ML081920353

2008 NRC RO and SRO Initial Examination - Corrected

ML081920353
Person / Time
Site:	Fermi
Issue date:	07/11/2008
From:	Hironori Peterson Operations Branch III
To:	Jennifer Davis Detroit Edison
References
50-341/08-301 50-341/08-301
Download: ML081920353 (93)

Text

QUESTION 1 Following a Loss of Coolant Accident, plant conditions are as follows:

• RPV Water Level is -25 inches.

• Reactor Pressure is 50 psig.

• The Reactor Building is NOT accessible due to High Radiation Levels.

Which ONE of the following paths can be used to inject from the Condensate Storage Tank (CST) into the reactor, using ONLY Main Control Room manipulations?

The CST can be injected by using the:

_____ A. Residual Heat Removal System with suction aligned from the Torus.

_____ B. Standby Liquid Control System with suction aligned from the Test Tank.

_____ C. Core Spray System with suction aligned from the Condensate Storage Tank.

_____ D. High Pressure Coolant Injection System with suction aligned from the Condensate Storage Tank.

QUESTION 2 The plant is in MODE 4, COLD SHUTDOWN, conditions are as follows:

• RPV Water Level is 230 inches.

• Reactor Coolant Temperature is 175°F.

• RHR Loop B is operating in Shutdown Cooling Mode at 9,200 gpm.

• Reactor Coolant System cooldown rate is 90°F per hour.

• E1150-F003B, Div 2 RHR Hx Outlet Vlv is THROTTLED OPEN 60 SECONDS.

• E1150-F048B, Div 2 RHR Hx Bypass Vlv is OPEN.

Which ONE of the following will REDUCE cooldown rate?

_____ A. FULLY OPEN E1150-F003B, Div 2 RHR Hx Outlet Valve.

_____ B. FULLY SHUT E1150-F048B, Div 2 RHR Hx Bypass Valve.

_____ C. THROTTLE SHUT E1150-F048B, Div 2 RHR Hx Bypass Valve.

_____ D. THROTTLE SHUT E1150-F003B, Div 2 RHR Hx Outlet Valve.

QUESTION 3 Following a transient, HPCI is being used to maintain RPV Water Level per 29.100.01 Sheet 1, RPV Control. Alarm 2D54, HPCI INVERTER CIRCUIT FAILURE actuates with the following indications failing DOWNSCALE:

• E41-R609, HPCI Pump Suction/Discharge Pressure.

• E41-R608, HPCI Turbine Steam Inlet/Exhaust Pressure.

• E41-R613, HPCI Pump Flow Indicator.

• E41-K615, HPCI Flow Controller.

With these indications, which ONE of the following caused the loss of AC power from the HPCI Inverter, and what actions are required?

AC power from the HPCI Inverter was produced by a LOSS of:

_____ A. Division 1 130 VDC power, it is required to OPERATE HPCI to maintain RPV Water Level.

_____ B. Division 1 130 VDC power, it is required to SHUTDOWN HPCI and use RCIC to maintain RPV Water Level.

_____ C. Division 2 130 VDC power, it is required to OPERATE HPCI to maintain RPV Water Level.

_____ D. Division 2 130 VDC power, it is required to SHUTDOWN HPCI and use RCIC to maintain RPV Water Level.

QUESTION 4 With the plant operating at full power, a relay malfunction resulted in the following annunciators:

• 1D1, DIV I CSS ACTUATED.

• 1D48, ADS ECCS PUMP CH A PERMISSIVE.

What will be the affect, if any, of this failure on Emergency Diesel Generators?

_____ A. Emergency Diesel Generators 11 and 12 will remain in STANDBY.

_____ B. Emergency Diesel Generators 11 and 12 will START and LOAD with ALL trips active.

_____ C. Emergency Diesel Generators 11 and 12 will START and LOAD with ONLY essential trips active.

_____ D. Emergency Diesel Generators 11 and 12 will START and operate UNLOADED with ONLY essential trips active QUESTION 5 Which ONE of the following Reactor Pressure Vessel piping taps is SHARED by the Standby Liquid Control System Injection Line?

_____ A. Jet Pump Differential Pressure tap

_____ B. Core Plate Differential Pressure tap

_____ C. Core Spray Line Break Detection tap

_____ D. Control Rod Drive Water Differential Pressure tap QUESTION 6 Following a manual reactor scram the following occurred:

• The Reactor Mode Switch was placed in SHUTDOWN.

• Scram Reset Switch, C7100-M605 was turned to the GP 1/4 AND GP 2/3 positions and released.

• SRM and IRM detectors were selected and the DRIVE IN pushbutton was depressed.

• A few minutes later a SECOND automatic scram signal was received.

• ALL RPV and Containment parameters remained constant prior to the second scram.

What was the cause of the SECOND scram and why did it occur?

_____ A. Alarm 3D51, SRM PERIOD SHORT, was received due to driving in SRM and IRM detectors and an automatic scram resulted.

_____ B. Alarm 3D97, APRM NEUTRON FLUX UPSCALE TRIP, was received due to the production of delayed neutrons from delayed neutron precursors.

_____ C. Alarm 3D86, MN STM LINE ISO VALVE CLOSURE CHANNEL TRIP, was received due to the failure to adjust Gland Seal Pressure resulting in an MSIV isolation on loss of vacuum and the subsequent scram.

_____ D. Alarm 3D94, DISCH WATER VOL HI LEVEL CHANNEL TRIP, was received and the reactor scram occurred because the SDV High Level Bypass Switch was not placed in BYPASS and the SDV filled faster than it drained after the first scram was reset.

QUESTION 7 A reactor startup is in progress. The reactor has been declared critical and the operator has established a 150 sec period. ALL IRMs are at 50/125 on range 4.

The following indications are observed:

• 3D63, IRM UPSCALE alarms.

• 3D59, IRM CH A/E/C/G UPSCALE TRIP/INOP alarms.

• 3D73, TRIP ACTUATORS A1/A2 TRIPPED alarms.

• 3D113, CONTROL ROD WITHDRAWAL BLOCK alarm These indications were CAUSED by:

_____ A. IRM E power supply failure

_____ B. IRM E being ranged to range 3.

_____ C. IRM E being ranged to range 5.

_____ D. IRM E being withdrawn from the core.

QUESTION 8 Which ONE of the following PROVIDES POWER for the Intermediate Range Channel B instrument drawer?

_____ A. 120/208 VAC Cabinet 72E-2B-1

_____ B. 120 VAC UPS Distribution Cabinet B

_____ C. 48/24 VDC DC Distribution Cabinet 2IA1-3

_____ D. 48/24 VDC DC Distribution Cabinet 2IB1-3 QUESTION 9 A reactor startup is in progress, the following conditions exist:

• The Mode Switch is in the START & HOT STBY position.

• NO SRMs are bypassed.

• The SRM detectors are PARTIALLY WITHDRAWN.

Which ONE of the following sets of conditions will RESULT in a rod withdrawal block?

_____ A. IRMs on range 3. ALL SRMs indicating 90 cps.

_____ B. IRMs on range 1. ALL SRMs indicating 120 cps.

_____ C. IRMs on range 2. SRM "A" indicating 90 cps, ALL OTHER SRMs indicating 150 cps.

_____ D. IRMs on range 4. SRM "A" indicating 90 cps, ALL OTHER SRMs indicating 120 cps.

QUESTION 10 Plant shutdown is in progress. A control rod is being inserted from position 48 to position 00. The B Level LPRM readings will most SIGNIFICANTLY DECREASE as the rod passes through positions _______.

_____ A. 08 to 04

_____ B. 20 to 16

_____ C. 32 to 28

_____ D. 44 to 40

QUESTION 11 Reactor Core Isolation Cooling (RCIC) is operating in the TEST MODE with the following conditions:

• E51-R614, RCIC Pump Flow Controller is in AUTOMATIC.

• RCIC Turbine Speed is 2950 rpm.

• P1100-F606, CST Common Return Isolation Valve is OPEN.

• E41-K820, Test Isolation/PCV E41-F011 Controller, is in MANUAL at 20%

OPEN.

Which ONE of the following describes the STABILIZED response of RCIC Turbine Speed AND system flow AFTER PCV E41-F011 is THROTTLED an ADDITIONAL 5% in the CLOSED direction?

_____ A. RCIC Turbine SPEED will be HIGHER System indicated FLOW will be HIGHER

_____ B. RCIC Turbine SPEED will be LOWER System indicated FLOW will be LOWER

_____ C. RCIC Turbine SPEED will be HIGHER System indicated FLOW will be AT THE INITIAL VALUE

_____ D. RCIC Turbine SPEED will be LOWER System indicated FLOW will be AT THE INITIAL VALUE

QUESTION 12 Ten minutes ago, the Primary Containment Pneumatic Supply System DIV I Inboard and Outboard Isolation Valves, T4901-F601 AND F465 SHUT.

How are the Automatic Depressurization System Valves affected?

_____ A. ADS Valves WILL operate if logic is actuated WITHOUT any further operator action.

_____ B. ADS Valves WILL NOT operate if logic is actuated. Operators MUST use Alternate Depressurization systems.

_____ C. ADS Valves WILL operate if logic is actuated. NIAS automatically aligns, WITHOUT any further operator action.

_____ D. ADS Valves WILL NOT operate if logic is actuated. Operators MUST clear and reset isolation and realign Nitrogen to the Drywell.

QUESTION 13 While monitoring the Primary Containment Isolation System (PCIS) Group Isolation mimic (ISO MIMIC) on P601, an Isolation signal is received.

For the statements below, which ONE is correct for valve groups that have received an isolation signal and the isolation is complete?

When the Isolation signal has initiated, for a group of valves NOT wired in series, a:

_____ A. GREY ISO SIG PRESENT message is displayed on the P601 PCIS mimic to indicate an isolation signal is present and when ALL isolation valves in that PCIS Group are CLOSED, they indicate RED.

_____ B. GREY ISO SIG PRESENT message is displayed on the P601 PCIS mimic to indicate an isolation signal is present and when ALL isolation valves in that PCIS Group are CLOSED, they indicate GREEN.

_____ C. YELLOW ISO SIG PRESENT message is displayed on the P601 PCIS mimic to indicate an isolation signal is present. When ALL isolation valves in that PCIS are CLOSED, they indicate RED.

_____ D. YELLOW ISO SIG PRESENT message is displayed on the P601 PCIS mimic to indicate an isolation signal is present. When ALL isolation valves in that PCIS are CLOSED, they indicate GREEN.

QUESTION 14 Safety Relief Valve G Tailpipe Vacuum Breaker is STUCK OPEN.

Which ONE of the following affects will result from this condition?

_____ A. Steam will be released to the TORUS when SRV G is OPEN.

_____ B. Steam will be released to the DRYWELL when SRV G is OPEN.

_____ C. Damage may occur to the SRV TAILPIPE when SRV G is OPEN.

_____ D. Water will be drawn up the SRV TAILPIPE after SRV G is SHUT.

QUESTION 15 Which ONE of the following indicates an OPEN Safety Relief Valve?

_____ A. Reactor Thermal Power LOWERING from 3430 Mwt to 3258 Mwt.

_____ B. Safety Relief Valve Tailpipe Temperature RISING from 170°F to 290°F.

_____ C. Total Indicated Steam Flow RISING from 13.4 Mlbm/hr to 14.1 Mlbm/hr.

_____ D. Total Feed flow LOWERING to 13.4 Mlbm/hr WITH Total Steam Flow at 14.1 Mlbm/hr.

QUESTION 16 During a startup, the North RFPT is operating.

The following conditions exist:

• Reactor Power is 1% CTP.

• Reactor pressure is 650 psig.

• SULCV is 40% open.

• SULCV M/A Station is in AUTO.

• North Feedwater Flow Control M/A Station is in MANUAL.

• The Interruptible Air Supply to the SULCV is LOST.

(1) How will this failure FIRST affect RPV Water Level? AND (2) Which ONE of the following actions will mitigate this failure?

_____ A. (1) RPV Water Level will RISE.

(2) PLACE the C32-R620, N21-F403 RPV Startup Level Controller SULCV M/A Station in MANUAL and lower the OUTPUT signal to CLOSE the SULCV.

