2008 NRC Flashcards by Nathan Waters

Operators are performing 2-OHP-4023-ECA-2.1, Uncontrolled Depressurization of All Steam Generators due to a steam leak inside containment along with failure of all SG stop valves to close.
Given the following plant conditions:
• Cooldown rate is stable at 155o F per hour.
• RCS cold leg temperatures are 340o F and lowering.
• Containment pressure is 8 psig and lowering.
• Narrow range Steam Generator levels indicate offscale low.
• Steam Generator AFW flow indicates 100x103 pph to each SG.
Which ONE of the following choices is correct for these plant conditions?
a. Adjust AFW flow to 60x103 pph on each Steam Generator.
b. Adjust AFW flow to 25x103 pph on each Steam Generator, AFTER at least one SG narrow range level is greater than 13%.
c. Adjust AFW flow to 25x103 pph on each Steam Generator.
d. Isolate AFW flow to ONLY three of the Steam Generators.

c. CORRECT. AFW flow is throttled to 25x103 pph on each Steam Generator if the cooldown rate is > 100oF per hour and level is less than 28% (Adverse Containment).

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Given the following plant conditions on Unit 2:
• Reactor is at 100% power.
• All loop Tcolds are 540°F.
• All loop Thots are 606°F.
• Average of the 5 highest Core exit TCs is 620°F.
• A malfunction occurred with the pressurizer pressure control system resulting in an RCS pressure of 1915 psig without a change in reactor power.
Which ONE of the following describes the change from normal full power values to RCS subcooling based on the pressure transient?
Subcooling will:
a. lower to 11°F
b. lower to 22°F
c. rise to 58°F
d. rise to 91°F

a. CORRECT. Saturation Temperature for 1930 psia (1915 psig) is ~631°F. Comparing thermocouples to Tsat yeilds. 631°F - 620.00°F = 11°F. Pressurizer Vapor space temperature is 650 at 2235 psig.

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During a Large Break LOCA, an evaluation of plant status is made during Step 11 of 1-OHP-4023-E-1, Loss of Reactor or Secondary Coolant. Part of this evaluation includes a check of ECCS pump compartment sump alarms and auxiliary building vent stack and area radiation monitors.
Which ONE of the following reasons describes the BASIS for checking these alarms and radiation monitors in this procedure?
a. Determine if local actions can be performed without excessive personnel exposure.
b. Determine if ECCS leakage exceeds that assumed in the Control Room dose analysis.
c. Determine if a transition should be made to address a LOCA outside of Containment.
d. Collect current radiation values to assist in Emergency Event classification.

c. CORRECT. Plant sump alarms and radiation monitors are both checked to identify leakage in the auxiliary building. This check is made to determine if the operator should make a transition to 1-OHP-4023-ECA-1.2, LOCA Outside Containment.

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Given the following plant conditions:
• A Reactor Trip and Safety Injection have occurred.
• While implementing 2-OHP-4023-E-0, Reactor Trip or Safety Injection Step 10, it is determined that AFW flow can NOT be established.
• All SG NR levels are off-scale low.
• All SG WR levels are 58% and lowering.
• The crew has just entered 2-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink.
• RCS Pressure is 175 psig and stable.
• Intact SG pressures are 475 psig and trending down.
Which ONE of the following describes the plant conditions and action required?
Steam Generators are:
a. required to provide secondary heat sink. Initiate Bleed and Feed per 2-OHP-4023-FR-H.1.
b. NOT required to provide secondary heat sink. Go to 2-OHP-4023-ES-0.0 Rediagnosis.
c. required to provide secondary heat sink. Remain in 2-OHP-4023-FR-H.1 to establish AFW Flow.
d. NOT required to provide secondary heat sink. Return to 2-OHP-4023-E-0.

d. CORRECT. Secondary heat sink is not required if SGs are at a higher pressure than the RCS. They act as a heat source. Crew returns to E-0.

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Given the following plant conditions on Unit 1:
• A small break LOCA is in progress.
• Only one train of SI has actuated.
• RCS Pressure is 1290 psig.
• RCS Temperature is 703oF.
In order to prevent fuel damage from inadequate core cooling, what is the reason for maintaining a secondary heat sink?
a. To provide an alternate means of RCS pressure control.
b. Reflux boiling is the primary means of heat removal prior to voiding in the hot legs.
c. To ensure removal of RCS heat since the RCPs are expected to be running.
d. RCS pressure may remain so high that cooling from injection flow alone is inadequate.

d. CORRECT. Mass loss out the break is not sufficient to lower RCS pressure to a point where energy loss through the break along with injection flow is sufficient to address all decay heat removal requirements. The SG will aid in removing some of the excess decay heat.

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Unit 2 Reactor Startup is in progress with Reactor Power at 2E-8 amps and rising.
Given the following conditions on RCP 23:
• Calculated #2 Seal Leak Rate is 1.6 gpm.
• Lower Bearing water temperature is 200°F and rising.
• Motor Bearing temperature is 174°F and stable.
• Seal Leakoff temperature is 179°F and rising.
• Seal Injection Flow is 10 gpm.
• Vibrations are at 16 mils and stable.
Which ONE of the following operator actions MUST be taken based upon these conditions?
a. Manually trip the reactor, Enter 2-OHP-4023-E-0, Reactor Trip or Safety Injection, perform immediate actions, then trip the No. 3 RCP.
b. Initiate reactor shutdown per 2-OHP-4021-001-003, Power Reduction and trip the No. 3 RCP after the reactor is shutdown.
c. Do NOT trip the reactor. Trip the No. 3 RCP and be in Hot Shutdown in 1 hour.
d. Do NOT trip the reactor. Trip the No. 3 RCP and close the No. 1 seal leakoff valve.

b. CORRECT. The #2 seal is failing and will require a controlled shutdown to remove the pump from service. Reactor trip is not required, since all other parameters are within limits of the procedure and fold-out page.

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While responding to a LOCA, a transition to OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation, was performed due to a loss of emergency coolant recirculation.
Given the following conditions:
• RWST level is 18% and lowering.
• Containment Pressure is 2.5 psig
• All ECCS pumps are running.
• Both CTS pumps were stopped.
• RCS Pressure is 940 psig.
Make up is being added to the RWST and ECCS is reduced to one train of SI flow.
Which ONE of the following describes the reason for these actions?
a. Restore RWST level so Containment Spray can be started.
b. Prevent damage to vital equipment by saving one ECCS train.
c. Establish conditions to allow restart of RCPs.
d. Delay the time to RWST depletion.

d. CORRECT. Makeup is added to the RWST to extend time the ECCS pumps can take suction from the RWST and supply core cooling. Reducing to one train of ECCS flow (1 CCP, SI, & RHR) delays the time to RWST depletion.

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Unit 2 Reactor power is at 100%.
Given the following plant conditions:
•	QRV-251 Charging Flow Controller is in MANUAL since Automatic control has failed.
•	PRZ level is stable at program level.
•	Charging and letdown are balanced.
Which ONE of the following describes the impact on the plant if a transient causes power to be lowered to 30% while 2-QRV-251 remains in MANUAL?
a.	Charging flow will RISE.
b.	Charging flow will LOWER.
c.	PRZ level will LOWER.
d.	PRZ level will RISE.

c. CORRECT. With QRV-251 Charging Flow Controller in Manual Charging flow will remain constant. As the RCS cools down, the Pressurizer Level will lower as the water contracts.

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Given the following plant conditions on Unit 2:
• A loss of all AC power has occured.
• The crew has just completed the immediate actions of 2-OHP-4023-ECA-0.0, Loss Of All AC Power.
Why does the procedure place a priority on checking for isolation of the RCS?
This action:
a. minimizes the RCS inventory depletion and maximizes the time until core uncovery.
b. ensures all RCS leak paths are isolated, thus preventing core uncovery.
c. maintains sufficient pressurizer level for proper control of the heaters and sprays for RCS pressure control.
d. prevents potential for rupturing the PRT due to excessive opening of the pressurizer PORVs.

a. CORRECT. The step for RCS isolation ensure that the only leakage path is by way of the RCP seals. This action minimizes the RCS leakage and extends the time until core uncovery.

