACTOR OPERATOR Page 67 ANSWER: 011 (1.00) ANSWER: 016 (1.00)

c. b.

REFERENCE:

Technical Specification 3.6.5.4 PEI T23 NEW BANK FUNDAMENTAL HIGHER 295024K101 ..(KA's) 295030A201 ..(KA's)

ANSWER: 012 (1.00) ANSWER: 017 (1.00)

a. d.

REFERENCE:

SDM B21/N11 ARI-H13P680-05-A1 BANK BANK HIGHER HIGHER 295025K309 ..(KA's) 295031K210 ..(KA's)

ANSWER: 013 (1.00) ANSWER: 018 (1.00)

b. d.

REFERENCE:

PEI Bases ARI-H13P680-05-A2 MODIFIED NEW HIGHER HIGHER 295026K201 ..(KA's) 295037K207 ..(KA's)

ANSWER: 014 (1.00) ANSWER: 019 (1.00)

d. d.

REFERENCE:

PEI Bases PEI-Bases D17 NEW BANK FUNDAMENTAL FUNDAMENTAL 295027A102 ..(KA's) 295038K302 ..(KA's)

ANSWER: 015 (1.00) ANSWER: 020 (1.00)

c. c.

REFERENCE:

PEI-SPI Supplement ONI-P54 BANK ARI-H13P904-01-A4 HIGHER NEW 295028A203 ..(KA's) FUNDAMENTAL 600000A216 ..(KA's)

SENIOR REACTOR OPERATOR Page 68 ANSWER: 021 (1.00) ANSWER: 026 (1.00)

a. a.

REFERENCE:

208-055 sheet 32 and 7 PEI-SPI Supplement Figure 4 NEW NEW HIGHER HIGHER 295007K203 ..(KA's) 295029A202 ..(KA's)

ANSWER: 022 (1.00) ANSWER: 027 (1.00)

c. c.

REFERENCE:

ARI-H13P680-03-A8 PEI Bases ARI-H13P680-05-A9 NEW NEW HIGHER FUNDAMENTAL 295036A102 ..(KA's) 295008K202 ..(KA's)

ANSWER: 028 (1.00)

ANSWER: 023 (1.00) c.

a.

REFERENCE:

ONI-R10 ARI-H13P601-20-E4 and F4 ONI-SPI A1 and A3 BANK BANK FUNDAMENTAL HIGHER 295011K101 ..(KA's) 203000K603 ..(KA's)

ANSWER: 024 (1.00) ANSWER: 029 (1.00)

b. a.

REFERENCE:

Technical Specification Bases 3.3.1.1 PDB-I0005 NEW BANK FUNDAMENTAL HIGHER 295014K301 ..(KA's) 205000A106 ..(KA's)

ANSWER: 025 (1.00) ANSWER: 030 (1.00)

d. c.

REFERENCE:

ARI-H13P601-19-B2 ONI-E12-1 BANK NEW HIGHER HIGHER 295017A106 ..(KA's) 209001 2.4.11 ..(KA's)

SENIOR REACTOR OPERATOR Page 69 ANSWER: 031 (1.00) ANSWER: 036 (1.00)

b. d.

REFERENCE:

208-065 sheet 3 and 12 ARI-H13-P680-06-E5 BANK SOI-C51(APRM)

HIGHER BANK 209002K101 ..(KA's) HIGHER 215005A406 ..(KA's)

ANSWER: 032 (1.00)

c. ANSWER: 037 (1.00)

REFERENCE:

b.

Technical Specification 3.1.7

REFERENCE:

BANK RCIC Pump Curves HIGHER TAF81834 211000 2.2.24 ..(KA's) BANK FUNDAMENTAL 217000A102 ..(KA's)

ANSWER: 033 (1.00) c.

REFERENCE:

ANSWER: 038 (1.00)

PDB Tab H Load Lists Tab 14 and 15 d.