_____ B. (1) RPV Water Level will RISE.

(2) TRIP the North RFPT, START the West Standby Feedwater Pump and control RPV Water Level using N2103-F003, SBFW 4" Disch Flow Ctrl Vlv.

_____ C. (1) RPV Water Level will LOWER.

(2) START the West Standby Feedwater Pump and control RPV Water Level using N2103-F003, SBFW 4" Disch Flow Ctrl Vlv.

_____ D. (1) RPV Water Level will LOWER.

(2) PLACE the Feedwater Flow Control M/A Station is in AUTO and OPEN N2100-F607, N RFP Disch Line Iso Valve.

QUESTION 17 Both HPCI and SGTS received an auto start signal due to Low RPV Water Level.

How would the HPCI system respond to a LOSS of SGTS?

_____ A. HPCI continues to operate properly, since Barometric Condenser Vacuum Pump is NOT REQUIRED for operation.

_____ B. HPCI cannot operate properly without a discharge path for the Barometric Condenser Vacuum Pump and will AUTOMATICALLY TRIP.

_____ C. HPCI cannot operate properly without a discharge path for the Barometric Condenser Vacuum Pump and must be MANUALLY TRIPPED.

_____ D. HPCI continues to operate properly because automatic trips associated with the Barometric Condenser Vacuum pump are automatically BYPASSED.

QUESTION 18 The plant is operating at full power with Div 2 SGTS OUT OF SERVICE for maintenance. The following indications occur:

• 3D85, PRIMARY CONTAINMENT HIGH PRESS CHANNEL TRIP, alarms.

• 8D35, DIV I SGTS AIR FLOW STOPPED, alarms

• T4600-F004A, Div 1 SGTS Exhaust Fan Inlet Isolation Damper, is CLOSED.

Which ONE of the following describes the affect of these conditions, if any, and the Emergency Operating Procedure Leg required to mitigate the condition, if any?

_____ A. NO CHANGE IN DIFFERENTIAL PRESSURE between the Reactor Building and the environs, NO EOP usage is required.

_____ B. INCREASING DIFFERENTIAL PRESSURE between the Reactor Building and the environs, which would FIRST result in an Entry Condition for the Secondary Containment Control EOP Leg.

_____ C. DECREASING DIFFERENTIAL PRESSURE between the Reactor Building and the environs, which would FIRST result in an Entry Condition for the Radiation Release Control EOP Leg.

_____ D. DECREASING DIFFERENTIAL PRESSURE between the Reactor Building and the environs, which would FIRST result in an Entry Condition for the Secondary Containment Control EOP Leg.

QUESTION 19 When operating the 480V ESF Bus Maintenance Tie Breakers for Live Bus Transfer, the 4160V ESF Buses are verified powered from their normal offsite power source.

Attempting LOCAL MANUAL operation of these breakers with either bus powered from an EDG may result in ____________.

_____ A. no breaker closure

_____ B. an overspeed trip of an EDG

_____ C. a sustained overload condition of the EDG

_____ D. equipment damage from paralleling out of phase QUESTION 20 Reactor Pressure is 400 psig and RHR Loop B is running in response to a valid LPCI initiation signal.

Which ONE of the following is the indicated flow on the Division II RHR System Flow Recorder?

_____ A. 0 gpm

_____ B. 3,000 gpm

_____ C. 10,000 gpm

_____ D. 20,000 gpm

QUESTION 21 With a loss of Division 2 ESF 260/130 VDC Batteries and Chargers, which ONE of the following will result?

_____ A. Breakers on 4160V Busses 64B and 11EA will lose control power.

_____ B. Breakers on 4160V Busses 65E and 13EC will lose control power.

_____ C. MCC 72CF Feed will auto throw-over from 72C Pos 3C to 72F Pos 5C.

_____ D. C11-F110B, Scram Pilot Air Header Backup Scram Valve, will actuate.

QUESTION 22 With the plant operating at full power, 10D72, BOP 260/130V BATTERY 2PC TROUBLE alarms. A COMPLETE LOSS of BOP DC has occurred.

Which ONE of the following operator actions is required under this condition?

_____ A. TRIP Breakers CM and CF using COP H11-P804 control switches.

_____ B. TRIP the Generator Field Breaker using COP-H11-P804 control switch.

_____ C. TRIP Breakers CM and CF using LOCAL Emergency Trip pushbuttons.

_____ D. TRIP the Turbine Generator by ARMING AND DEPRESSING the Turbine Trip pushbutton.

QUESTION 23 Following a Loss of Offsite Power, the grid has been restored, conditions are as follows:

• EDG 11 is supplying ESF Bus 64B by way of EDG Bus 11EA and ESF-EDG Bus Tie Breaker B8.

• Synchroscope Switch for ESF Bus 64B Normal Feeder Breaker B6 is ON.

• ALL conditions are met for paralleling the EDG with Offsite Power.

When the B6 Breaker is CLOSED, the operating mode of EDG 11 will shift to the

___(1)___ mode.

Following breaker closure, the EDG 11 Governor must be immediately adjusted to prevent a(n) ___(2)___ condition.

_____ A. (1) Speed Droop (2) overload

_____ B. (1) Speed Droop (2) reverse power

_____ C. (1) Isochronous (2) overload

_____ D. (1) Isochronous (2) reverse power QUESTION 24 Which ONE of the following Air Compressors will DIRECTLY TRIP due to a Low Cooling Water Flow signal? (NOT a High Temperature signal.)

_____ A. D001, Control Air Compressor

_____ B. D002, Control Air Compressor

_____ C. D001, East Station Air Compressor

_____ D. D002, Center Station Air Compressor

QUESTION 25 Interruptible Air System (IAS) Air Header Pressure supplied to the RBCCW Temperature Control Valve and Differential Pressure Control Valve is LOWERING due to an Air Header leak causing RBCCW Air Operated Valves to move towards their FAIL position.

How will Non Interruptible Air System (NIAS) Aftercooler Air Temperature be affected by this condition?

Aftercooler Air Temperature will:

_____ A. RISE due to Differential Pressure Control Valve failing SHUT.

_____ B. LOWER due to Differential Pressure Control Valve failing OPEN.

_____ C. RISE due to RBCCW Temperature Control Valve failing SHUT.

_____ D. LOWER due RBCCW Temperature Control Valve failing OPEN.

QUESTION 26 With the plant operating at 100% power, when the following occurs:

• 9D10, DIV 1 480 V ESS BUS 72C BKR TRIPPED, alarms.

• CMC Switch for BUS 64C POS C11, 4160V FEED TO BUS 72C, indicates TRIPPED.

Given these indications, which ONE of the following correctly describes the impact on the Reactor Building Closed Cooling Water (RBCCW) Pumps?

_____ A. ONLY P4200-C001, North RBCCW Pump, has lost power.

_____ B. ONLY P4200-C003, South RBCCW Pump, has lost power.

_____ C. BOTH P4200-C001 AND P4200-C002, North and Center RBCCW Pumps, have lost power.

_____ D. BOTH P4200-C003 AND P4200-C002, South and Center RBCCW Pumps, have lost power.

QUESTION 27 The reactor is at 5% power during a plant startup. While a control rod is being withdrawn, Rod Select Power is LOST.

Which ONE of the following describes the affect of this loss?

When Rod Select Power is LOST, the control rod motion will:

_____ A. STOP; the control rod may eventually settle due to leakage but NO settle function will occur.

_____ B. STOP; the control rod settles to the next notch as the Settle Bus is automatically energized for 4.4 seconds.

_____ C. CONTINUE UNTIL the Rod Movement Control Switch is released; the control rod settles to the next notch as the Settle Bus is automatically energized for 4.4 seconds.

_____ D. CONTINUE ONLY if the Rod Out Notch Override Switch is positioned to Emergency In; the control rod may eventually settle due to leakage but NO settle function will occur.

QUESTION 28 Following a transient, the following conditions exist:

• RPV Water Level is 192 inches.

• Reactor Pressure is 1000 psig.

• Blowdown Mode of RWCU is being used as an Alternate Pressure Control System per EOP-1 RPV Control.

• Pressure on the Blowdown Line between RWCU and the Main Condenser is RISING.

Which ONE of the following describes the FIRST AUTOMATIC response of RWCU valves to this condition?

_____ A. BEFORE piping failure occurs, the G3300-F033, RWCU Blowdown FCV will automatically close due to Pressure Upstream of F033.

_____ B. BEFORE piping failure occurs, the G3300-F033, RWCU Blowdown FCV will automatically close due to Pressure Downstream of F033.

_____ C. AFTER piping failure occurs, G3352-F001 and G3352-F004, RWCU Containment Isolation Valves will automatically close due to HIGH Differential Flow.

_____ D. AFTER piping failure occurs, G3352-F001 and G3352-F004, RWCU Containment Isolation Valves will automatically close due to HIGH Area Temperature.

QUESTION 29 The plant was operating at rated power when a manual reactor scram was inserted.

The following conditions exist:

• ALL control rods have fully inserted.

• The Reactor Mode Switch has been placed in SHUTDOWN.

• The scram has NOT been reset.

What are the MINIMUM actions necessary to reset the Control Rod Drift indications on the Full Core Display (vertical section of panel H11-P603)?

_____ A. MOMENTARILY rotate ROD DRIFT ALARM Switch to RESET.

_____ B. RESET the reactor scram. After Control Rods have settled at position 00, MOMENTARILY rotate ROD DRIFT ALARM Switch to RESET.

_____ C. RESET the reactor scram. SELECT each Control Rod with a drift alarm.

After Control Rod has settled at position 00, MOMENTARILY rotate ROD DRIFT ALARM Switch to RESET. REPEAT for each Control Rod with a drift alarm.

_____ D. SELECT each Control Rod with a drift alarm. MOMENTARILY place ROD MOVEMENT CONTROL Switch to OUT NOTCH. After Control Rod has settled at position 00, MOMENTARILY rotate ROD DRIFT ALARM Switch to RESET. REPEAT for each Control Rod with a drift alarm.

QUESTION 30 Traversing In-core Probe (TIP) Channel A Detector is INSERTING INTO THE CORE for calibration of Local Power Range Monitors (LPRMs). The IN CORE Light is LIT.

The CORE TOP Light is NOT LIT. A TRIP of ONE Reactor Feed Pump occurs and RPV Water Level lowers to 160 inches before recovering to 195 inches.

Which of the following describes the automatic TIP response?

_____ A. C51F001A, TIP Channel A Shear Valve FIRES, ISOLATING the drive mechanism.

_____ B. The TIP detector WITHDRAWS into the shield chamber, AND C51F002A, TIP Channel A Ball Valve, CLOSES.

_____ C. The TIP detector WITHDRAWS AND STOPS outboard of the Indexer, and C51F002A, TIP Channel A Ball Valve, CLOSES.

_____ D. The TIP drive CONTINUES TO INSERT the detector to the Core Top Limit AND completes the TIP trace. The detector then withdraws into the shield chamber and C51F002A, TIP Channel A Ball Valve, CLOSES.

QUESTION 31 With Reactor Power at 65%, the TRIP SETPOINT of the Rod Block Monitor is:

_____ A. 107.2 %; and if exceeded, a Control Rod Block WILL result.

_____ B. 107.2%; and if exceeded, a Control Rod Block WILL NOT result.

_____ C. 112.2 %; and if exceeded, a Control Rod Block WILL result.

_____ D. 112.2 %; and if exceeded, a Control Rod Block WILL NOT result.

QUESTION 32 Core reload is in progress. The following indications are observed after a Fuel Bundle has been inserted into the core:

• Source Range Monitors (SRMs) indicate 250 cps, slowly rising.

• SRM Period is slightly positive and stable.

• All control rods are fully inserted.

Which ONE of the following statements describes present conditions and what action should be performed?

The indicated conditions are:

_____ A. NORMAL. Determination of SRM signal to noise ratio should be directed.

_____ B. ABNORMAL. Immediate core unloading near the SRMs should be performed.

_____ C. NORMAL. Continued Core Reload should be directed per the Fuel Movement Sheets.

_____ D. ABNORMAL. Fuel Handling shall be terminated until a complete evaluation is performed

QUESTION 33 The plant was operating at normal pressure with C11-F002A, CRD Flow Control Valve A in service and in AUTOMATIC control. The following CRD indications were present:

• Cooling Water Flow is 45 gpm.