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Given the following plant conditions on Unit 2:
•	The unit was operating 100%.
•	A loss of 250 VDC Train A occurs.
Initially, CRID 1 and 2 power supply will be \_\_ (1)\_\_ aligned to \_\_(2)\_\_.
a.	1) automatically
        2) a 600 VAC vital bus.
b.	1) manually
        2) a 600 VAC vital bus.
c.	1) automatically
        2) CRP-3.
d.	1) manually
        2) CRP-3.

a. CORRECT. Auto transfer will occur to the vital bus on a loss of the normal 250VDC feed to the inverter.

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Unit 2 was operating at 100% when a Large Break Loss of Coolant Accident occurred thirty minutes ago.
Given the following current plant conditions:
• All Control Rods are inserted.
• RCS Pressure is 40 psig.
• Pressurizer Level is 0%.
• All equipment has functioned as designed.
• RCS Subcooling is 0°F.
• Containment Radiation levels are in alarm and slowly rising.
• Containment Pressure is 4 psig and slowly lowering.
Which ONE of the following describes the status of automatic Containment Isolation?
a. Phase A was DIRECTLY actuated by a High Containment Pressure Signal, and Phase B was DIRECTLY actuated by a High-High Containment Pressure Signal.
b. Phase A was DIRECTLY actuated by the Safety Injection Signal and Phase B was DIRECTLY actuated by a High-High Containment Pressure Signal.
c. Phase A was DIRECTLY actuated by the Safety Injection Signal and Phase B has NOT actuated.
d. Phase A has NOT actuated and Phase B has NOT actuated.

b. CORRECT. Phase A has actuated on the Safety Injection Signal and Phase B has actuated on High-High Containment Pressure Signal.

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Which ONE of the following describes the procedural actions in response to addressing a leaking Pressurizer (PRZ) PORV?

a. 1. All PORV block valves are initially closed to lower tailpipe temperature.
2. One PORV block valve is opened at a time.
3. Leakage is determined by a rise in tailpipe temperature after each PORV block valve is re-opened.
b. 1. PORV block valves are closed one at a time.
2. Temperature on the tailpipe is monitored by the operator.
3. Leakage is determined by a lowering of tailpipe temperature after each PORV block valve is closed.
c. 1. PORV block valves are closed one at a time.
2. Temperature on the Pressurizer Relief Tank (PRT) is monitored by the operator.
3. Leakage is determined by a lowering PRT temperature after each PORV block valve is closed.
d. 1. All PORV block valves are initially closed to stabilize Pressurizer Relief Tank (PRT) temperature.
2. One PORV block valve is opened at a time.
3. Leakage is determined by a rise in PRT temperature after each PORV block valve is re-opened.

a. CORRECT. The procedure requires that all PORV Block Valves be initially closed. Once tailpipe temperature is lowering, the block valves are opened 1 at a time to check for a rise in tailpipe temperature.

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Unit 1 Reactor has been holding at 30% for the last hour due to Chemistry.
Given the following plant conditions:
• West Main Feedwater pump is stopped.
• East Main Feedwater pump is operating.
• AFW pumps are aligned for automatic operation.
• AMSAC is aligned in NORMAL.
The East Main Feedwater pump trips and a manual Reactor Trip is initiated.
The current conditions exist:
• Narrow Range Steam Generator levels lowered to 33%.
• Steam Dumps indicate 10% open.
Which ONE of the following statements correctly describes the AFW pump status after the Reactor Trip with no further operator actions?
a. The Motor Driven and Turbine Driven AFW pumps have NOT started.
b. The Motor Driven AFW Pumps have auto started but the Turbine Driven AFW pump has NOT started.
c. The Turbine Driven AFW Pump has auto started but the Motor Driven AFW pumps have NOT started.
d. The Motor Driven and Turbine Driven AFW pumps have all auto started.

b. CORRECT. The Motor Driven AFW pumps will auto start on the loss of both Main FW pumps but the Turbine driven AFW pump will not.

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Given the following plant conditions:
• An AEO has reported a major air leak in the plant air system.
• The crew is implementing 2-OHP-4022-064-001, Control Air Malfunction.
Which ONE of the following pressures would require the SRO to direct a manual Reactor Trip and enter 2-OHP-4023-E-0, Reactor Trip or Safety Injection?
a. Plant Air header pressure lowers to 79
b. Plant Air header pressure lowers to 84
c. Control Air header pressure lowers to 79
d. Control Air header pressure lowers to 84

c. CORRECT. 02-OHP-4022-064-001, Control Air Malfunction directs a reactor trip if Control Air Pressure is <80 psig.

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Given the following plant conditions on Unit 2:
• A Loss of All AC Power has occurred.
• The crew is performing the actions of 2-OHP-4023-ECA-0.0, Loss of All AC Power.
• The operators have been unable to restore power.
Which ONE of the following describes the actions required for the safeguards equipment and why?
a. The control switches for the safeguards equipment are placed in Pull-to-Lock to prevent a potential bus overload when power is restored.
b. The control switches for the safeguards equipment are placed in Pull-to-Lock to prevent the possibility of an uncontrolled cooldown and depressurization of the RCS when power is restored.
c. The control switches for the safeguards equipment are verified to be in automatic alignment on the 4KV bus(es) that get power back so if SI is required, it will occur without operator action.
d. The control switches for the safeguards equipment for ONE train are placed in Pull-to-Lock and the control switches for the other train are verified to be in automatic. This ensures alignment for injection without operator action.

a. CORRECT. All major loads on the bus are stripped prior to energizing the bus to prevent excessive loading as pumps initially start.

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Given the following plant conditions on Unit 2:
• The unit was operating at 12% power.
• Reserve Aux Transformer 201AB was supplying Buses 2A and 2B.
• Reserve Aux Transformer 201CD was supplying Buses 2C and 2D.
An electrical transient occured on the Grid.
During the transient the following values were reached:
Electrical Bus
• Parameter 2A 2B 2C 2D
• Frequency (Hz) 57 57 59.7 59.7
• Local Voltage (VAC) 4100 4100 4200 4200
• CR Indication (VAC) 118.2 118.2 121.1 121.1
• Ann. 207, Drop 10, RCP BUSSES UNDER FREQ TRIP is LIT
Which ONE of the following describes the expected status of the reactor trip breakers and the Reactor Coolant Pumps?
Note: Assume no operator action.
a. Reactor Trip Breakers remain Closed
All 4 RCPs remain running.
b. Reactor Trip Breakers Trip Open.
All 4 RCPs remain running.
c. Reactor Trip Breakers Trip Open.
All 4 RCPs trip off.
d. Reactor Trip Breakers remain Closed
ONLY 2 RCPs remain running.

b. CORRECT. The RCPs no Longer trip on Underfrequency (Recent Plant Modification)

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The crew was operating Unit 2 at 100% when High secondary radiation and lowering PRZ level caused the crew to suspect a SG tube leak or rupture.
Which ONE of the following describes the THRESHOLD condition for initiating a reactor trip and safety injection per 02-OHP-4022-002-021, Steam Generator Tube Leak?
The THRESHOLD is when Pressurizer level can not be maintained with __(1)__ CCP(s) in service and letdown __(2)__ .
a. 1. one
2. in service
b. 1. one
2. isolated
c. 1. two
2. in service
d. 1. two
2. isolated

b. CORRECT. The procedure directs the operator to maximize charging flow and reduce/then isolate letdown. If PZR level can NOT be maintained in this configuration it is a rupture.