NEW

REFERENCE:

FUNDAMENTAL ARI-H13P601-19-A9 212000K201 ..(KA's) BANK HIGHER 218000K501 ..(KA's)

ANSWER: 034 (1.00) d.

REFERENCE:

ANSWER: 039 (1.00)

ARI-H13P680-06-C2 c.

MODIFIED

REFERENCE:

HIGHER PDB-I0005 215003K304 ..(KA's) NEW HIGHER 223002K108 ..(KA's)

ANSWER: 035 (1.00) d.

REFERENCE:

ANSWER: 040 (1.00)

ARI-H13P680-06-C1 d.

SOI-C51 SRM

REFERENCE:

BANK PDB-I0005 HIGHER ARI-H13P601-19-A3 215004K604 ..(KA's) MODIFIED HIGHER 223002A302 ..(KA's)

SENIOR REACTOR OPERATOR Page 70 ANSWER: 041 (1.00) ANSWER: 046 (1.00)

c. b.

REFERENCE:

302-271 IOI-0003 NEW MODIFIED FUNDAMENTAL HIGHER 239002K301 ..(KA's) 262001A404 ..(KA's)

ANSWER: 042 (1.00) ANSWER: 047 (1.00)

a. a.

REFERENCE:

Technical Specification 3.6.2.1 ARI-H13P680-06-A4 BANK PDB-H008 FUNDAMENTAL BANK 239002A404 ..(KA's) FUNDAMENTAL 262002K602 ..(KA's)

ANSWER: 043 (1.00)

c. ANSWER: 048 (1.00)

REFERENCE:

c.

REFERENCE:

SOI-C34 ONI-SPI D1 NEW NEW HIGHER FUNDAMENTAL 259002A101 ..(KA's) 263000 2.4.34 ..(KA's)

ANSWER: 044 (1.00) ANSWER: 049 (1.00)

c. d.

REFERENCE:

912-605 SOI-R43 NEW ARI-H13P877-01-C2 FUNDAMENTAL NEW 261000K404 ..(KA's) HIGHER 264000A207 ..(KA's)

ANSWER: 045 (1.00)

c. ANSWER: 050 (1.00)

REFERENCE:

c.

ARI-H13P877-02-B1

REFERENCE:

G4, ARI-H13P601-22-D2 SOI-R43 NEW BANK HIGHER HIGHER 262001A211 ..(KA's) 264000A404 ..(KA's)

SENIOR REACTOR OPERATOR Page 71 ANSWER: 051 (1.00) ANSWER: 056 (1.00)

a. d.

REFERENCE:

ONI-P52 ARI-H13P680-04-A5 BANK ARI-H13P680-04-B5 FUNDAMENTAL NEW 300000K302 ..(KA's) HIGHER 202002A402 ..(KA's)

ANSWER: 052 (1.00)

d. ANSWER: 057 (1.00)

REFERENCE:

b.

SOI-P43

REFERENCE:

NEW E SOI-M51/56 FUNDAMENTAL OT-3401-000-05 400000K102 ..(KA's) NEW HIGHER 223001K509 ..(KA's)

ANSWER: 053 (1.00) c.

REFERENCE:

ANSWER: 058 (1.00)

ARI-H13P680-06-A2 c.

PDB-A006

REFERENCE:

NEW ARI-H13P601-20-A4 FUNDAMENTAL NEW OPRM K4.02 (KAs) HIGHER 226001K409 ..(KA's)

ANSWER: 054 (1.00)

d. ANSWER: 059 (1.00)

REFERENCE:

a.

SOI-C11(RC&IS)

REFERENCE:

BANK ONI-B21-1 HIGHER SVI-B21-T2005 201003A201 ..(KA's) NEW HIGHER 239001A109 ..(KA's)

ANSWER: 055 (1.00) a.

REFERENCE:

ANSWER: 060 (1.00)

PDB-H006 b.