• Drive Water D/P is 235 psid.

• Cooling Water D/P is 15 psid.

• NO Rod Motion is in progress.

The P603 operator adjusts C1152-F003, CRD Drive/Cooling Water PCV in the CLOSED direction.

What will be the FINAL effect on the CRD System parameters?

_____ A. Drive Water D/P INCREASES and Cooling Water Flow DECREASES.

_____ B. Drive Water D/P DECREASES and Cooling Water Flow DECREASES.

_____ C. Drive Water D/P INCREASES and Cooling Water Flow REMAINS THE SAME.

_____ D. Drive Water D/P DECREASES and Cooling Water Flow REMAINS THE SAME.

QUESTION 34 With the plant operating at full power, when the following occurs:

• 4D97, GENERATOR BUS COOLING TEMPERATURE HIGH alarms.

• GENERATOR BUS CLG FAN DISCHARGE PRESSURE Red Light is ON.

Which ONE of the following failures is INDICATED by these conditions?

_____ A. 480 VAC Bus 72R has TRIPPED.

_____ B. ALL TBCCW Pumps have TRIPPED.

_____ C. Iso Phase Bus South Cooling Fan, S1200-C002, has TRIPPED.

_____ D. TBCCW Cooling Water Valve, P43-F209 has LOST Instrument Air.

QUESTION 35 With the plant in COLD SHUTDOWN, the following conditions exist:

• Control Rods are being exercised.

• Control Rod Drive is providing RPV Makeup.

• Reactor Water Cleanup Blowdown Valve is throttled 30%.

• RPV Water Level is 195 inches, STABLE.

Which ONE of the following describes the EXPECTED affect, if any, on RPV Water Level when the LAST operating Condenser Pump TRIPS?

_____ A. RPV Water Level will NOT CHANGE, due to a redundant suction supply.

_____ B. RPV Water Level will LOWER, due a Low Suction Pressure TRIP of the operating CRD Pump.

_____ C. RPV Water Level will LOWER THEN RISE, due the effect of suction transfer on the operating CRD Pump.

_____ D. RPV Water Level will LOWER THEN RISE, due a Low Suction Pressure TRIP of the operating CRD Pump followed by a Reactor Scram.

QUESTION 36 With the plant operating at full power, the following alarms and indications exist:

• 6D21, E/W OFF GAS RECOMBINER TEMPERATURE HIGH/LOW alarms.

• The West Off Gas Recombiner is in service and is indicating 805°F on N62-R815, Off Gas Components Temperature Recorder.

Which ONE of the following should be performed to control Off Gas Recombiner Temperature?

_____ A. VERIFY N62-F400, 18" Manifold Steam Supply TCV, is OPEN.

_____ B. VERIFY N62-F400, 18" Manifold Steam Supply TCV, is SHUT.

_____ C. VERIFY N6200-D010, West Off Gas Chiller Unit is RUNNING.

_____ D. VERIFY N62-N013A, C Thermostatic Controlled Electric Heaters, at 600°F.

QUESTION 37 Which ONE of the following provides power for D11-K609A, Fuel Pool (EAST) Vent Exhaust Duct Radiation Monitor?

_____ A. 24/48 VDC

_____ B. 130/260 VDC

_____ C. 120 VAC RPS

_____ D. 120 VAC UPS

QUESTION 38 The plant is in MODE 5 with movement of RECENTLY irradiated fuel in progress.

Due to a damper malfunction, Reactor Building Vacuum is 0 inches water gauge.

Which ONE of the following actions is REQUIRED by Technical Specifications?

_____ A. Suspend fuel movement IMMEDIATELY.

_____ B. Restore Secondary Containment Pressure WITHIN ONE HOUR.

_____ C. Start BOTH Divisions of Standby Gas Treatment System IMMEDIATELY.

_____ D. Verify at least ONE door is closed at each Reactor Building access WITHIN ONE HOUR.

QUESTION 39 Following a trip of ONE Reactor Recirculation Pump, why is it necessary to limit operating Recirculation Pump Speed to 75%?

_____ A. To PREVENT Recirculation Pump runout due to reduced backpressure.

_____ B. To PREVENT excessive vibration of Reactor Vessel internal components.

_____ C. To REDUCE Reactor Power to within the Technical Specification Limit for Single Loop Operation.

_____ D. To REDUCE APRM Simulated Thermal Power Trip Setpoints until the setpoints are adjusted for Single Loop Operation.

QUESTION 40 The reactor has scrammed due to a LOSS of Offsite Power.

ONLY EDGs 13 & 14 have started and loaded.

What is the SOURCE of power to the station DC loads?

_____ A. Div 1 Chargers are supplying Div 1 DC loads.

Div 2 Chargers are supplying Div 2 DC loads.

_____ B. Div 1 Batteries are supplying Div 1 DC loads.

Div 2 Chargers are supplying Div 2 DC loads.

_____ C. Div 1 Chargers are supplying Div 1 DC loads.

Div 2 Batteries are supplying Div 2 DC loads.

_____ D. Div 1 Batteries are supplying Div 1 DC loads.

Div 2 Batteries are supplying Div 2 DC loads.

QUESTION 41 The plant is operating at full power when the following annunciators and indications are received:

• 9D17, DIV I ESS 130 V BATTERY 2PA TROUBLE

• 9D21, DIV I EDG SEQUENCER TROUBLE

• 1D6, DIV I CSS LOGIC POWER FAILURE

• 1D8, RHR LOGIC A 125 VDC BUS POWER FAILURE

• 1D62, STM LK DET HPCI LOGIC POWER FAILURE

• Div I DC powered valves position indicating lights are OFF.

• Breaker position indicating lights for Div I ESF Bus breakers are OFF.

Based on these alarms, select the correct DIAGNOSIS AND AFFECT, if any, on Division I EDGs ability to mitigate a Loss of Offsite Power.

_____ A. ONLY the Division I Batteries have been lost. Division I EDGs will NOT START if Offsite Power is LOST.

_____ B. ONLY the Division I Batteries have been lost. Division I EDGs will AUTO START if Offsite Power is LOST.

_____ C. BOTH Division I Battery Chargers AND BOTH Division I Batteries have been lost. Division I EDGs will NOT START if Offsite Power is LOST.

_____ D. BOTH Division I Battery Chargers AND BOTH Division I Batteries have been lost. Division I EDGs will AUTO START if Offsite Power is QUESTION 42 The plant is at 35% power when the turbine trips. The reactor will ______________.

_____ A. SCRAM AND reactor pressure will INCREASE due to decay heat

_____ B. SCRAM AND reactor pressure will REMAIN CONSTANT due to BPV operation

_____ C. REMAIN OPERATING AND reactor power will INCREASE due to Feedwater Temperature change

_____ D. REMAIN OPERATING AND reactor power will DECREASE due to Feedwater Temperature change

QUESTION 43 With the plant operating at 85% power, the in-service 52 inch Manifold Pressure Transmitter fails HIGH.

Which ONE of the following describes the affect of this failure, with no operator actions?

Reactor Pressure and Reactor Power will:

_____ A. RISE until an automatic Reactor Scram occurs. Pressure will then be controlled by MANUAL operation of the Bypass Valves.

_____ B. RISE until an automatic Reactor Scram occurs. Pressure will then be controlled by AUTOMATIC operation of the Bypass Valves.

_____ C. LOWER until an automatic Reactor Scram occurs. Pressure will then be controlled by AUTOMATIC operation of the Bypass Valves.

_____ D. LOWER until an automatic Reactor Scram occurs. Pressure will then be controlled by AUTOMATIC operation of the Safety Relief Valves.

QUESTION 44 The plant has entered 20.000.19, SHUTDOWN FROM OUTSIDE THE CONTROL ROOM.

The transfer of certain CR controls to OUTSIDE the CR is needed to:

_____ A. continue to remove decay heat.

_____ B. allow rapid re-entry into the CR.

_____ C. preclude the effects of hot shorts.

_____ D. allow proper fire fighting response.

QUESTION 45 Foreign material has partially covered the tube sheet at the inlet to the Stator Cooling Water Heat Exchanger.

Which ONE of the following describes the AUTOMATIC response of the system?

The Temperature Control Valve throttling:

_____ A. Stator Water will reposition to permit more flow to BYPASS the Heat Exchanger.

_____ B. Stator Water will reposition to permit more flow THROUGH the Heat Exchanger.

_____ C. Turbine Building Closed Cooling Water will reposition to permit more flow to BYPASS the Heat Exchanger.

_____ D. Turbine Building Closed Cooling Water will reposition to permit more flow THROUGH the Heat Exchanger.

QUESTION 46 The plant is operating at 100% power.

• P50-R802, Station Air Header Pressure is 90 psig (lowering).

Which ONE of the following is required per 20.129.01, Loss of Station and / or Control Air and the reason for that action?

It is required to:

_____ A. START ANY available Station Air Compressors to prevent the INBOARD Main Steam Isolation Valves from drifting shut.

_____ B. START ANY available Station Air Compressors to prevent the OUTBOARD Main Steam Isolation Valves from drifting shut.

_____ C. CLOSE P5000-F401, Station Air to TB Hdr Iso Vlv to prevent the INBOARD Main Steam Isolation Valves from drifting shut.

_____ D. CLOSE P5000-F401, Station Air to TB Hdr Iso Vlv to prevent the OUTBOARD Main Steam Isolation Valves from drifting shut.

QUESTION 47 The plant is in MODE 4, with RHR Pump A operating in Shutdown Cooling Mode AND BOTH Reactor Recirculation Pumps shut down.

Due to a leak between E1150-F008, RHR SDC Outboard Isolation Valve and E1150-F009, RHR SDC Inboard Suction Isolation Valve, RPV Water Level lowered to 170 inches.

Per 23.800.04, Alternate Coolant Circulation and Decay Heat Removal, which ONE of the following will provide Core Circulation?

_____ A. Natural Circulation

_____ B. Reactor Recirculation Pump START

_____ C. Reactor Water Cleanup Pump START

_____ D. Residual Heat Removal Pump START QUESTION 48 With the plant in MODE 5, REFUELING, with Core Alterations is progress.

Which ONE of the following is the MINIMUM acceptable Water Level above the Reactor Vessel Flange, and the reason for that limit?

_____ A. 20.5 feet provides adequate Iodine absorption following an accident.

_____ B. 20.5 feet provides adequate shielding of personnel during core alterations.

_____ C. 22.0 feet provides adequate Iodine absorption following an accident.

_____ D. 22.0 feet provides adequate shielding of personnel during core alterations.

QUESTION 49 Which malfunction below, were it to occur during a Design Basis Loss of Coolant Accident (DBA-LOCA), would threaten Primary Containment Integrity?

Consider each malfunction, independently, as the ONLY malfunction.

Assume that NO operator action is taken in response to the malfunction.

_____ A. Safety Relief Valve B2104-F013H has a break in its Tailpipe located in the Drywell.

_____ B. SST 65 TRIPS and Emergency Diesel Generator 13 FAILS TO START.

_____ C. ALL Drywell Spray Valves FAIL TO OPEN when attempting to initiate Drywell Spray.

_____ D. ALL Torus to Drywell Vacuum Breaker Check Valves FAIL TO OPEN when Drywell Spray is in operation.

QUESTION 50 Following a Main Steam Line Break, outside of containment, AND an automatic reactor scram, the following conditions exist:

• ALL Control Rods inserted.

• Reactor Pressure is 1115 psig, LOWERING.

• RPV Water Level lowered to 150 inches AND recovered.

Which ONE of the following describes how Decay Heat will be removed from the reactor WITHOUT operator action?

_____ A. Safety Relief Valves will AUTOMATICALLY OPEN.

_____ B. Main Turbine Bypass Valves will AUTOMATICALLY OPEN.

_____ C. Reactor Core Isolation Cooling (RCIC) will AUTOMATICALLY START.

_____ D. Reactor Feedwater Pump Turbines will OPERATE on MINIMUM FLOW.

QUESTION 51 Following a plant transient and a reactor scram, the following conditions exist:

• RHR Pump A is injecting with ONE Pump at 13,500 gpm.

• Div 1 Core Spray is injecting with TWO Pumps at 7,750 gpm.

• Torus Pressure is 5.5 psig.