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Given the following plant conditions:
• Both Units are in Mode 1.
• Engineering has determined that a relay has failed that will prevent the SI auto-start of the U1 East Essential Service Water (ESW) pump.
• All of the other U1 East ESW Pump start signals will function as designed.
• The Unit 1 East ESW discharge crosstie valves are de-energized in the open position while maintenance replaces the motors on the valves.
Which ONE of the following describes the operability and Technical Specification (TS) applicability associated with these conditions?
a. All ESW related trains are still OPERABLE because Unit 1 East ESW pump can still be manually started, and a service water TS LCO action statement would NOT be entered.
b. Only Unit 1 East ESW Train is INOPERABLE and a service water TS LCO action statement would be entered because the auto start is required to be operable.
c. All ESW related trains are still OPERABLE because the Unit 1 East ESW pump will start automatically if the discharge pressure falls below 40 psig and a service water TS LCO action statement would NOT be entered.
d. Both Unit 1 East and Unit 2 West ESW Trains are INOPERABLE and the ESW TS LCO action statements for both units must be entered because the crossties are open.

d. CORRECT. If the crossties are open then the LCO action statement also applies to Unit 2.

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Given the following plant conditions:
• The plant was at 80% power and stable at EOL.
• A fault caused a continuous rod withdrawal.
• The rod motion was stopped after 20 steps by placing the Rod Control to MANUAL.
• The reactor did NOT trip.
What were the final effects of the fuel temperature and moderator temperature coefficients when the plant had stabilized?
a. The fuel temperature and moderator temperature coefficients have added positive reactivity
b. The fuel temperature and moderator temperature coefficients have added negative reactivity.
c. The fuel temperature coefficient added negative reactivity and moderator temperature coefficient have added positive reactivity.
d. The fuel temperature coefficient added positive reactivity and moderator temperature coefficient have added negative reactivity

b. CORRECT. Both the MTC and FTC are negative at EOL.

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Given the following plant conditions:
• A Small Break LOCA occurred 12 hours ago.
• Containment pressure is 1.2 psig.
• Containment air temperature is 215ºF.
• OHP-4023-FR-Z-3, Response to High Containment Radiation Level, is entered.
Which ONE of the following verifications is a major action category of OHP-4023-FR-Z-3?
a. Both Containment Recirculation Fans are running.
b. Upper and Lower Containment Ventilation Fans are running.
c. Containment Ventilation Isolation has occurred.
d. Control Room Ventilation System is in ISOLATE.

c. CORRECT. Step 1 of the procedure ensures that CVI has occured.

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Given the following plant conditions on Unit 2:
• A LOCA occurred thirty minutes ago.
• Containment Pressure has risen to 5 psig.
• The crew has completed steps of 2-OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation, to align RHR/CTS suctions to the recirculation sump and the CCP/SI suctions to RHR Discharge.
• ONLY the Train A CCP, SI, RHR, and CTS pumps are operating.
• The next step of 2-OHP-4023-ES-1.3 directs the crew to “Check if RHR Spray is Required”.
Based on the indications above, which ONE of the following would best describe the required action AND the reason for the decision?
a. Place RHR spray in service NOW since ALL of the requirements are met.
b. Place RHR spray in service ONLY after RHR has injected for 50 minutes to ensure adequate core cooling.
c. Do NOT place RHR spray in service because the RHR pump suction is NOT aligned to the RWST.
d. Do NOT place RHR spray in service because ONLY one RHR pump is operating.

b. CORRECT. RHR is required if only 1 CTS pump is operating. After RHR has injected for 50 minutes the core is sufficiently cooled to allow RHR to be diverted to support spray functions.

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Given the following plant conditions on Unit 1:
• The unit is at 100% power.
• Pressurizer Level Control selector is in the 1-2 position.
• Pressurizer Level Channel 1 (NLP-151) fails low.
Which ONE of the following describes the reason that Charging Flow and/or PZR Level control are placed in Manual in accordance with OHP-4022-013-010, Pressurizer Level Instrument Malfunction?
If NO operator action is taken following the channel failure:
a. charging flow will lower and PRZ level will lower until heaters are de-energized.
Rx Trip will occur on OTDT from the lowering pressure.
b. charging flow rises and PRZ level rises.
PRZ PORVs will open as the steam space is compressed.
c. charging flow will lower and PRZ level will lower until heaters are de-energized.
PZR pressure will lower to Rx Trip setpoint.
d. charging flow rises and PRZ level rises.
PRZ level will rise to a Hi Lvl Rx Trip setpoint.

d. CORRECT. Charging flow rises and PRZ level rises (due to charging flow and loss of letdown). PRZ level will continue to rise to a Hi Lvl Rx Trip.

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Given the following plant conditions on Unit 1:
• A Small Break LOCA has occurred.
• Offsite Power was lost on the reactor trip.
• The actions of 1-OHP-4023-ES-1.2, Post LOCA Cooldown And Depressurization, are in progress.
• Both CCPs are running with suction aligned to the RWST.
• Both RHR Pumps are stopped in Neutral.
• Both SI Pumps are running.
• The crew is ready to depressurize the RCS to refill the Pressurizer.
Which ONE of the following is the FIRST method available to the operator to commence the RCS depressurization?
The operator will open:
a. One PZR PORV to vent the PZR.
b. All Pressurizer PORVs to vent the PZR.
c. The PZR Aux Spray Valve to spray down the PZR steam space.
d. One PZR Normal Spray Control valve to spray down the PZR steam space.

a. CORRECT. Since normal spray is not available (RCP Busses De-energized from loss of offsite power), the next choice is to use one PZR PORV.

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Given the following plant conditions on Unit 2:
• The unit is at 100% power
• One Control Rod in Bank D Group 1 was found stuck at 190 steps.
• While aligning the remainder of the rods in Bank D to 190 steps an additional Control Rod in Bank D Group 2 was found stuck at 210 steps.
• It has been determined that both rods are mechanically bound.
In accordance with Technical Specifications, which ONE of the following describes the action required within one hour?
a. Determine that Shutdown Margin requirements are satisfied.
b. Determine that QPTR requirements are satisfied or enter the applicable action statement.
c. Verify all peaking factors are within acceptable limits.
d. Align the remainder of rods in the affected banks within 12 steps of the stuck rods.

a. CORRECT. Technical Specifications 3.1.4, Rod Group Alignment Limits, Condition A, requires SDM verification within 1 hour.

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Following a rapid power reduction from 80% power due to a Feedwater pump trip, the following plant conditions exist on Unit 1: • Reactor power is 59%. • Turbine Power is 620 Mwe. • Rod Control is in MANUAL. • All other controls in AUTO. • Feedwater and Steam Flows are equal. • An Emergency Boration is being performed per 1-OHP-4021-005-007, Operation Of Emergency Boration Flow Paths, due to the ROD BANK D LOW-LOW (Rod Insertion Limit) alarm being lit. Given these conditions, which ONE of the following would be used to verify that negative reactivity is being added? a. Tref lowering with a negative Start-up Rate. b. Tavg lowering with approximately -0.3 dpm Start-up Rate. c. ROD BANK D LOW-LOW alarm clearing with Tref Lowering d. A flow of 45 gpm indicated on QFI-410 with Tavg Lowering

d. CORRECT. Procedually flow rate is > 44 gpm. Tavg lowering would indicate that flow is reaching the core. b. is incorrect because that is a post trip SUR.

During the response to a LOCA, the crew implements 2-OHP-4023-FR-C-1, Response to Inadequate Core Cooling. The crew is about to depressurize the SGs to atmospheric conditions. Given the following plant conditions: • Containment pressure is 3.2 psig and lowering. • A loss of CRID 1 and 2 has resulted in no power available to the SG PORV Controllers. Which ONE of the following procedural actions will be needed to depressurize the SGs? a. Dump steam to condensers using the steam dump system. b. Depressurize the SGs using the local SG PORV backup control stations. c. Dump steam using the Turbine Driven AFW Steam Supply. d. Remove control air from the SG PORVs to fail the valves open.

b. CORRECT. 2-OHP-4023-FR-C.1 dictates to use the local PORV stations in the even that the normal steam dumps and control of SG PORVs from the control room is unavailable.