BANK

REFERENCE:

FUNDAMENTAL 208-045 202001K201 ..(KA's) 208-151 BANK HIGHER 241000A408 ..(KA's)

SENIOR REACTOR OPERATOR Page 72 ANSWER: 061 (1.00) ANSWER: 066 (1.00)

c. b.

REFERENCE:

ARI-H13P680-08-B6 PYBP-POS-1-5 ARI-H13P680-07-D9 BANK NEW HIGHER HIGHER 2.1.1 ..(KA's) 245000A312 ..(KA's)

ANSWER: 067 (1.00)

ANSWER: 062 (1.00) c.

a.

REFERENCE:

NOP-OP-1002 ARI-H13P870-04-C3 NEW NEW FUNDAMENTAL HIGHER 2.1.2 ..(KA's) 256000K106 ..(KA's)

ANSWER: 068 (1.00)

ANSWER: 063 (1.00) b.

c.

REFERENCE:

PDB-C005 OAI-1703 attachment 11 BANK SOI-C34 HIGHER BANK 2.1.25 ..(KA's)

HIGHER 259001A308 ..(KA's)

ANSWER: 069 (1.00) a.

ANSWER: 064 (1.00)

REFERENCE:

a. NOP-OP-1001

REFERENCE:

NEW ONI-D17 FUNDAMENTAL NEW 2.2.13 ..(KA's)

FUNDAMENTAL 272000K120 ..(KA's)

ANSWER: 070 (1.00) d.

ANSWER: 065 (1.00)

REFERENCE:

a. SOI-F15

REFERENCE:

BANK SOI-P54(WTR) FUNDAMENTAL NEW 2.2.26 ..(KA's)

FUNDAMENTAL 286000K407 ..(KA's)

SENIOR REACTOR OPERATOR Page 73 ANSWER: 071 (1.00) ANSWER: 076 (1.00)

b. c.

REFERENCE:

NOP-OP-1002 Technical Specifications 3.1.5 NEW BANK FUNDAMENTAL HIGHER 2.3.2 ..(KA's) 2.1.11 ..(KA's)

ANSWER: 072 (1.00) ANSWER: 077 (1.00)

a. a.

REFERENCE:

ARI-H13P970-01-A8 Technical Specification 2.1 Safety Limits BANK BANK FUNDAMENTAL HIGHER 2.3.11 ..(KA's) 2.1.32 ..(KA's)

ANSWER: 073 (1.00) ANSWER: 078 (1.00)

d. c.

REFERENCE:

PEI-Bases PAP-1402 NEW NEW FUNDAMENTAL HIGHER 2.4.2 ..(KA's) 2.2.14 ..(KA's)

ANSWER: 074 (1.00) ANSWER: 079 (1.00)

b. d.

REFERENCE:

PEI-Bases PAP-1604 NEW MODIFIED FUNDAMENTAL HIGHER 2.4.5 ..(KA's) BANK FUNDAMENTAL 2.3.1 ..(KA's)

ANSWER: 075 (1.00) b.

REFERENCE:

ANSWER: 080 (1.00)

ONI-R61 and ARI-H13P680-07-E15 a.

NEW

REFERENCE:

FUNDAMENTAL PEI-Bases 2.4.32 ..(KA's) BANK HIGHER 2.3.8 ..(KA's)

SENIOR REACTOR OPERATOR Page 74 ANSWER: 081 (1.00) ANSWER: 086 (1.00)

b. a.

REFERENCE:

PEI-Bases PEI-Bases BANK NEW HIGHER FUNDAMENTAL 2.4.6 ..(KA's) 295025A204 ..(KA's)

ANSWER: 082 (1.00) ANSWER: 087 (1.00)

c. b.

REFERENCE:

EPI-A1, Emergency Action Levels PEI-Bases BANK BANK HIGHER FUNDAMENTAL 2.4.41 ..(KA's) 295026A201 ..(KA's)

ANSWER: 083 (1.00) ANSWER: 088 (1.00)

d. d.