• Torus Level is -60 inches.

• Torus Temperature is 205°F.

• RPV Pressure is 85 psig (steady).

• RPV Water Level is -10 inches (rising).

The Nuclear Operator in the reactor building calls to report the RHR and Core Spray Pumps are rattling.

To maintain long term injection capability, which of the following is the MAXIMUM injection permissible?

_____ A. Core Spray Div 1 Flow - 7,000 gpm RHR Flow - 11,000 gpm

_____ B. Core Spray Div 1 Flow - 7,000 gpm RHR Flow - 12,500 gpm

_____ C. Core Spray Div 1 Flow - 8,000 gpm RHR Flow - 11,000 gpm

_____ D. Core Spray Div 1 Flow - 8,000 gpm RHR Flow - 12,500 gpm

QUESTION 52 E2101-C001A and C, Div 1 Core Spray Pumps are being operated in Full Flow Test, when a transient occurs resulting in the following plant conditions:

• Drywell Pressure is 1.48 psig.

• RPV Water Level is 30 inches, lowering 1 inch per minute.

• Reactor Pressure is 520 psig, lowering 5 psig per minute.

• ALL High Pressure Injection Systems have failed to inject.

• ALL RHR Pumps will NOT start.

• Core Spray Pumps B and D will NOT start.

Which ONE of the following describes the affect of Core Spray Pump operation on RPV Water Level?

Core Spray Pumps A and C:

_____ A. REMAIN in Full Flow Test, and RPV Water Level continues to LOWER.

_____ B. TRIP and RPV Water Level continues to LOWER because a start signal has NOT been received.

_____ C. REMAIN operating and RPV Water Level continues to LOWER because a permissive condition has NOT been satisfied.

_____ D. REMAIN operating and RPV Water Level will RISE because ALL permissive conditions HAVE been satisfied.

QUESTION 53 The plant is in an emergency condition and the following Primary Containment parameters exist:

• Torus Water Level is 0 inches

• Drywell Temperature is 275°F.

• Drywell Pressure is 4 psig.

• Torus Pressure is 3 psig.

If Drywell Sprays were INITIATED, which ONE of the following will occur?

_____ A. The Torus to Drywell Vacuum Breakers will NOT operate due to low Differential Pressure.

_____ B. Convective Cooling WILL RESULT in Nitrogen being drawn into the Drywell by operation of the Torus to Drywell Vacuum Breakers.

_____ C. The Torus to Drywell Vacuum Breakers capacity WILL BE EXCEEDED and damage to the Primary Containment Vent system will occur.

_____ D. Evaporative Cooling WILL RESULT in Oxygen being drawn in to the Torus by operation of the Reactor Building to Torus Vacuum Breakers.

QUESTION 54 RCIC is being used to control RPV Water Level with its suction aligned to the Torus when a leak in the Torus occurs.

Which ONE of the following will occur FIRST as Torus Level continues to lower?

_____ A. RCIC will TRIP due to Low Suction Pressure.

_____ B. RCIC will TRIP due to Low Cooling Water Flow.

_____ C. RCIC suction will AUTO TRANSFER to the CST due to Low Suction Pressure.

_____ D. RCIC suction will AUTO TRANSFER to the CST due to Low Torus Water Level.

QUESTION 55 While mitigating an ATWS per 29.100.01 Sheet 1A, based on the attached curve, what is the significance of Torus Water Temperature reaching 120°F while Reactor Power is 10% ?

_____ A. If Emergency Depressurization is conducted at this point, the Heat Capacity Limit will NOT be exceeded.

_____ B. If Torus Water Temperature continues to increase AND is being used as the injection source, Reactor Power will LOWER.

_____ C. If Standby Liquid is injected at this point, Hot Shutdown Boron Weight will be injected before the Heat Capacity Limit is reached.

_____ D. If ALL injection to the RPV is Terminated and Prevented at this point, RPV Water Level will remain ABOVE TAF when Reactor Power reaches 3%.

QUESTION 56 Why does 29.100.01 Sheet 5, RADIOACTIVITY RELEASE CONTROL LEG permit the RESTART of isolated HVAC Systems?

While executing the Radioactivity Release Control Leg, restarting HVAC Systems:

_____ A. ensures a POSITIVE pressure is maintained in the Control Room.

_____ B. ensures ACCESSIBILITY is maintained INSIDE the Secondary Containment.

_____ C. provides FILTRATION and ADSORPTION of radioactivity and an elevated release path.

_____ D. ensures ACCESSIBILITY is maintained in buildings OUTSIDE the Secondary Containment.

QUESTION 57 Following a Loss of Offsite Power, a fire is in progress.

RCIC has started WITHOUT an initiation signal.

Which ONE of the following actions is REQUIRED and WHY?

Complete the plant shutdown using:

_____ A. RCIC; because NO operator action is required to achieve injection.

_____ B. Standby Feedwater; because RCIC CANNOT be relied upon as a makeup source and is required to be disabled.

_____ C. HPCI; because RCIC CANNOT be relied upon as a makeup source and is required to remain running as a backup source.

_____ D. HPCI; because RCIC AND Standby Feedwater CANNOT be relied upon as a makeup source and are required to remain running as a backup source.

QUESTION 58 Following a Grid Disturbance, conditions are as follows:

• Generator Power is 1200 Mwe.

• Reactive Power is 360 MVAR (LAG).

• Generator Hydrogen Pressure is 75 psig.

The System Dispatcher has requested additional reactive load support to maintain grid voltage.

Considering the attached Capability Curve, which ONE of the following actions is required?

_____ A. RAISE Recirculation Flow to increase the Reactive Load on the Generator.

_____ B. LOWER Recirculation Flow, because Generator Load limits have been EXCEEDED.

_____ C. MANUALLY RAISE the Voltage Regulator setting to increase the Reactive Load on the Generator.

_____ D. MANUALLY LOWER the Voltage Regulator setting, because Reactive Load limits have been EXCEEDED.

QUESTION 59 The plant is at 100% when a Main Turbine Trip occurred.

Which ONE of the following describes the plant conditions that will CAUSE a Main Turbine Trip AND the BASIS for that trip?

The Main Turbine has tripped due to:

_____ A. TWO of the Narrow Range Level instruments having a level of 214".

This will prevent the erosion of the Main Turbine Blades.

_____ B. the SELECTED Narrow Range Level instrument having a level of 214".

This will prevent the erosion of the Main Turbine Blades.

_____ C. TWO of the Narrow Range Level instruments having a level of 214".

This will prevent the erosion of the Main Steam piping and Main Control Valve seats.

_____ D. the SELECTED Narrow Range Level instrument having a level of 214".

This will prevent the erosion of the Main Steam piping and Main Control Valve seats.

QUESTION 60 With the plant operating at full power, a Nitrogen Regulator failure caused Drywell Pressure to rise to 1.75 psig.

• NO RPS actuation occurred.

• ALL OTHER isolations and actuations occurred.

Which ONE of the following describes the resulting trend of Drywell Temperature?

Drywell Temperature will:

_____ A. RISE due to isolation of EECW.

_____ B. RISE due to continued Nitrogen Addition.

_____ C. RISE due to the Two Speed Drywell Cooling Fans shifting from FAST to OFF.

_____ D. LOWER due Two Speed Drywell Cooling Fans shifting from SLOW to FAST.

QUESTION 61 The plant is operating at full power and the following conditions exist:

• 8D41, Div 1 DW Temperature High alarms.

• 14 DW Cooling Fans are operating.

• T47-R803A, point 16 indicates > 185°F.

• P42-K803, RBCCW to TCV P42-F400 CTRLR, valve position indicates 100%.

• Lake Temperature is 71°F.

The AVERAGE Drywell Temperature has risen from 132°F to 135°F during the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Which ONE of the following actions is appropriate?

_____ A. Operate ALL available Drywell Cooling per 29.100.01, Sheet 2, Primary Containment Control.

_____ B. SHIFT DW Cooling Fans 1, 2, 3, and 4 to LOW speed per 23.415, Drywell Cooling System.

_____ C. PLACE RBCCW Supplemental Cooling in service, per 23.127.01, RBCCW Supplemental Cooling System.

_____ D. Manually INITIATE EECW and EESW Systems per 20.127.01, Loss of Reactor Building Closed Cooling Water System.

QUESTION 62 The reactor is in MODE 2, with Reactor Pressure at 800 psig when the following occurs:

• The operating CRD pump TRIPS.

• 3D10, CRD ACCUMULATOR TROUBLE, alarms for Control Rod 30-27.

• Control Rod 30-27 is at position 48.

What action is REQUIRED in accordance with procedure 20.106.01, CRD Hydraulic System Failure?

_____ A. PLACE the Mode Switch in SHUTDOWN.

_____ B. IMMEDIATELY START the standby CRD pump.

_____ C. Within 20 minutes, CLOSE C1100-F034, CRD Charging Water Header Isolation Valve

_____ D. Within 20 minutes START at least one CRD pump and FULLY INSERT Control Rod 30-27.

QUESTION 63 Which ONE of the following is the reason for having the Main Steam Tunnel High Temperature Isolation?

The Main Steam Tunnel High Temperature Isolation will:

_____ A. LIMIT the escape of radioactivity from the MSL Tunnel to the Reactor Building HVAC system.

_____ B. PREVENT exceeding the Environmental Qualification temperature limits on the MSIV control solenoids.

_____ C. PROTECT the integrity of the Secondary Containment AND ensure the continued operability of safe shutdown equipment.

_____ D. MINIMIZE radioactive releases to the environment AND limit the inventory loss from the reactor under all accident conditions.

QUESTION 64 The plant is operating at 100% power. 16D6, REAC/AUX BLDG FIRST FLOOR HIGH RADN alarms. Area Radiation Monitors associated with this alarm indicate:

• D21-K702 (RB1 RR Airlock) indicates 4 mr/hr.

• D21-K712 (RB1 Inside TIP Room) indicates 420 mr/hr.

• D21-K713 (RB1 Outside TIP Room) indicates 85 mr/hr.

• D21-K732 (AB1 Near Blowout Panel) indicates 3 mr/hr.

• D21-K733 (RB1 South Airlock) indicates 2 mr/hr.

• D21-K745 (Drywell) indicates 12 mr/hr.

Which ONE of the following plant conditions would be consistent with these indications?

_____ A. A steam leak has developed in RCIC piping.

_____ B. Spent Fuel Handling operations are in progress.

_____ C. Traversing In-core Probe movement is in progress.

_____ D. SRM detectors are being withdrawn for post maintenance testing.

QUESTION 65 29.100.01, Sheet 5, Secondary Containment and Rad Release, directs operating available sump pumps whenever Secondary Containment area or sump levels exceed their Max Normal Operating levels.

What is the BASIS for this action?

This action is BASED on:

_____ A. MINIMIZING the spread of contamination within the Secondary Containment.

_____ B. MAINTAINING water levels below the point at which equipment required for safe shutdown will fail.

_____ C. PREVENTING the uncontrolled release of liquid radioactive effluents from the Secondary Containment.

_____ D. CONTAINING leakage from a primary system within systems design for storage of radioactive liquids.

QUESTION 66 During a Refuel Outage, with the Reactor Mode Switch in REFUEL, the following conditions exist:

• 3D113, CONTROL ROD WITHDRAWAL BLOCK alarms.

• ONE Control Rod is SELECTED.

• ALL Control Rods are INSERTED to position 00.

Which ONE of the following conditions caused the alarm?

The Refuel Bridge is:

_____ A. over the core with the Grapple FULL UP and the Trolley Hoist is NOT LOADED.

_____ B. over the core with the Grapple FULL DOWN and the Trolley Hoist is NOT LOADED.

_____ C. NOT over the core with the Grapple FULL UP and the Trolley Hoist is LOADED.

_____ D. NOT over the core with the Grapple FULL DOWN and the Trolley Hoist is LOADED.

QUESTION 67 After a ONE WEEK VACATION, a Nuclear Station Operator is scheduled to work the following schedule:

Sunday Monday Tuesday Wednesday Thursday Friday Saturday OFF 12 hrs 12 hrs 12 hrs 12 hrs 8 hrs OFF

Per MGA 17, Working Hour Limits, which ONE of the following is the MAXIMUM ADDITIONAL hours this person can be scheduled to work WITHOUT exceeding any administrative limits?