Given the following plant conditions on Unit 2: • The unit is operating at 40% power. • Chemistry has reported the following confirmed sample results. • Dose Equivalent I-131: 125 μCi/gram • Gross Activity: 60 μCi/gram • E-bar: 1.40 Mev Which ONE of the following describes the required actions and/or limits? Note: Tech Spec 3.4.16, RCS Specific Activity is attached. a. The Dose Equivalent I-131 value requires that power be maintained at slightly less than 64%. b. The Gross Activity value requires that power be maintained at slightly less than 80%. c. The Dose Equivalent I-131 value requires the plant to be in Mode 3 with Tavg less than 500°F within 6 hours d. The Gross Activity value requires the plant to be in Mode 3 with Tavg less than 500°F within 6 hours

a. CORRECT. 100/E-bar is 71 so gross activity is less than required. Since Dose Equivalent I-131 is > 1 μCi/gram Action A.1 applies. The required action states that LCO 3.0.4.c is applicable which allows power to be raised to the limit per the applicable figure 3.4.16-1 (~64%)

Given the following sequence of events: • Unit 1 and Unit 2 were operating at 100% power. • Unit 1 and Unit 2 East Essential Service Water (ESW) pumps were operating with the Unit Crossties open. • Unit 2 tripped due to a turbine Electro-Hydraulic Control fluid leak. • Unit 1 remained on line. • The Unit 2 Reserve Transformers are unavailable. • Both Unit 2 Emergency Diesel Generators (EDGs) started and loaded; however, Bus T21D failed to energize. Assuming NO operator actions, which ONE of the following describes the ESW cooling water status for the Unit 2 EDGs? a. 2CD EDG must be tripped immediately as ESW cooling has been lost. b. 2CD EDG has ESW cooling supplied by the Unit 2 West ESW Pump. c. 2AB EDG must be tripped immediately as ESW cooling has been lost. d. 2CD EDG has ESW cooling supplied by the Unit 1 West ESW Pump.

a. 2CD EDG must be tripped immediately as ESW cooling has been lost. * Note. The 2008 exam has a different answer based on configuration of the normal cooling valves being open. This is no longer the case. The normal cooling valves are normally closed and require the associated 600 VAC bus to open and provide cooling.

Given the following plant conditions on Unit 2: • The East CCW HX is in service with the West CCW Pump running. • CCW Surge Tank level is stable. • CRS-4301, East CCW HX Radiation Monitor, generates an External Failure Alarm due to a faulty flow switch. Which ONE of the following describes the response of the CCW system for the given conditions? a. No automatic actions will occur since the West CCW pump is running b. No automatic actions will occur since the CRS-4401, West CCW HX Radiation Monitor is still functioning. c. 2-CMO-420, West CCW HX Outlet, opens and 2-CMO-410, East CCW HX Outlet, closes d. 2-CRV-412, CCW Surge Tank Vent Valve, will automatically close.

d. CORRECT. 2-CRV-412, CCW Surge Tank Vent Shutoff Valve closes on High Rad Level Alarm, Low Sample Flow, External Failure on CRS-4300/4400, Channel 4301 - East and/or Channel 4401-West.

Given the following plant conditions on Unit 1: • The unit was operating at 100% power. • A Loss of Control Air has occured. • Control Air Header is reading 0 psig. Which ONE of the following describes the effect of the loss of control air on the Main Steam Isolation Valves (MSIVs). The MSIVs: a. will remain open and can be closed using the hydraulic unit. b. will remain open and can be closed by locally aligning nitrogen to one MSIV dump valve. c. close due to MSIV dump valves failing open on loss of air. d. close due to loss of control air to the hydraulic unit.

c. CORRECT. MSIV dump dumps fail open on loss of air causing the MSIV to close on steam delta-p on the steam actuating piston.

Unit 2 was operating at about 100% power when a Complete Loss of Onsite and Offsite AC power occurred 3 hours ago. Unit 2 dispatched operators after 30 minutes to shed the large Non-Essential DC loads. Given the following plant conditions: • The crew transitioned to 2-OHP-4023-ECA-0-0, Loss Of All AC Power, and stabilized SG pressures. • Power has just been restored from Emergency Power. • While performing Step 30, power could NOT be restored to the battery chargers for the N train, 2AB, and 2CD 250VDC buses. (The actions of step 30 to restore 600V AC Busses and Control Room Cooling were successfully performed.) Which ONE of the following describes the impact of failing to restore the battery chargers? When the Batteries completely discharge: a. all AFW flow will be lost when the AFW pump discharge valves fail closed b. the ability to start and stop ECCS pumps from the control room will be lost. c. the Emergency Power feed breaker will trip open resulting in another Loss of AC. d. all vital instrumentation will be lost.

b. CORRECT. A loss of DC control power will prevent breaker operations with the control switch (and trip functions).

Unit 1 was operating at 100% power when a Large Break LOCA occured. The West RHR pump tripped on overload and can NOT be restarted. SI has been reset and the crew has just stopped both Emergency Diesel Generators and returned to standby in accordance with 1-OHP-4023-E-1, Loss of Reactor or Secondary Coolant. The following sequence of events occurs: • The T11D, 4kV AC ESF Bus subsequently loses normal power but is re-energized by the 1CD Emergency Diesel Generator. • T11D Automatic load sequencing is complete. Which ONE of the following statements correctly describes the status the East RHR? The East RHR Pump: a. has tripped and automatically restarted. b. was NOT affected by the loss of Bus T11D. c. has tripped and may be manually started immediately. d. has tripped and may NOT be manually started until the load conservation signal has been reset.

c. CORRECT. The ECCS timer and SI signals are the only auto starts for the RHR Pumps. The pumps will NOT automatically start following a load shed. The East RHR is powered from T11D while the West RHR is powered from T11A.

Unit 1 is in Mode 3. The 4160 VAC distribution system is being supplied by the Reserve Auxiliary Transformers (RATs). Due to a system disturbance, indicated voltage on the safeguards buses drops. Given the following plant conditions: • T11A Voltage Indication is 112 Volts. • T11B Voltage Indication is 114 Volts. • T11C Voltage Indication is 113 Volts. • T11D Voltage Indication is 114 Volts. Which ONE of the following describes the FINAL plant response if voltage remains at these values for an extended period? a. All safeguards busses will be energized by their respective EDG. b. T11A and T11C busses will be energized by their respective EDG. c. T11A and T11B busses will be energized by its respective EDG. d. ONLY T11A bus will be energized by its respective EDG.

c. CORRECT. An Undervoltage condition of 113 V will energize 62-1 T11A. After a 111 Second delay it will open T11A9 and T11B1 causing T11A and T11B to lose power. This will cause the EDG to start and energize T11A and T11B.

Which ONE of the following describes the effects of continued power operation with a leaking pressurizer PORV? (Assuming NO operator actions.) a. There are NO adverse effects. The PRT is designed to handle continuous in-leakage. b. The cyclic temperature stresses in combination with inner wall erosion on the PORV tailpipe may lead to premature piping failure. c. The PRT rupture disc may break with subsequent elevated radiation, temperature and pressure indications in containment. d. Mechanical breakdown of the PORV seating surface may cause the PORV to fail when needed for overpressure protection.

c. CORRECT. The tank design is based on the requirement to condense and cool a discharge of pressurizer steam equal to 110 percent of the volume above the 100%-power pressurizer water level set-point. If the temperature in the tank rises above 126°F during plant operation, the tank is cooled by spraying in cool water and draining out the warm mixture to the Waste Disposal System. The tank is not designed to accept a continuous discharge from the pressurizer.