REFERENCE:

PEI-Bases PEI-SPI Supplement NEW BANK HIGHER HIGHER 295005A204 ..(KA's) 295028A201 ..(KA's)

ANSWER: 084 (1.00) ANSWER: 089 (1.00)

a. b.

REFERENCE:

Technical Specification 3.5.1 ARI-H13P604-01-A4 and A5 BANK NEW HIGHER HIGHER 295019 2.2.23 ..(KA's) 295038 2.4.10 ..(KA's)

ANSWER: 085 (1.00) ANSWER: 090 (1.00)

c. c.

REFERENCE:

EPI-A1 Fission Product Barrier Matrix PEI-Bases NEW BANK HIGHER HIGHER 295024 2.4.45 ..(KA's) 295009A201 ..(KA's)

SENIOR REACTOR OPERATOR Page 75 ANSWER: 091 (1.00) ANSWER: 096 (1.00)

b. d.

REFERENCE:

ONI-C71-1 Technical Specification 3.3.1.1 and 1.0 PEI-SPI 1.3 BANK BANK HIGHER HIGHER 212000A203 ..(KA's) 295015 2.1.7 ..(KA's)

ANSWER: 097 (1.00)

ANSWER: 092 (1.00) b.

c.

REFERENCE:

Technical Specification 3.5.1 and 3.5.3 PDB-I0005 SOI-E51 Technical Specification 3.4.10, ONI-C71-2 BANK NEW HIGHER HIGHER 217000 2.1.12 ..(KA's) 295020A206 ..(KA's)

ANSWER: 098 (1.00)

ANSWER: 093 (1.00) d.

b.

REFERENCE:

SVI-B21-T1176 PDB-G0001 Technical Specification 3.4.11 SOI-E21 BANK MODIFIED HIGHER HIGHER 216000 2.2.12 ..(KA's) 209001A208 ..(KA's)

ANSWER: 099 (1.00)

ANSWER: 094 (1.00) a.

a.

REFERENCE:

SOI-E12 Technical Specification 3.3.1.2 NEW Core ALT definition HIGHER NEW 219000A205 ..(KA's)

FUNDAMENTAL 215004 2.2.27 ..(KA's)

ANSWER: 100 (1.00) c.

ANSWER: 095 (1.00)

REFERENCE:

d. SOI-F15

REFERENCE:

Technical Specification 3.9.1 Technical Specification 3.3.1.1, ORM 6.5.4 ORM 6.2.5, ARI-H13P680-06-B5 and C4 NEW BANK HIGHER HIGHER 234000A201 ..(KA's) 215005A205 ..(KA's)

SENIOR REACTOR OPERATOR Page 76

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

SENIOR REACTOR OPERATOR Page 77 ANSWER KEY MULTIPLE CHOICE 001 a 021 a 041 c 061 c 081 b 002 a 022 c 042 a 062 a 082 c 003 d 023 a 043 c 063 c 083 d 004 b 024 b 044 c 064 a 084 a 005 b 025 d 045 c 065 a 085 c 006 c 026 a 046 b 066 b 086 a 007 d 027 c 047 a 067 c 087 b 008 a 028 c 048 c 068 b 088 d 009 b 029 a 049 d 069 a 089 b 010 a 030 c 050 c 070 d 090 c 011 c 031 b 051 a 071 b 091 b 012 a 032 c 052 d 072 a 092 c 013 b 033 c 053 c 073 d 093 b 014 d 034 d 054 d 074 b 094 a 015 c 035 d 055 a 075 b 095 d 016 b 036 d 056 d 076 c 096 d 017 d 037 b 057 b 077 a 097 b 018 d 038 d 058 c 078 c 098 d 019 d 039 c 059 a 079 d 099 a 020 c 040 d 060 b 080 a 100 c

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