_____ A. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> on Tuesday

_____ B. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> on Thursday

_____ C. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> on Saturday

_____ D. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on Sunday

QUESTION 68 The plant is operating at 50% power. What is the MAXIMUM amount of TOTAL Reactor Coolant System Leakage allowed for continued plant operation?

_____ A. 2 gpm

_____ B. 5 gpm

_____ C. 25 gpm

_____ D. 50 gpm

QUESTION 69 Which ONE of the following is required when a visible break CANNOT be used to disconnect a piece of equipment from its power supply?

_____ A. Independent verification of the danger tag.

_____ B. An approved grounding device installed on the load side.

_____ C. A safety observer is stationed for all work performed on the equipment.

_____ D. An approved blocking device and a method for determining that power is removed.

QUESTION 70 With the plant operating at 80% power, at 0800 on February 8, 2008, EDG 11 is discovered INOPERABLE.

Which ONE of the following describes LATEST TIME that SR 3.8.1.1 must be completed WITHOUT entering into a condition which requires a unit shutdown?

_____ A. 0815 on February 8, 2008

_____ B. 0850 on February 8, 2008

_____ C. 0905 on February 8, 2008

_____ D. 0915 on February 9, 2008 QUESTION 71 With core alterations in progress, a fuel assembly contacts the core top guide, resulting in 16D1, RB REFUELING AREA FIFTH FLOOR HIGH RADN alarm.

Indications are as follows:

• AREA Radiation Monitor 15, RB5 Spent Fuel Pool AREA Radiation Monitor (ARM) indicates 25 mr/hr, rising.

• AREA Radiation Monitor 17, RB5 Refuel Floor Lo Range AREA Radiation Monitor (ARM) indicates 30 mr/hr, rising.

Which ONE of the following is the Control Room action required?

_____ A. Ensure automatic isolations have occurred.

_____ B. Alert personnel by using the Plant Area alarm.

_____ C. Ensure that Standby Gas Treatment is operating.

_____ D. Ensure that Control Room HVAC is operating in the filtered mode QUESTION 72 Which ONE of the following conditions will cause the Division 1 AXM to automatically shift from STANDBY to OPERATE?

_____ A. Automatic start of Division 1 SGTS.

_____ B. Automatic shift of CCHVAC to the Recirc Mode.

_____ C. High Radiation Trip of Div 1 or 2 Containment High Range Radiation Monitors.

_____ D. High Radiation Alarm on the Div 1 SGTS SPING Medium Range Noble Gas Channel.

QUESTION 73 With the plant operating at 100% power, the following conditions exist:

• Drywell Pressure increased to 1.75 psig.

• NO RPS actuations occurred.

• NO Control Rod motion occurred.

With these conditions, which ONE of the following actions is REQUIRED?

It is IMMEDIATELY required to:

_____ A. ENTER 29.100.01 Sheet 1, RPV Control ONLY, THEN place the Reactor Mode Switch in SHUTDOWN.

_____ B. PLACE the Reactor Mode Switch in SHUTDOWN, THEN enter 29.100.01 Sheet 2, Primary Containment Control ONLY.

_____ C. INITIATE Standby Liquid Control THEN enter 29.100.01 Sheet 1A, ATWS Control AND 29.100.01 Sheet 2, Primary Containment Control.

_____ D. PLACE the Reactor Mode Switch in SHUTDOWN AND enter 29.100.01 Sheet 1, RPV Control AND 29.100.01 Sheet 2, Primary Containment Control.

QUESTION 74 Which ONE of the following is an IMMEDIATE Action for a CONFIRMED fire in accordance with 20.000.22, Plant Fires?

_____ A. Identify the type or class of fire.

_____ B. Announce the fire alarm over the Hi-Com system.

_____ C. Dispatch an operator to verify the magnitude and location of the fire.

_____ D. Establish communications between the Control Room and the Fire Brigade.

QUESTION 75 An ALERT Emergency Action Level has been declared. The Technical Support Center and the Emergency Operations Facility are NOT activated.

Per EP-290, Emergency Notifications, which ONE of the following communications methods is used to make an INITIAL notification to the US Nuclear Regulator Commission?

INITIAL notification to the US Nuclear Regulator Commission is made by:

_____ A. contacting the NRC Resident Inspector.

_____ B. using the HPN (Health Physics Network) telephone system.

_____ C. using the ENS (Emergency Notification System) telephone.

_____ D. using the ECOS (Emergency Call Out System) telephone system.

SRO Tier K/A Number Statement IR Origin Source Question 76 1 295004 2.2.25 4.2 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents F 1 Technical Specification B3.8.5 Rev 31 Partial or Total Loss of DC Pwr Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

QUESTION 76 With the plant in MODE 4, Cold Shutdown, which ONE of the following describes the MINIMUM DC Sources REQUIRED OPERABLE and the reason for that requirement?

_____ A. ONE subsystem consisting of TWO 130 VDC Batteries AND TWO Chargers. This will provide power necessary to mitigate a Design Basis Loss of Coolant Accident.

_____ B. ONE subsystem consisting of TWO 130 VDC Batteries AND TWO Chargers. This will provide power necessary to mitigate an inadvertent Reactor Vessel Draindown.

_____ C. TWO subsystems consisting of TWO 130 VDC Batteries AND TWO Chargers. This will provide power necessary to mitigate a Design Basis Loss of Coolant Accident.

_____ D. TWO subsystems consisting of TWO 130 VDC Batteries AND TWO Chargers. This will provide power necessary to mitigate an inadvertent Reactor Vessel Draindown.

Correct Answer: B LCO 3.8.5 requires ONE 130 VDC subsystem OPERABLE in MODE 4.

This is based on mitigating fuel handling accidents and reactor vessel Draindown postulated to occur during COLD SHUTDOWN conditions.

Plausible Distractors:

A is plausible; DBA LOCA is the analyzed failure for MODES 1, 2, and 3.

C is plausible; In MODE 4, TWO subsystems OPERABLE is NOT the MINIMUM. DBA LOCA is the analyzed failure for MODES 1, 2, and 3.

D is plausible; In MODE 4, TWO subsystems OPERABLE is NOT the MINIMUM.

Objective Link: LP-OP-315-0164-C012

SRO Tier K/A Number Statement IR Origin Source Question 77 1 295016 2.4.4 4.7 M Fermi-2 Bank EQ-OP-213-0427-000-0001-002 LOK Grp 10 CFR 55.43(b)5 LOD (1-5) Reference Documents H 1 20.000.18 Rev 38 Control Room Abandonment Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

QUESTION 77 The plant was operating at full power when a fire in the Cable Spreading Room (Zone 11) occurred. A Loss of Offsite Power has occurred resulting in an EOP entry condition due to RPV Water Level. This transient has been complicated by the spurious operation of numerous components and smoke in the control room.

Which ONE of the following procedures contains measures which will MITIGATE the SPURIOUS OPERATION of components?

_____ A. 29.100.01, Sheet 1, RPV CONTROL

_____ B. 20.300.OFFSITE, LOSS OF OFFSITE POWER

_____ C. 20.000.18, SHUTDOWN FROM THE DEDICATED SHUTDOWN PANEL

_____ D. 20.000.19, SHUTDOWN FROM OUTSIDE THE CONTROL ROOM Correct Answer: C 20.000.18, SHUTDOWN FROM THE DEDICATED SHUTDOWN PANEL will mitigate the effects of spurious operation of components?

Plausible Distractors:

A is plausible; will NOT mitigate spurious operation, but would be appropriate in the absence of spurious operation.

B is plausible; will NOT mitigate spurious operation, but would be appropriate in the absence of spurious operation.

D is plausible; would be appropriate if Control Room evacuation were required and in the absence of spurious operation.

Objective Link: LP-OP-315-0199-A001

SRO Tier K/A Number Statement IR Origin Source Question 78 1 295024 EA2.01 4.4 B Fermi-2 Bank EQ-OP-202-0121-000-A002-007 LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 29.100.01 Sheet 2 Rev 9 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Drywell Pressure QUESTION 78 During an accident condition after Emergency RPV Depressurization, the following conditions exist:

• RPV Water Level is 20 inches and lowering.

• RPV Pressure is 70 psig, lowering.

• Drywell Temperature is 250°F, rising.

• Drywell Pressure is 42 psig, rising.

• Torus Pressure is 42.5 psig, rising.

• Primary Containment Water Level is 580 ft, RISING.

• Torus Venting is in progress.

Which ONE of the following actions is required?

_____ A. VENT the Drywell.

_____ B. RESTART Drywell Coolers.

_____ C. PREVENT Core Spray and LPCI.

_____ D. ENTER the Reactor Flooding Procedure.

Correct Answer: A With Drywell Pressure approaching PCPL, it is required to vent the Drywell IRRESPECTIVE of offsite release rate limits. Torus Venting is not maintaining Containment below PCPL.

Plausible Distractors:

B is plausible; CANNOT restart Drywell Coolers >242°F.

C is plausible; RPV Water Level is very LOW, preventing Core Spray and LPCI threatens adequate core cooling.

D is plausible; would be true if RPV Water Level cannot be determined. With RPV Pressure at 70 psig, Drywell Temperature at 250°F is BELOW saturation temperature.

Objective Link: LP-OP-802-3004-0002

SRO Tier K/A Number Statement IR Origin Source Question 79 1 295025 EA2.06 3.8 B 2003 LaSalle NRC Exam LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 29.100.01 Sheet 1A ATWS RPV Control Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor water level QUESTION 79 An ATWS is in progress following a condenser boot rupture, with plant conditions as follows:

• RPV Water Level is 150 inches.

• APRM DOWNSCALE Lights are NOT lit.

• Suppression Pool Temperature is 118°F.

• Low-Low Set is controlling reactor pressure at 1020 psig.

If the above parameters remain CONSTANT, which ONE of the following is the HIGHEST RPV Water Level that may be MAINTAINED?

_____ A. +214 inches

_____ B. +114 inches

_____ C. 0 inches

_____ D. -25 (minus 25) inches Correct Answer: B with Reactor Power > 3% (APRM DOWNSCALE Lights NOT Lit) and RPV Water Level above 114 inches, it is required to Terminate and Prevent Injection until RPV Water Level lowers to 114 inches. High Reactor Pressure condition is met by having Low Low Set controlling Reactor Pressure.

Plausible Distractors:

A is plausible; +214 inches is the HIGHEST RPV Water Level allowed if power were BELOW 3%. (APRM DOWNSCALE Lights LIT)

C is plausible; 0 inches is the end point of the Terminate and Prevent Injection statement IF RPV Water Level was initially between +114 and 0 inches.

D is plausible; is the LOWEST RPV Water Level allowed.

Objective Link: LP-OP-802-3003-0010

SRO Tier K/A Number Statement IR Origin Source Question 80 1 295030 EA2.01 4.2 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 29.100.01 Sheet 2 Rev 9 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL : Suppression pool level QUESTION 80 Following a transient, the following conditions exist:

• ONLY HPCI and RCIC are injecting.

• RPV Water Level is 16 inches, LOWERING 2 inches per minute.

• Reactor Pressure is 1000 psig.

• ALL Control Rods are fully inserted.

• Torus Water Level is -38 inches, LOWERING 2 inches per minute.

Which ONE of the following actions should be ordered FIRST?

_____ A. DEPRESSURIZE the reactor by opening Turbine Bypass Valves.

_____ B. DEPRESSURIZE the reactor by opening FIVE Safety Relief Valves.

_____ C. SHUTDOWN the HPCI Turbine to prevent direct pressurization of the Torus.

_____ D. PREVENT Core Spray AND LPCI Pump Injection, because injection is NOT needed.

Correct Answer: B Emergency Depressurization is required with Torus Water Level < 38 inches.

Plausible Distractors:

A is plausible; would be true prior to reaching ED criteria.

C is plausible; would be true if Torus Water level were approaching -68 inches. RPV Water Level is very low.

D is plausible; would be true if RPV Water Level were substantially higher. RPV Water Level is very low and injection systems (HPCI and RCIC) will be lost after Emergency Depressurization.

With these conditions, it is NOT appropriate to secure CS and LPCI.