Given the following plant conditions: • Lower Containment Cooling NESW supply is throttled to all ventilation units. • A failure of a Lower Containment Cooling Supply regulator results in full flow through the cooling coils. • Average containment temperature lowers from 119°F to 109°F. • Assume RCS Pressure and Temperature remain Constant. Which ONE of the following describes the change in indicated Pressurizer level due to this lowering in Containment temperature? a. Level indicates higher than actual due to reference leg density lowering. b. Level indicates lower than actual due to reference leg density rising. c. Level indicates higher than actual due to reference leg density rising. d. Level indicates lower than actual due to reference leg density lowering.

b. CORRECT. Pressurizer Level uses a wet reference leg DP level indicator. This compares the pressure of the full reference leg with the pressure of the actual water in the pressurizer. When these are equal the level indicates 100%. As the temperature in Containment and therefore the reference leg lowers the density & weight of the reference leg rises. This means that the level in the pressurizer will indicate lower for the same initial actual level.

Which ONE of the following describes the leakoff flow of the RCP #2 seal during normal operations? a. 3 gph to the Reactor Coolant Drain Tank via RCP standpipe. b. 2 gpm to the Volume Control Tank. c. 100 cc/hr to the Reactor Coolant Pump standpipe. d. 100 cc/hr directly to the Reactor Coolant Drain Tank.

a. CORRECT. RCP #2 Seal Leakoff flowpath is directed to the #2 Seal Standpipe, which flows through and orifice to the RCDT.

Which ONE of the following sets of parameters lists the design features provided for CTS pump protection when pump suction is aligned to the containment recirc sump? 1) Minimum Recirculation Level lights 2) Recirc Sump Level Low Alarm 3) Low Suction Pressure CTS Pump Trip 4) Low Sump Level CTS Pump Trip 5) Low CTS Pump Suction Pressure Alarm 6) Low Sump Level Status Light a. 1, 2, 6 b. 1, 3, 5 c. 2, 3, 4 d. 4, 5, 6

a. CORRECT. Indications of adequate level in the containment recirc sump include minimum level status lights, low recirc sump level alarm, and associated status lights.

Which ONE of the following sets contains ONLY conditions that will cause the UNIT 1 pressurizer heaters to automatically deenergize? 1. Pressurizer Level Control Channel - 5% below program level. 2. Pressurizer Level Control Channel - less than 17% level. 3. Pressurizer Level Bistable Channel - less than 17% level. 4. Pressurizer Level Cold Calibration Channel - less than 17% level. a. 1 and 3 b. 2 and 3 c. 2 and 4 d. 1 and 4

b. CORRECT. Protection is provided when either the CONTROL or BISTABLE channel is <17%

With an event in progress, noble gas readings are rising rapidly on the Unit Vent Radiation Monitor VRS-1500. Which ONE of the following directly shifts the sample air flow through VRS-1500 to the emergency lineup (opens 1-VRV-317, Bypass Valve and closes 1-VRV-318, Monitor Valve)? a. High alarm on VRS-1505, Unit Vent Effluent Low Range Noble Gas Radiation Monitor. b. Alert alarm on VRS-1505, Unit Vent Effluent Low Range Noble Gas Radiation Monitor. c. Alert alarm on VRS-1507, Unit Vent Effluent Mid Range Noble Gas Radiation Monitor. d. High alarm on VRS-1509, Unit Vent Effluent High Range Noble Gas Radiation Monitor.

d. CORRECT. High alarm on Unit vent effluent high range noble gas radiation monitor VRS-1509 will Close1-VRV-318 and Open 1-VRV-317.

Given the following plant conditions: • Containment pressure instrument Channel #3, 2-PPP-301, declared inoperable. • Required actions per 2-OHP-4022-013-011 Containment Instrumentation Malfunction have been completed. CRID 2 has just lost power. Which ONE of the following statements explains the impact on the Safety Injection system and expected operator actions? a. No Safety Injection logic is satisfied. Implement 2-OHP-4021-082-008, Operation of CRID Power Supplies, to address the CRID Failure. b. Only Train "A" safety injection actuation logic was satisfied, but equipment will NOT automatically actuate. Implement 2-OHP-4023-E-0, Reactor Trip or Safety Injection, and inform the operators that they must individually reposition/start each Train "A" component. c. Train "A" and "B" safety injection actuation logic were satisfied. Implement 2-OHP-4023-E-0, Reactor Trip or Safety Injection. Both Trains equipment will automatically actuate. d. Train "A" and "B" safety injection actuation logic were satisfied, but Train "A" equipment will NOT automatically actuate. Implement 2-OHP-4023-E-0, Reactor Trip or Safety Injection, and inform the operators that they must individually reposition/start each Train "A" component.

c. CORRECT. Channels 2, 3, & 4 of Containment Pressure cause an SI. This signal will result from the bistables being tripped on Channel 3 & the Loss of Power to Channel 2 bistables. The SI signal is actuated on Both SI trains. Train A SSPS output is powered from CRID 1 & Train B is powered from CRID 4 so all actuations will occur automatically.

Given the following plant conditions on Unit 1: • The unit is at 100% power when a Reactor Trip and SI occurs. • While verifying AFW flow the operator notes the following: • The West MDAFP has tripped on overload. • The TDAFP has failed to start. • The East MDAFP total flow is 250 x 103 pph. Which ONE of the following describes the capabilities of the East AFW Pump with regards to the minimum AFW flow for RCS Decay Heat removal and the SGs being supplied from this pump? The East MDAFP will: a. supply flow to ALL 4 SGs. This flow meets the minimum required for RCS Decay Heat Removal. b. supply flow to the 2 and 3 SGs. This flow meets the minimum required for RCS Decay Heat Removal. c. supply flow to ALL 4 SGs. This flow alone does NOT meet the minimum required for RCS Decay Heat Removal. d. supply flow to the 2 and 3 SGs. This flow alone does NOT meet the minimum required for RCS Decay Heat Removal.

b. CORRECT. Minimum flow required for initial decay heat removal is 240K pph and the East MDAFP is aligned to automatically feed 2 and 3 SGs.

Unit 2 has experienced a loss of both CCW pumps in Mode 3. Given the following plant conditions: • NEITHER Unit 2 CCW pump can be restarted. • BOTH Unit 2 CCPs are running because a CCP swap was in progress. • 2-OHP-4022-016-004, Loss of Component Cooling Water, is in progress. Which ONE of the following describes the procedural requirements for CCP operation based on these conditions? a. Immediately stop both CCPs. b. Immediately stop one CCP; stop the second CCP within 1-1/2 minutes of the event. c. Stop BOTH CCPs within 1-1/2 minutes of the event. d. Immediately stop one CCP; run the second CCP as long as it continues to operate.

d. CORRECT. 02-OHP-4022-016-004 has a note prior to step 4 that describes the possible damage that may occur to a CCP on the loss of CCW. The note and procedure directs that one CCP be saved until CCW is restored. The other pump should be run as long as possible to allow time to align CVCS crosstie

Given the following plant conditions: • The plant is in Mode 3 performing a cooldown in preparation for a refueling outage. • A malfunction of the Steam Generator Power Operated Relief Valves causes the cooldown rate to exceed Technical Specification limits. Which ONE of the following actions is required per TECHNICAL SPECIFICATIONS and why? a. Restore cooldown rate to Technical Specifications limits within 1 hour to provide adequate margin from ductile failure of the reactor vessel. b. Immediately stop any further cooldown. Maintain temperature for 6 hours to allow temperature stabilization throughout the reactor vessel wall. c. Stop cooldown within 15 minutes. Maintain temperature for 12 hours to allow temperature stabilization throughout the reactor vessel wall. d. Restore cooldown rate to Technical Specifications limits within 30 minutes to provide adequate margin from brittle failure of the reactor vessel.

d. CORRECT. Technical Specification 3.4.3 requires the RCS temperature to be restored to within Limits in 30 minutes. The concern of excessive cooldown rates to brittle failure caused by the tensile stresses on the inner wall.