Objective Link: LP-OP-802-3004-0001

SRO Tier K/A Number Statement IR Origin Source Question 81 1 295038 EA2.03 4.3 B 2001 Fermi-2 NRC Exam LOK Grp 10 CFR 55.43(b)5 LOD (1-5) Reference Documents H 1 29.100.01 Sheet 5 Rev 7 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE : Radiation levels QUESTION 81 While operating the reactor in MODE 1, a restriction of cooling water flow through a fuel bundle causes fuel clad overheating and fission product release into the reactor coolant. The following plant conditions exist:

• Reactor Power is 18%.

• Reactor Pressure is 940 psig.

• RPV Water Level is 100 inches.

• Main Steam Line B Inboard AND Outboard MSIVs have failed OPEN.

• Main Turbine is TRIPPED.

• Site Boundary Release corresponds to 4.90 REM to an Adults Thyroid AND is RISING.

Given these conditions, which ONE of the following actions is REQUIRED?

_____ A. Use the Safety Relief Valves to perform a reactor cool down at LESS THAN a 90°F/hr rate.

_____ B. Use HPCI and RCIC in the Test Mode to perform a reactor cool down at LESS THAN a 90°F/hr rate.

_____ C. Use the Safety Relief Valves to perform a reactor cool down at GREATER THAN a 90°F/hr rate.

_____ D. Use Main Turbine Bypass Valves to perform a reactor cool down at GREATER THAN a 90°F/hr rate.

Correct Answer: C With Site Boundary Dose approaching the EPA PAG, or General Emergency EAL, it is required to perform an Emergency Depressurization. ED utilizes Safety Relief Valves and GREATER THAN a 90°F/hr rate.

Plausible Distractors:

A is plausible; >90°F/hr is expected and permitted during an Emergency Depressurization to quickly put the reactor in a low energy state and reduce the release rate.

B is plausible; >90°F/hr is expected and permitted during an Emergency Depressurization to quickly put the reactor in a low energy state and reduce the release rate.

D is plausible; using Turbine Bypass Valves is NOT permitted after Emergency Depressurization is required.

Objective Link: LP-OP-802-3005-0009 SRO Tier K/A Number Statement IR Origin Source Question 82 1 600000 2.4.41 4.6 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 EP-101 Enclosure A TAB H Rev 30 Plant Fire On Site - Knowledge of the emergency action level thresholds and classifications.

QUESTION 82 The plant is operating at full power.

At time 1200, a fire was reported in Bus 64B, position B10, the power supply breaker for E2101-C001A Division 1 Core Spray Pump A. At time 1215, the Fire Brigade has requested that the Bus be deenergized prior to attempting extinguishment.

Which ONE of the following Emergency Action Levels (EAL) is required, and what is the criterion for that EAL?

It is required to declare an:

_____ A. UNUSUAL EVENT, due a fire inside the PROTECTED AREA.

_____ B. UNUSUAL EVENT, due to Loss of Offsite Power.

_____ C. ALERT, due to Loss of Offsite and Onsite AC Power.

_____ D. ALERT, due to a fire involving SAFE SHUTDOWN EQUIPMENT.

Correct Answer: D ALERT EAL HA2 is required due to a fire involving SAFE SHUTDOWN EQUIPMENT. Candidate must know loss of one bus does not indicate a Loss of Offsite Power, and that Bus 64B is a power source for Safe Shutdown equipment Plausible Distractors:

A is plausible; fires on site for 15 minutes require HU 2 Unusual Event. This has been exceeded by HA2.

B is plausible; SU1 is not met because SST 64 and 65 are energized.

C is plausible; SA1 is only applicable in Mode 4 and 5.

Objective Link: LP-ER-832-0001-0004

SRO Tier K/A Number Statement IR Origin Source Question 83 1 295008 AA2.05 3.1 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 2 23.107 Rev 105 Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL : Swell QUESTION 83 The plant was operating at full power, when the following occurred:

• BOTH Feedwater Pumps TRIPPED.

• The reactor automatically scrammed.

• ONLY ONE Control Rod is at position 48.

• ALL OTHER Control Rods are FULLY INSERTED.

• HPCI initiation RAISED RPV Water Level from 110 inches.

• HPCI was MANUALLY TRIPPED as RPV Water Level reached 210 inches.

Plant conditions are currently:

• Reactor pressure 700 psig, rising at 10 psig per minute.

• MSIVs are OPEN.

• The OPERATING CRD Pump TRIPPED.

What is the expected RPV Water Level response over the next TEN MINUTES, and what action will be required to be directed?

Over the next TEN MINUTES, RPV Water Level will . . .

_____ A. RISE due to SWELL. It is required to direct operators to allow steam off to lower RPV Water Level BELOW 214 inches.

_____ B. LOWER due to SHRINK . It is required to direct operators to use HPCI to maintain RPV Water Level ABOVE 173.4 inches.

_____ C. LOWER due to SHRINK. It is required to direct operators to use ONLY RCIC to maintain RPV Water Level ABOVE 0 inches.

_____ D. RISE due to SWELL. It is required to direct operators to TERMINATE AND PREVENT Injection Systems to lower RPV Water Level BELOW 114 inches.

Correct Answer: A HPCI injected (100 inches x 200 gal per inch=) 20,000 gallons of cold CST water. As this water is heated, SWELL occurs. It is required to maintain RPV Water Level below Level 8 (214 inches). The SRO is required to direct operators to allow steaming to lower RPV Water Level. Shrink cannot occur because heatup and pressurization of saturated system is in progress with NO steam voids. ALL SRVs and TBVs are shut for the next ten minutes because Reactor Pressure will be below 800 psig.

Plausible Distractors:

B is plausible; identifies misconception about shrink and swell.

C is plausible; identifies misconception about shrink and swell. 0 inches is the MINIMUM for the ATWS RPV Water Level Control Band. ATWS is plausible because ONE Control Rod did not fully insert.

D is plausible; 114 inches is the MAXIMUM for the ATWS RPV Water Level Control Band.

ATWS is plausible because ONE Control Rod did not fully insert.

Objective Link: None

SRO Tier K/A Number Statement IR Origin Source Question 84 1 295012 2.2.22 4.7 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 2 TS Basis B3.6.1.5 Rev 0 High Drywell Temperature - Knowledge of limiting conditions for operations and safety limits.

QUESTION 84 With the plant operating at full power, the following conditions exist:

• Drywell Temperature is 155°F.

• Drywell Pressure is 0.70 psig.

• Torus Water Level is (-1) inch.

• Torus Water Temperature is 94°F.

Which ONE of the following actions is required with these conditions and why?

_____ A. RESTORE Drywell Temperature below 145°F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to preserve the function of RPV Water Level Instrumentation.

_____ B. RAISE Torus Water Level to zero inches within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to preserve to preserve Net Positive Suction Head to ECCS Pumps.

_____ C. RAISE Torus Water Level to zero inches within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to preserve the pressure suppression function of the Primary Containment.

_____ D. RESTORE Drywell Temperature below 145°F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to preserve the initial conditions assumed in the Loss of Coolant Accident Analysis.

Correct Answer: D Drywell Average Air Temperature exceeds 145°F, which is listed in LCO 3.6.1.5 Plausible Distractors:

A is plausible; RPV Water Level Instruments can be threatened by High Drywell Temperature in Emergency Conditions.

B is plausible; Torus Water Level is low, but within the LCO. Low Torus Water Level can threaten ECCS Pumps from cavitation in Emergency Conditions.

C is plausible; Torus Water Level is low, but within the LCO. Low Torus Water Level can threaten the Suppression function in Emergency Conditions.

Objective Link: None

SRO Tier K/A Number Statement IR Origin Source Question 85 1 295033 EA2.02 3.2 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents F 2 29.100.01 Sheet 5 Rev 7 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS : Equipment operability QUESTION 85 During the execution of 29.100.01 Sheet 5, Secondary Containment and Radiation Release, the operability of equipment required to perform a safe shutdown is assured by which ONE of the following?

Emergency Depressurization is REQUIRED when the:

_____ A. Radiation Level in any ONE AREA exceeds the MAX NORMAL value.

_____ B. Radiation Levels in MORE THAN ONE AREA exceed the MAX SAFE value.

_____ C. Radiation Levels in MORE THAN ONE AREA exceed the MAX NORMAL value.

_____ D. Radiation Level in any ONE AREA exceeds the MAX SAFE value and Water Level exceeds a MAX SAFE Water Level in the SAME AREA.

Correct Answer: B Emergency Depressurization is required when the Radiation Levels in MORE THAN ONE AREA exceed the MAX SAFE value.

Plausible Distractors:

A is plausible; and is an entry condition for 29.100.01 Sheet 5.

C is plausible; does NOT require Emergency Depressurization.

D is plausible; does NOT require Emergency Depressurization.

Objective Link: LP-OP-802-3005-0009

SRO Tier K/A Number Statement IR Origin Source Question 86 2 223002 2.2.36 4.2 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 Technical Specification LCO 3.6.1.3 Amendment 134 and B3.6.1.3 rev 0 PCIS/Nuclear Steam Supply Shutoff - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

QUESTION 86 The plant is in MODE 2, STARTUP, following a Refueling Outage. Engineering has determined that ALL the MSIVs have had unqualified valve control manifolds installed during outage maintenance which will cause LONGER stroke times over the cycle.

Which ONE of the following is the MINIMUM Required Action and the reason for that action, according to Technical Specifications?

_____ A. NO ACTIONS are required, because Primary Containment Isolation capability is NOT REQUIRED OPERABLE in MODE 2.

_____ B. It is REQUIRED to SHUT ONLY ONE MSIV in each Main Steam Line.

The basis for this action is to limit the severity of the MAXIMUM Reactor Pressure spike following a spurious MSIV closure at power.

_____ C. It is REQUIRED to SHUT ONLY ONE MSIV in each Main Steam Line.

The basis for this action is to limit the MAXIMUM Radiological Release following a Design Basis Accident.

_____ D. It is REQUIRED to SHUT BOTH MSIVs in all Main Steam Lines.

The basis for this action is to limit the MAXIMUM Radiological Release following a Design Basis Accident.

Correct Answer: C It is REQUIRED to SHUT ONLY ONE MSIV in each Main Steam Line.

The basis for this action is to limit the MAXIMUM Radiological Release following a Design Basis Accident.

Plausible Distractors:

A is plausible; would be true for MODE 4, COLD SHUTDOWN.

B is plausible; would be true if the Maintenance error resulted in SHORTER stroke times over the cycle.

D is plausible; is NOT the MINIMUM Required Action.

Objective Link: LP-OP-804-0001-0012

SRO Tier K/A Number Statement IR Origin Source Question 87 2 218000 A2.06 4.3 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 29.100.01 Sheet 1 Rev 11 Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS initiation signals present QUESTION 87 Following a Loss of Offsite Power, the following conditions occur at the listed time:

• 12:00 Reactor Scram occurred, all Control Rods are inserted.

• 12:01 ONLY EDG 14 has started and loaded.

• 12:05 Drywell Pressure is 1.0 psig and stable.

• 12:10 RPV Water Level is 64 inches, lowering 4 inches per minute.

Given these conditions; (1) Which ONE of the following describes the response of the Automatic Depressurization System (ADS)? AND (2) What operator actions which should be ordered?

_____ A. (1) ADS will OPEN Safety Relief Valves at 12:20.

(2) It is required to INHIBIT ADS prior to automatic actuation and MANUALLY Emergency Depressurize the reactor at a specified RPV Water Level.

_____ B. (1) ADS will OPEN Safety Relief Valves at 12:27.

(2) It is required to INHIBIT ADS prior to automatic actuation and MANUALLY Emergency Depressurize the reactor at a specified RPV Water Level.

_____ C. (1) ADS will OPEN Safety Relief Valves at 12:20.

(2) It is required to VERIFY ADS automatically actuates and MAXIMIZE Injection with Low Pressure ECCS Pumps and restore RPV Water Level to a specified Water Level Band.

_____ D. (1) ADS will OPEN Safety Relief Valves at 12:27.

(2) It is required to VERIFY ADS automatically actuates and MAXIMIZE Injection with Low Pressure ECCS Pumps and restore RPV Water Level to a specified Water Level Band.

Correct Answer: B With NO High Drywell Pressure signal present, L1 (31.8 inches) will cause ADS Timer to initiate in 7 minutes. The ADS Timer lasts 105 seconds or 2 minutes. A total of 9 minutes later, SRVs will OPEN. L1 will be reached in 8 minutes. 8 minutes + 9 minutes = 17 minutes 12:10 + 17 minutes = 12:27 29.100.01 Sheet 1 requires ADS INHIBITED when L1 is reached.

Plausible Distractors:

A is plausible; time would be true with Drywell Pressure above 1.68 psig.

C is plausible; time would be true with Drywell Pressure above 1.68 psig, it is required to verify ALL other ECCS actuations and MAXIMIZE Injection.

D is plausible; it is required to verify ALL other ECCS actuations and MAXIMIZE Injection.

Objective Link: LP-OP-315-0142-OBJ C

SRO Tier K/A Number Statement IR Origin Source Question 88 2 215003 2.4.31 4.1 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 LCO 3.3.1.1 Amendment 134 Intermediate Range Monitors - Emergency Procedures / Plan: Knowledge of annunciator alarms, indications, or response procedures.

QUESTION 88 A reactor startup is in progress with Intermediate Range Monitor (IRM) Channel A is INOPERABLE and BYPASSED, when the following occurs:

• IRM Channel D indicates upscale at 125/125, irrespective of Range Switch position.

• IRM Channels B, C, E, F, G, and H indicate 32/40 on Range 7.

• ALL Average Power Range Monitors (APRMs) are DOWNSCALE.

Which ONE of the following actions should be directed?

_____ A. PLACE IRM Channel D in a TRIPPED condition and continue the Reactor Startup.

_____ B. SHUTDOWN per GOP 22.000.04, Plant Shutdown from 25% power; because REQUIRED Intermediate Range Monitors are INOPERABLE.

_____ C. BYPASS the IRM Channel D ROD BLOCK using the joystick per 23.603, Intermediate Range Monitors; RESET the Half Scram, and CONTINUE the Reactor Startup.

_____ D. BYPASS IRM Channel D ROD BLOCK by placing the Reactor Mode Switch in RUN per GOP 22.000.02, Plant Startup from 25% power; RESET the Half Scram, and CONTINUE the Reactor Startup.

Correct Answer: C 3 IRMs per RPS Trip System are OPERABLE. Directed actions include BYPASSING the IRM Channel D ROD BLOCK using the joystick per 23.603, Intermediate Range Monitors; RESETTING the Half Scram, and continuing the Reactor Startup.

Plausible Distractors:

A is plausible; IRM is in a TRIPPED condition, needs to be BYPASSED to continue startup.

B is plausible; would be true with an additional IRM in either TRIP System INOPERABLE.

D is plausible; Reactor Power is too low to place the Reactor Mode Switch in RUN.

Objective Link: None

SRO Tier K/A Number Statement IR Origin Source Question 89 2 215005 A2.03 3.8 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 LCO 3.3.1.1. Action A.1 Amendment 139 Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions : Inoperative trip (all causes)

QUESTION 89 The plant is operating at 50% power with Average Power Range Monitor (APRM)

Channel 1 INOPERABLE and BYPASSED. APRM Channel 3 has multiple in-service LPRM inputs as follows:

A Level LPRMs .5 B Level LPRMs .5 C Level LPRMs .4 D Level LPRMs .6 3D118, LPRM DOWNSCALE, alarms.

The P603 selects the LPRM Display on the APRM 3 ODA and determines that LPRM 08-17C is DOWNSCALE. The STA recommends bypassing LPRM 08-17C.

Which ONE of the following describes the affect of BYPASSING LPRM 08-17C on APRM Channel 3, and what action is required?

When LPRM 08-17C is BYPASSED, APRM Channel 3 will be INOPERABLE because there are ONLY:

_____ A. 19 LPRMs providing input. Technical Specifications are satisfied by the TWO remaining APRMs.

_____ B. 3 LPRMs providing input at the C Level. Technical Specifications are satisfied by the TWO remaining APRMs.

_____ C. 19 LPRMs providing input. Technical Specifications will be satisfied by placing APRM Channel 3 in TRIPPED condition.

_____ D. 3 LPRMs providing input at the C Level. Technical Specifications will be satisfied by placing APRM Channel 3 in TRIPPED condition.

Correct Answer: C With 19 TOTAL LPRMs providing input, APRM Channel 3 is made INOPERABLE. 3 is the REQUIRED Number of APRM Channels, and with 1 REQUIRED APRM INOPERABLE, a channel must be placed in TRIP.

Plausible Distractors:

A is plausible; misconception that MINIMUM Required Channels are 2.

B is plausible; misconception that 3 LPRMs per axial location renders an APRM INOPERABLE and misconception that MINIMUM Required Channels are 2.

INOPERABLE.

D is plausible; misconception that 3 LPRMs per axial location renders an APRM INOPERABLE.

Objective Link: None

SRO Tier K/A Number Statement IR Origin Source Question 90 2 262001 2.4.8 4.5 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 20.307.01 Rev 21 AC Electrical Distribution Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

QUESTION 90 Following a Loss of Coolant Accident with an electrical plant malfunction, plant conditions are as follows:

• RPV Water Level is 35 inches, LOWERING.

• Reactor Pressure is 250 psig, LOWERING.

• 345 kV Mat Power Indicating Lights are OFF.

• EDG 13 is LOADED carrying Bus 65E.

• Bus 65F and Bus 14ED Power Indicating Lights are OFF.

• Bus 72F and Bus 72ED Power Indicating Lights are OFF.

• 65F-F6 Breaker is TRIPPED.

• 65F-F8 Breaker is CLOSED.

Which ONE of the following lists the electrical procedure which SHOULD be executed to provide MAXIMUM Low Pressure ECCS Injection to support 29.100.01, Sheet 1, RPV Control actions?

_____ A. 20.300.65F, Loss of Bus 65F, due to Bus 65F being LOCKED OUT.

_____ B. 20.300.72F, Loss of Bus 72F, due to Bus 72F being LOCKED OUT.

_____ C. 20.307.01, Emergency Diesel Generator Failure, due to EDG 14 failing to start.

_____ D. 20.300.SBO, Loss of Offsite and Onsite Power, due to a combination of electrical malfunctions.

Correct Answer: C A loss of 345kv power is indicated, resulting in loss of ESF Div 2 AC Power. An EDG 14 START FAILURE has resulted in Bus 65F being deenergized.

20.307.01 may restore power to a Core Spray Pump and an RHR Pump to allow Maximum ECCS Injection.

Plausible Distractors:

A is plausible; 65F LOCKOUT condition is excluded by 65F-F8 Breaker being CLOSED.

B is plausible; 72F LOCKOUT condition is excluded by 65F-F6 Breaker being TRIPPED.

D is plausible; Station Blackout is excluded by Div 1 ESF Buses energized and EDG 13 LOADED.

Objective Link: LP-OP-802-2001- OBJ A

SRO Tier K/A Number Statement IR Origin Source Question 91 2 219000 2.4.11 4.2 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 2 LCO 3.6.2.1 Amendment 134 RHR/LPCI: Torus/Pool Cooling Mode - Knowledge of abnormal condition procedures.

QUESTION 91 With the plant operating at full power, the following conditions exist:

• ONE Safety Relief Valve OPENED.

• ALL available Torus Cooling was initiated.

Which ONE of the following describes how Torus Cooling operation will affect Torus Temperature?

Operating ALL available Torus Cooling with ONE FULLY OPEN Safety Relief Valve will:

_____ A. LOWER Torus Water Temperature below the LOWEST Technical Specification LCO value.

_____ B. MAINTAIN Torus Water Temperature at a CONSTANT temperature until the Safety Relief Valve is successfully CLOSED.

_____ C. NOT MAINTAIN Torus Water Temperature BELOW the Technical Specification LCO value which requires a Reactor Shutdown.

_____ D. MAINTAIN Torus Water Temperature BELOW the Technical Specification LCO value applicable when testing which adds heat to the Torus is in progress.

Correct Answer: C ONE SRV open substantially exceeds the capacity of all available Torus Cooling. Torus Water Temperature will exceed 110°F, which Technical Specification LCO 3.6.2.1 Condition D requires a reactor shutdown.

BOTH RHR HXs 41.6 x 106 BTU/hr x 2 = 93.2 x 106 BTU/hr ONE SRV = 6% power 3430 Mwt x 6% = 205.8 Mwt x 3.413 x 106 BTU/ Mw-hr =

702.4 x 106 BTU/hr Plausible Distractors:

A is plausible; would be true if RHR capacity exceeded SRV heat addition.

B is plausible; would be true if RHR capacity matched SRV heat addition.

D is plausible; TS LCO 3.6.2.1C specifies a higher allowable Torus Temperature when testing is in progress.

Objective Link: LP-OP-315-0141-C013

SRO Tier K/A Number Statement IR Origin Source Question 92 2 223001 2.2.42 4.6 B Fermi-2 Bank EQ-OP-315-0116-000-0003-016 LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 2 LCO 3.4.4 Amendment 134 Primary Containment and Auxiliaries Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

QUESTION 92 The plant is operating at full power.

The following Drywell Floor and Equipment Drain Sump Effluent Integrator readings (total gallons pumped) have been noted for the past 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

TIME Floor Drain Equipment Drain Leak Rate Leak Rate Integrator Integrator Floor (gpm) Equipment (gpm) 0000 89321 27861 2.3 16.4 0800 90543 35805 2.54 16.55 1600 92079 44181 3.2 17.45 0000 94383 52821 4.8 18.0 With these conditions, which ONE of the following is correct?

_____ A. NO Drywell Leakage limit has been exceeded.

_____ B. TOTAL LEAKAGE has exceeded the leakage limit.

_____ C. IDENTIFIED LEAKAGE has exceeded the leakage limit.

_____ D. UNIDENTIFIED LEAKAGE INCREASE has exceeded limits within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

Correct Answer: D a 2 gpm increase in UNIDENTIFIED DRYWELL LEAKAGE within a twenty four hour period has been exceeded. 2.3 gpm increased to 4.8 gpm in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Plausible Distractors:

A is plausible; would be true in MODE 2 STARTUP, because the 2 gpm increase in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is NA in MODE 2.

B is plausible; would be true if 5.0 gpm Equipment Leakage were exceeded (4.8 gpm max).

C is plausible; would be true if 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Total Leakage exceed 25 gpm (20.46 gpm actual).

Objective Link: LP-OP-315-0116-C013

SRO Tier K/A Number Statement IR Origin Source Question 93 2 271000 A2.17 3.1 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 2 ARP 3D12 Rev 12 Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Reactor power changes QUESTION 93 During a reactor startup, Control Rod withdrawal is in progress at 20% power.

The following indications are received:

• 3D8, DIV I/II OFF GAS RADN MONITOR UPSCALE alarms.

• 3D12, DIV I/II OFF GAS RADN MONITOR HIGH-HIGH alarms.

• D11-K601A and B Off Gas Radiation Monitors indicate 1200 mr/hr, RISING.

• At H21-P275A, Hydrogen Analyzer Panel, BOTH Channels indicate 1.2% H2.

• Main Condenser Vacuum is 1.0 psia.

• Off Gas Flow is 10 cfm.

Which ONE of the following CAUSED these indications, and what action is REQUIRED?

These indications are CAUSED by:

_____ A. a fuel cladding failure. It is required to enter 20.000.07, Fuel Cladding Failure.

_____ B. increased Main Condenser air inleakage. It is required to enter 20.125.01, Loss of Main Condenser Vacuum.

_____ C. an Off Gas Recombiner malfunction. It is required to enter 20.712.01, High Hydrogen Concentration / Explosion in the Off-Gas System.

_____ D. an expected increase in Nitrogen-16 (N16) production from the reactor. It is required to notify Chemistry of the power increase and obtain samples per 74.000.19, Chemistry Routine Surveillances.

Correct Answer: A A fuel cladding failure will cause Off Gas Radiation to increase.

It is required to enter 20.000.07, Fuel Cladding Failure.

Plausible Distractors:

B is plausible; increased air inleakage initially increases radionuclide transport rate, short lived nuclides will cause a momentary increase in Radiation Level until air dilution subsequently lowers it. Excluded by Main Condenser Vacuum and OG Flow.

C is plausible; Off Gas alarms can be indicative of Recombiner Failure. Excluded by low Hydrogen Analyzer indication.