Given the following plant conditions on Unit 2: • The unit is operating at 100% power. • A small instrument air leak inside Containment causes a slow rise in Containment pressure. • Containment pressure is currently 0.19 psig. In order to ensure that adequate margin to Containment Technical Specification pressure limits is maintained, which ONE of the following indicates the appropriate action to reduce Containment pressure ? a. Maximize NESW cooling to the Containment Ventilation Units b. Lower pressure in containment using the Containment Purge System. c. Vent containment using the Containment Pressure Relief system. d. Verify all Upper/Lower Containment Ventilation Fans (CUV/CLV) are running.

c. CORRECT. With the Containment Pressure rising due to air line leakage, the only way to reduce pressure is to purge air from Containment. This is accomplished with the Containment Pressure Relief System.

Unit 2 is at 76% power with all control systems in AUTOMATIC. Which ONE of the following describes the plant response to a trip of the East Main Feed Pump? Note: Assume no operator action. a. 1. The Turbine will NOT automatically runback causing SG water levels to start lowering resulting in the Feedwater Regulating Valves opening further. 2. A lower feedwater header pressure causes the West MFP speed to rise until it trips on overspeed. 3. Both the Steam-driven and Motor-driven Auxiliary Feedwater Pumps start when SG levels reach the Low-Low level setpoint. b. 1. The Turbine will automatically runback. 2. The SG pressures will rise due to the runback requiring the West MFP speed to rise. 3. As the MFP speed rises, the MFP Suction Pressure will lower causing the Low Pressure Heater Bypass CRV-224 to open and the Middle Heater Drain Pump to start. c. 1. The Turbine will NOT automatically runback. 2. The West MFP transfers to Speed Control with the Speed set to Maximum but will not maintain SG levels. 3. Both the Steam-driven and Motor-driven Auxiliary Feedwater Pumps start when SG levels reach the Low-Low level setpoint. d. 1. The Turbine will automatically runback. 2. The West MFP transfers to Speed Control with the Speed set to Maximum. 3. NO automatic pump starts occur.

d. CORRECT. On the loss of the East Main FW Pump a turbine runback is automatically initiated, reduced flow will cause the FW regulating valves will open further as the SGs try to maintain normal level & FW flow matched to steam flow. FW Pump transfers to Speed Control with the Speed set to Maximum.

Unit 2 is operating at 100% Power when the following alarm actuates: • Ann. 210, Drop 50, REACTOR PROT TRAIN B TROUBLE. Investigation reveals the Train B General Warning Lamp Lit and one of the 48 VDC power supplies on Train B tripped. Which ONE of the following describes the impact, if any, of this failure? a. The Train B Reactor Trip breaker Shunt trip is unavailable. b. NONE of Train B SSPS equipment will automatically actuate when required. c. A portion of the Train B SSPS equipment will NOT automatically actuate when required. d. ALL of Train B SSPS equipment will automatically actuate when required.

d. CORRECT. The 48 VDC power supplies are 100% redundant within the SSPS train. The RPS/SSPS will still function as designed but the redundancy is compromised.

Given the following plant conditions: • Unit 2 was operating at 100% power. • The Unit 2 North NESW Pump was running. • The Unit 2 South NESW Pump was stopped and in AUTO. A large steamline break inside containment occurred. • Unit 2 offsite power is lost on the trip. • Both Unit 2 Diesel Generators started and are supplying their respective busses. • Unit 2 Containment Pressure is 3.2 psig. • UNIT 1 has remained operating at 100% power throughout the event. Which ONE of the following describes the response of the NESW pumps and the required actions; if any, to restore Unit 2 NESW? a. The Safety Injection Sequencer will start both Unit 2 NESW pumps after the appropriate time delay. b. The Safety Injection Sequencer will start the North NESW Pump after the appropriate time delay. The South NESW Pump will start on Low System Pressure if the North NESW Pump fails to restart. c. NESW Load Conservation will prevent the start of both Unit 2 NESW pumps. Either pump may be started by placing the control switch directly to Run after the Containment Spray Pumps have been stopped. d. NESW Load Conservation will prevent the start of both Unit 2 NESW pumps. Either pump may be started by placing the control switch to Trip/Lockout and then to Run after at least 75 seconds have elapsed since the time of the Containment Spray Signal.

d. CORRECT. NESW Load Conservation actuates on a CTS with Load Shed (or EDG on Bus). This prevents the NESW pumps from starting for 75 seconds or until the CTS signal is cleared. The Control Switch will need to be placed in Trip/Lockout to clear the signal.

Given the following plant conditions on Unit 1: • The unit was operating at 100% power when an INADVERTENT Phase A Containment Isolation occurred on Train A. • The Crew has reset Phase A Containment Isolation and attempted to restore Control Air to Containment. • The Control Air Containment Isolation Valves could not be opened. Which ONE of the following describes short-term impact of the loss of air on the restoration efforts of the crew and the required compensatory actions? a. RCP NESW Motor Air cooling water can NOT be restored. Trip the reactor and stop 3 RCPs. Perform a containment pressure relief. b. Glycol Cooling to the ice condenser can NOT be restored. Stop all Unit 1 Air Handling Units (AHUs). Monitor ice bed temperatures to ensure they remain at an acceptable level. c. RCS overpressure protection has been lost (PORVs will NOT open). Begin a reactor shutdown and be in Mode 3 within 6 hours. d. RCP Seal Injection is available but Seal Return can NOT be restored. Drain the PRT as required to maintain an acceptable level.

b. CORRECT. Glycol Cooling inside Containment Isolation valves VCR-11 and VCR-21 will NOT open. The AHU's are stopped if the glycol system is shutdown for more than 30 minutes. Technical Specifications requires that temperatures are maintained < 27°F.

Given the following plant conditions: • Reactor Power is 95%. • The East CCP pump trips. • The West CCP is in Standby. Which ONE of the following describes the effect of these conditions on the CVCS System operation assuming NO operation action? a. QRV-251 (CCP Discharge Flow Control valve) throttles to minimum flow. b. QRV-160, 161, and 162 (Letdown Orifice Isolation valves) fully close. c. CRV-470 (Letdown Temperature Control valve) throttles in the open direction. d. QRV-111 and 112 (RC Letdown to Regen Hx valves) fully close.

b. CORRECT. The Letdown Orifice Isolation valves are directly interlocked to close (isolates letdown) on a loss of the CCP.

Given the following plant conditions: • Unit 2 is at 55% • Both Main Feedwater Pumps are in operation. Which ONE of the following conditions will DIRECTLY cause a trip of the East Main Feedwater Pump (MFP)? a. Main Turbine Trip b. Loss of MFP Turbine Condenser Vacuum c. East MFP Turbine Seal Water Delta-P Low d. East MFP Turbine Thrust Bearing Temperature High

b. CORRECT. Condenser vacuum of <15.0 in Hg will cause a Triconix Trip of the operating MFP.

``` A Safety Injection (SI) occurred 20 minutes ago due to a large break LOCA, with a failure of reactor trip breaker "B" to open. Which ONE of the following describes the SI signal response after pushing (then releasing) both SI reset pushbuttons? SI STATUS Train A Train B a. Reset NOT Reset b. Reset Reset c. NOT Reset NOT Reset d. NOT Reset Reset ```

a. CORRECT. With a failure of Train B reactor Trip Breaker to open, a P-4 signal is not generated on Train B. With a Large Break LOCA the SI signal will still be present preventing Train B from being reset. The SI reset and P-4 block features are train specific.

``` Which ONE of the following describes the effect on the reactor trip breakers of actuating the manual REACTOR TRIP switches? Shunt trip relays de-energize, causing the reactor trip breaker trip coils to (1) , AND Reactor trip breaker undervoltage coils (2) . a. 1) energize 2) energize b. 1) de-energize 2) de-energize c. 1) energize 2) de-energize d. 1) de-energize 2) energize ```

c. CORRECT. Manual trip de-energizes the undervoltage trip coil and energizes the reactor trip breaker trip coil.

``` Which ONE of the following describes the Technical Specification Limits and the ESFAS response associated with the UPPER Containment Ventilation fans? • 1) Upper Containment temperature limits - 2) Fans Trip from a. 1) >60°F and <120°F 2) Phase A actuation b. 1) >60°F and <120°F 2) Phase B actuation c. 1) >60°F and <100°F 2) Phase A actuation d. 1) >60°F and <100°F 2) Phase B actuation ```

c. CORRECT. Technical Specifications Limits upper Containment to >60°F and <100°F. The fans trip on a Phase A Actuation.

Unit 2 was operating at 95% power. Following an Ice Condenser Inspection the AEO reports that: • One Intermediate Deck Door was slightly open and incapable of being fully closed due to Ice Buildup. • Another Intermediate Deck Door has excessive Ice Buildup around the hinges and could NOT be moved. Which ONE of the following actions, if any, are required to continue operations at power? Note: Technical Specifications 3.6.11 and 3.6.12 are attached. a. No action is required since the Ice Condenser Inlet Doors are all capable of opening. b. Monitor the ice bed temperature less than or equal to 27°F every 4 hours for a maximum of 14 days. c. Restore the Ice Bed to Operable status within 48 hours or be in Mode 3 within 6 hours and Mode 5 within 36 hours. d. Restore the Doors to Operable status within 1 hour or be in Mode 3 within 6 hours and Mode 5 within 36 hours.

b. CORRECT. If the Ice Condenser Intermediate deck doors are inoperable then TS 3.6.12 Action B applies. This requires temperature to be monitored every 4 hours for up to 14 days. The TS allows seperate entry for each Door.

Given the following plant conditions on Unit 2: • A tube rupture has occurred in the SG#23. • A Manual Reactor Trip and Safety Injection were performed. • RCS Pressure is currently 2035 psig. • RCS Tave is 549°F. • SG#23 has been isolated in accordance with procedure 2-OHP-4023-E-3, Steam Generator Tube Rupture. The Unit Supervisor has directed you to perform the cooldown to 475°F using the Steam Dumps. Which ONE of the following describes how the cooldown would be performed and the reason? a. Place the Steam Dumps in Steam Pressure mode and fully open them to obtain a cooldown at the maximum rate. The Steamline Break SI, Main Steam Isolation will NOT occur since SI is already actuated b. Place the Steam Dumps in Steam Pressure mode and open them in a controlled manner to prevent causing a Steamline Isolation. Block the Steamline Break SI, Main Steam Isolation as temperature lowers. c. Block the Steamline Break SI, Main Steam Isolation and then place the Steam Dumps in Steam Pressure mode and fully open them to obtain a cooldown at the maximum rate. d. Block the Steamline Break SI, Main Steam Isolation and then place the Steam Dumps in Steam Pressure mode and open them in a controlled manner to prevent exceeding steam flows of 1.42 x106 pph and causing a Steamline Isolation.

b. CORRECT. The note before step 7 warns that a steam line isolation may occur if the dumps are opened rapidly. The dumps should be opened in a controlled manner and to obtain the maximum cooldown rate and then the Steam Line SI can be blocked when P-12 (541°F) is reached.

Both Units are responding to a loss of Plant Air (PA) and Control Air (CA) event. Given the following plant conditions: • Both Unit's Plant Air Compressors (PACs) have failed. • Both Unit's Control Air Compressors (CACs) have failed. • Both Unit's PA Header pressures are 70 psig and lowering. • Both Unit's CA Header pressures are 80 psig and lowering. • An AEO has been dispatched to start and align the Back-up Plant Air Compressor (B/U PAC), per 1-OHP-4022-064-001, Control Air Malfunction. Which ONE of the following statements describes the B/U PAC ability to repressurize both Unit's Control Air Systems under these conditions? a. PRV-10 or PRV-11, Plant Air Header Crosstie Valves to Unit 2, must be jumpered and reopened. b. PRV-20 or PRV-21, Plant Air Header Crosstie Valves to Unit 1, must be jumpered and reopened. c. The B/U PAC will supply air to both Unit's Control Air Headers. d. The B/U PAC will supply air to both Unit's Plant Air Receivers.

c. CORRECT. The Backup Air compressor discharges into the section of piping between the Plant Air header crosstie valves. This is also where the Control Air headers are connected allowing the backup air compressor to supply control air.

Which ONE of the following describes the power supply for the Unit 1 West CCP during normal plant operation? a. Bus T11A b. Bus T11B c. Bus T11C d. Bus T11D

a. CORRECT. The West CCP is a Train B pump which is supplied by the T11A Bus.

Unit 2 was operating at 100% power when main condenser vacuum dropped from 28 inches vacuum to 25 inches vacuum and stabilized. Given the following plant conditions: • All 4 condenser steam air ejectors (SJAE) are in service. • The system lineup is in the normal configuration. Which ONE of the following conditions describes the cause of this problem? a. Loss of a running Hotwell Pump's seal water supply. b. Loss of the Main Steam supply to the SJAEs. c. Loss of Condensate flow to the SJAE condensers. d. A SJAE condenser drain was left open.

c. CORRECT. Removes the ability of SJAEs to remove noncondensibles and causes vacuum to lower.

Rod Control was in AUTO with Unit 1 power at 79% when MPC-253, Turbine Impulse Pressure Channel 1, failed low. Which ONE of the following describes the expected response of the rod control system? a. Rods will initially move in due to power mismatch. AUTO rod control will then withdraw rods to restore temperature. b. Rods will move in due to power mismatch and due to a temperature error until placed in MANUAL. AUTO rod control may be restored after MPC-253 has been defeated and the associated bistables have been tripped. c. Rods will initially move in due to power mismatch but will stop as the temperature error becomes large enough. AUTO rod control is NOT available until after Conditional C-5 bistable has been tripped. d. Rods will move in due to power mismatch and due to a temperature error until placed in MANUAL. AUTO rod control is NOT available even after MPC-253 bistables have been tripped.

d. CORRECT. The failure of MPC-253 would have caused the control Rods to Insert (Tref lowered) since it feeds Tref. MPC-253 also feed the 15% C-5 interlock which prevents Auto Rod Withdrawal. Since the Tref function is not selectable - Auto Rod Control is not available.

The In-Service Waste Gas Decay Tank #2 has a HYDROGEN concentration of 4.8%. Per TRM 8.7.12, Explosive Gas Mixture, which ONE of the following is the HIGHEST OXYGEN concentration allowed in Gas Decay Tank #2? a. 2% b. 3% c. 4% d. 5%

b. CORRECT. TRM 8.7.12, Explosive Gas Mixture, TRO 8.7.12, requires Oxygen Concentration of less than or equal to 3% when Hydrogen concentration is > 4%

Given the following conditions in Unit 1: • The unit is at 20% power with a power ascension in progress. • Rods are in Manual • RCP 11 trips due to an overcurrent condition. • No operator action has been taken. Which ONE of the following describes the INITIAL reactor and the UNAFFECTED Loops response? a. A reactor trip WILL NOT occur and the UNAFFECTED Loops Delta-T will lower. b. A reactor trip WILL NOT occur and the UNAFFECTED Loops Delta-T will rise. c. A reactor trip WILL occur and the UNAFFECTED Loops Delta-T will lower. d. A reactor trip WILL occur and the UNAFFECTED Loops Delta-T will rise.

b. CORRECT. A single RCP trip will not cause an automatic reactor trip below P-8. Once the RCP trips, the unaffected loops Delta-T rises to continue producing the same power level as before.

Unit 1 is conducting a reactor startup following a refueling outage. Given the following plant conditions: • Source Range and Intermediate Range Nuclear Instruments are slowly rising. • Rods are in manual with no rod motion. • Source Range Instruments read as follows: • N-31: 2.1x104 cps. • N-32: 2.0x104 cps. • Intermediate Range Instruments read as follows: • N-35: 2.5x10-11 amps. • N-36: 1.0x10-9 amps. Which ONE of the following best explains the indications? a. N-35 compensating voltage is set too high b. N-35 compensating voltage is set too low c. N-36 compensating voltage is set too high d. N-36 compensating voltage is set too low

a. CORRECT. N-35 reads too low for the conditions given, compensating voltage is too high.

The control room operators are performing 1-OHP-4023-FR-C.1, Inadequate Core Cooling . They are NOT able to establish high head ECCS flow. Given the following plant conditions: • SG depressurization proves to be ineffective. • SG NR levels are stable at 20%. • All core exit TCs are 1250°F and slowly rising. The operators were attempting to establish conditions for RCP restart, but are unable to establish RCP seal injection or 200 psid across the #1 seal. Which ONE of the following describes the current status of the fuel and the required actions? a. The fuel is NOT significantly damaged. The crew is required start one RCP at a time until core exit TCs are less than 1200°F. b. The fuel is significantly damaged so the crew should NOT start the RCPs. They are required to open all PRZ PORVs and block valves. c. The fuel is significantly damaged. The crew is required start all RCPs simultaneously to reduce core exit TCs to less than 1200°F. d. The fuel is NOT significantly damaged so the crew should NOT start the RCPs. They are required to continue attempts to establish high head injection.

a. CORRECT. The 1200°F value was chosen because it was significantly below the point at which the fuel was damaged, but high enough that extreme measures are required to recover cooling. Since adequate SG levels for heat sink exist, the RCPs are started in an attempt to circulate coolant/steam through the SG tubes in an attempt to cool the RCS. Adequate support condition for the RCPs are desired but NOT required since Core Cooling is severely challenged.

Given the following plant conditions on Unit 1: • The unit is at 100% power and stable. • Steam Generator Level Controls are in AUTOMATIC. • Steam Generator #12 Feed Flow Channel 1, 1-FFC-220, is selected to the Steam Generator Level Control System. If 1-FFC-220 instantaneously fails offscale low, which ONE of the following describes the expected plant response? Note: Assume no operator action. The Steam Generator Level Control system will: a. initially lower feed flow and then slowly return SG#12 level to approximately program level. b. automatically transfer the SG#12 FW Regulating Valve Controller to Manual to maintain the current valve position. c. initially raise feed flow and then slowly return SG#12 level to approximately program level. d. automatically transfer the SG#12 FW Regulating Valve Controller to Feed Flow Channel 2, 1-FFC-221 to allow continued automatic operation.

b. CORRECT. A sudden failure offscale will cause the Taylor Controller to shift to manual at the current position.

Unit 2 was operating at 100% power when a fire occurred in the Unit 2 Charging Pump Area resulting in a loss of seal injection to all the RCPs. The Shift Manager determines that a Reactor Trip is required based on the fire impacting control of the plant from the Control Room. Which ONE of the following describes the correct Operator response? Immediately trip the Reactor and RCPs and implement: a. 2-OHP-4023-E-0, Reactor Trip or Safety Injection. 12-OHP-4025-001-002, Fire Response Guidelines, for CCP area may be performed concurrently after the immediate actions are complete. b. 2-OHP-4023-E-0, Reactor Trip or Safety Injection. 12-OHP-4025-001-002, Fire Response Guidelines, for CCP area is NOT needed since the EOP network addresses a loss of seal injection due to fire. c. 2-OHP-4023-E-0, Reactor Trip or Safety Injection. Steps from 12-OHP-4025-001-002, Fire Response Guidelines, for CCP area may NOT be performed until completion of 2-OHP-4023-ES-0.1, Reactor Trip Response. d. 12-OHP-4025-001-002, Fire Response Guidelines, for CCP area until restoration of seal injection. Perform 2-OHP-4023-E-0, Reactor Trip or Safety Injection, steps as time allows.

a. CORRECT. OHI-4023, Abnormal/Emergency Procedure User's Guide allows Abnormal Procedures to be implemented concurrently with Emergency Procedures.

Given the following plant conditions on Unit 2: • The unit is in Mode 6 with refueling activities in progress. • Containment purge is in service. • A fuel element accidentally dropped into the cavity. • All radiation monitor TRIP/BLOCK switches are in their NORMAL position. • The Manipulator Crane area radiation monitor has a HIGH alarm. • ERS-2305 and ERS-2405, Lower Containment Noble Gas Low Range Radiation Monitors, have a HIGH alarm. Which ONE of the following actions would occur, assuming that operators follow the required actions of 12-OHP-4022-018-004, Irradiated Fuel Handling Accident In Containment Building - Control Room Actions and all equipment responds as designed? a. Containment evacuation alarm sounds automatically. Containment purge stops automatically. b. Containment evacuation alarm is manually actuated by the control room operator. Containment purge stops automatically. c. Containment evacuation alarm sounds automatically. Containment purge is stopped manually by the control room operator. d. Containment evacuation alarm is manually actuated by the control room operator. Containment purge is stopped manually by the control room operator.

b. CORRECT. OHP 4022.018.004, Step 1 directs the operator to actuate the Containment Evacuation alarm. For the protection of personnel, it is important to evacuate the affected area until radiation surveys can be completed. Step 2 follows up the alarm with a page announcement notifying all non-essential personnel to evacuate the containment. Steps 5 through 9 verify the containment purge and pressure relief systems are shutdown and isolated. This will limit the exposure of personnel outside containment.

Given the following plant conditions on Unit 1: • The unit is operating at 100%. • The feedwater temperature input to the LEFM thermal power calculation was incorrectly calibrated to 7oF higher than actual feedwater temperature. • Calibration of the power range nuclear instruments (NIs) is being performed. How will LEFM power compare to actual thermal power and how will adjustment of the Nis be affected using the calculated value of LEFM? a. Calculated thermal power is lower than actual power. NI adjustment will be less conservative. b. Calculated thermal power is higher than actual power. NI adjustment will be less conservative. c. Calculated thermal power is lower than actual power. NI adjustment will be more conservative. d. Calculated thermal power is higher than actual power. NI adjustment will be more conservative.

a. CORRECT. Due to actual FW temperature being 7 degrees lower than indicated/calibrated FW temperature, calculated thermal power will indicate lower than actual thermal power. This will cause NI calibration to result in setpoints that are less conservative (i.e. farther from the trip setpoint).