D is plausible; higher power does raise N-16 production, Off Gas Delay piping is designed to preclude N-16 changes from causing High Radiation alarms.

Objective Link: LP-OP-315-0135- OBJ B

SRO Tier K/A Number Statement IR Origin Source Question 94 3 Generic 2.1.43 4.3 N NA LOK Grp 10 CFR 55.43(b) 6 LOD (1-5) Reference Documents F NA GOP 22.000.03 Rev 74 Ability to use procedures to determine the effects on reactivity of plant changes, such as RCS temperature, secondary plant, fuel depletion, etc.

QUESTION 94 With the plant in end of cycle coast down, all Control Rods are at position 48 and Reactor Power is 97%. Preparations are being made to shutdown for a Refuel Outage.

Which ONE of the following is an APPROVED method of adding positive reactivity and obtaining additional energy from the core?

_____ A. SECURE Heater Drain Pumps per 23.108, Extraction Steam and Heater Drains.

_____ B. RAISE Core Flow to 110 Mlbm/hr per 22.000.03, Power Operation 25%

to 100% to 25%.

_____ C. RAISE the Pressure Regulator setting per 22.000.03, Power Operation 25% to 100% to 25%.

_____ D. BYPASS Feedwater Heater #6 North and South by opening N2100-F603,

1. 6 FW Heaters Bypass Valve per 23.108, Extraction Steam and Heater Drains.

Correct Answer: C RAISE the Pressure Regulator setting per 22.000.03, Power Operation 25% to 100% to 25% is approved at this power level.

Plausible Distractors:

A is plausible; can be performed at <65% power, will cause a Recirculation Runback at this power level.

B is plausible; lower rod lines accommodate greater core flow, it is NOT permitted to exceed the Licensed Core Flow Limit.

D is plausible; can be performed <50% power, when isolating Feedwater Heaters.

Objective Link: LP-OP-802-1002-0001

SRO Tier K/A Number Statement IR Origin Source Question 95 3 Generic 2.2.14 4.3 B Fermi-2 Bank LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents F NA MOP05 Rev 18 Knowledge of the process for controlling equipment configuration or status.

QUESTION 95 A Work Request has been released to replace the Div 1 RHRSW Radiation Monitor Sample Pump. Due to parts difficulties, the work group is requesting that the Work Request be DEACTIVATED with the Sample Pump removed.

Per MOP05, Control of Equipment, which ONE of the following describes when the Shift Manager MAY deactivate the package?

The Shift Manager MAY deactivate the package:

_____ A. once the protection has been released AND the existing configuration has been evaluated per ODE 6, Operator Challenges.

_____ B. once the protection has been released AND the existing configuration has been evaluated per MES12, Performing Temporary Modifications

_____ C. after noting the deactivation on the Safety Tagging Record AND the existing configuration has been evaluated per ODE 6, Operator Challenges.

_____ D. after noting the deactivation on the Safety Tagging Record AND the existing configuration has been evaluated per MES12, Performing Temporary Modifications Correct Answer: B Per MOP05, Control of Equipment, a Work Request shall not be deactivated when either personnel protection is in effect OR an interim alteration exists that has not been evaluated in accordance with MES12, Performing Temporary Modifications.

Plausible Distractors:

Noting deactivation on the Safety Tagging Record is considered plausible but incorrect. Notes are used frequently on tagouts, but are not in accordance with MOP05. Operator Challenges involve documentation of work arounds, and do not meet the requirements of Temporary Modification Evaluations.

Objective Link: : LP-OP-802-4101-0022

SRO Tier K/A Number Statement IR Origin Source Question 96 3 Generic 2.2.35 4.5 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents F NA Table 1.1-1 Amendment 134 Ability to determine Technical Specification Mode of Operation.

QUESTION 96 Following a Refueling Outage, the following conditions exist:

• ALL RPV Head Closure Bolts are FULLY TENSIONED.

• Reactor Coolant System Temperature is 185°F.

• The Reactor Mode Switch is in REFUEL and Control Rod exercising is in progress.

NOTE: For this particular instance, Special Operations TS 3.10.4 does NOT apply.

Which ONE of the following is the correct MODE of operation, based on these conditions?

_____ A. MODE 2, STARTUP

_____ B. MODE 3, HOT SHUTDOWN

_____ C. MODE 4, COLD SHUTDOWN

_____ D. MODE 5, REFUEL Correct Answer: A MODE 2 STARTUP, with the Reactor Mode Switch in REFUEL and all RPV Head Closure Bolts Fully Tensioned.

Plausible Distractors:

B is plausible; would be true if RCS Temperature exceeded 200°F with the Reactor Mode Switch in SHUTDOWN C is plausible; would be true with Reactor Mode Switch in SHUTDOWN.

D is plausible; would be true with ONE RPV Head Closure Bolt Less Than Fully Tensioned.

Objective Link: LP-OP-8004-0001-0004

SRO Tier K/A Number Statement IR Origin Source Question 97 3 Generic 2.3.5 2.9 N NA LOK Grp 10 CFR 55.43(b) 4 LOD (1-5) Reference Documents H NA EP-547 Rev 6 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(10 CFR 55.43(b) 4 - assessment of an abnormal radiation hazard)

QUESTION 97 Emergency Operations Facilities are NOT manned.

Following an accident, it is required to estimate Core / Fuel Damage using the following Containment High Range Radiation Monitor (CHRRM) readings and conditions:

• Reactor was SHUTDOWN at 1200.

• CHRRM Readings were taken at 1300.

• DIV 1 CHRRM indicates 2.0 x 104 R/hr.

• DIV 2 CHRRM indicates 1.5 x 104 R/hr.

Note: See attached references.

Which ONE of the following is the CORRECT Core / Fuel Damage calculation, based on these readings?

% Gap Release  % of Fermi-2  % of Regulatory Upper Bound Guide 1.3 (H) LOCA LOCA (J) (K)

_____ A. 21.4 5.0 1.9

_____ B. 28.6 6.7 2.5

_____ C. 115.4 30.0 8.8

_____ D. 153.8 40.0 11.8 Correct Answer: B Calculated using Div 1 CHRRM and Enclosure B 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> values for (E) 7 x 104 R/hr, (F) 3 x 105 R/hr , (G)8 x 105 R/hr

Plausible Distractors:

A is plausible; miscalculated using LOWEST reading CHRRM.

C is plausible; miscalculated using 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> vice 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after Shutdown and LOWEST reading CHRRM.

D is plausible; miscalculated using 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> vice 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after Shutdown.

Objective Link: LP-ER-832-0001-0016 LP-OP-801-0001-A002

SRO Tier K/A Number Statement IR Origin Source Question 98 3 Generic 2.3.4 3.7 B 2003 Fermi-2 NRC Exam LOK Grp 10 CFR 55.43(b)4 LOD (1-5) Reference Documents F NA MRP05 Rev 6 Knowledge of radiation exposure limits under normal or emergency conditions.

QUESTION 98 During a DECLARED EMERGENCY, a leak develops in an area that is accessible, but now radiologically contaminated. The Shift Manager has directed that an investigation be performed IMMEDIATELY.

In accordance with MRP05, ALARA / RWPs, what are the RWP REQUIREMENTS for entry into the area for investigation?

_____ A. A written Specific RWP must be issued.

_____ B. A General RWP already exists for this type of event.

_____ C. A revision to the General RWP for that area must be issued.

_____ D. A verbally issued RWP may be used for timely plant response.

Correct Answer: D A verbally issued RWP may be used for timely plant response.

Plausible Distractors:

A is plausible; but not required under emergency conditions.

B is plausible; existing General RWP would not be useful since conditions have changed.

C is plausible; but not required under emergency conditions.

Objective Link: LP-OP-802-4101-0032

SRO Tier K/A Number Statement IR Origin Source Question 99 3 Generic 2.4.20 4.3 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H NA 29.100.01 Sheet 1A Rev 10 Knowledge of operational implications of EOP warnings, cautions, and notes.

QUESTION 99 While executing 29.100.01 Sheet 1A, ATWS RPV Control, Emergency Depressurization is REQUIRED. Conditions are as follows:

• Standby Liquid Injection has been INITIATED.

• RPV Level is -15 inches.

• Reactor Pressure is 1000 psig.

• Injection has been Terminated and Prevented as required.

• FIVE Safety Relief Valves have been OPENED.

• Reactor Pressure is 850 psig LOWERING.

• RPV Water Level is -30 inches LOWERING.

Which ONE of the following actions is required to be ordered?

_____ A. RAISE RPV Water Level above -25 inches using INSIDE the shroud systems FIRST.

_____ B. RAISE RPV Water Level above -25 inches using OUTSIDE the shroud systems FIRST.

_____ C. WAIT until RPV Pressure LOWERS to 230 psig, THEN RAISE RPV Water Level above -25 inches using INSIDE the shroud systems FIRST.

_____ D. WAIT until RPV Pressure LOWERS to 230 psig, THEN RAISE RPV Water Level above -25 inches using OUTSIDE the shroud systems FIRST.

Correct Answer: D with ATWS conditions, ED requires initial pressure reduction to MSCP without injection. THEN, using OUTSIDE the shroud systems FIRST, RPV Water Level may be restored.

Plausible Distractors:

A is plausible; below MINIMUM ATWS Level Control Band.

B is plausible; Would be true if reactor was shutdown under all conditions, per RPV Control.

D is plausible; OUTSIDE the shroud sources are required to be used FIRST, then augmented with INSIDE the shroud sources.

Objective Link: None

SRO Tier K/A Number Statement IR Origin Source Question 100 3 Generic 2.4.23 4.4 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H NA 29.100.01 Sheet 1A Rev 10 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

QUESTION 100 The plant is operating at full power, when a transient occurs, resulting in the following conditions:

• NO Control Rod movement occurred.

• Blue Scram Valve Lights on the Full Core Display are ALL LIT.

• The Turbine Generator is ON LINE.

• Reactor Power is 30%.

• ADS is INHIBITED.

• FSQ 1-8 is COMPLETE.

• RPV Water Level is at 60 inches, LOWERING at 4 inches per minute.

• Torus Water Temperature is 90°F, due to HPCI starting.

Which ONE of the following EOP ACTIONS should have HIGHEST PRIORITY, based on these conditions?

_____ A. Start Torus Cooling per 23.205.

_____ B. Vent the Scram Air Header per 29.ESP.03.

_____ C. Deenergize Scram Solenoids per 29.ESP.03.

_____ D. Defeat RPV Level 1 MSIV Isolation Signals per 29.ESP.11.

Correct Answer: D RPV Level 1 will isolate MSIVs in three minutes. At 30% power, MSIV Closure represents a substantial containment threat.

Plausible Distractors:

A is plausible; Torus Cooling is required, but cannot mitigate the heat load of 30% power dumped to the Suppression Pool.

B is plausible; Control Rod insertion is required. Blue Scram Valve Lights indicate Scram Valves are OPEN and efforts to vent scram air header will be ineffective.

C is plausible; Control Rod insertion is required. Blue Scram Valve Lights indicate Scram Valves are OPEN and efforts to deenergize scram solenoids will be ineffective.

Objective Link: LP-OP-802-3003-003

FERMI 2008 Written Exam Answer Key RO 1. A 41. C SRO 76. B

2. D 42. B 77. C

3. D 43. D 78. A

4. D 44. A 79. B

5. B 45. D 80. B

6. D 46. B 81. C

7. B 47. B 82. D

8. D 48. A 83. A

9. C 49. C AND D 84. D

10. C 50. A 85. B

11. C 51. A 86. C

12. A 52. C 87. B

13. D 53. D 88. C

14. B 54. A 89. C

15. B 55. C 90. C

16. B 56. D 91. C

17. A 57. B 92. D

18. D 58. C 93. A

19. D 59. A 94. C

20. A 60. A 95. B

21. B 61. C 96. A

22. C 62. A 97. B

23. B 63. D 98. D

24. D 64. C 99. D

25. D 65. B 100. D

26. A 66. B

27. A 67. D

28. B 68. C

29. B 69. D

30. B 70. B

31. C 71. B

32. D 72. D

33. C 73. D

34. B 74. B

35. A 75. C

36. A

37. D

38. A

39. B

40. D 1