``` Refer to the following list of valve operations: 1. Close discharge valve. 2. Close suction valve. 3. Open discharge valve. 4. Open suction valve. Which ONE of the following describes the required sequence of valve operations when tagging out and subsequently restoring to service of a centrifugal pump? TAGOUT then RESTORE a. 1,2 then 3,4 b. 1,2 then 4,3 c. 2,1 then 3,4 d. 2,1 then 4,3 ```

b. CORRECT. Close discharge before suction, and open suction before discharge.

A maintenance visual inspection requires momentarily placing a 'B' train pump control switch in PULL-TO-LOCKOUT then back to AUTO. The unit condition is such that BOTH trains are required to auto start. Which ONE of the following describes the status of the affected ESF system? The 'B' train pump is INOPERABLE until: a. the control switch is independently verified in the AUTO position. b. the pump's monthly surveillance has been performed. c. the pump's auto start function is tested. d. the pump is manually started.

a. CORRECT. The B train pump may be considered Operable after being returned to the correct position and being independently verified.

2008 NRC Exam - Question # 70 Given the following plant conditions on Unit 2: • A normal plant cooldown is in progress: • RCS loop Tavg readings are all lowering with the current values: • Loop 1: 538°F. • Loop 2: 542°F. • Loop 3: 537°F. • Loop 4: 540°F. • Steam header pressure is 900 psig and lowering. • Steam Dump Mode Selector switch in STM PRESS MODE. • Steam Dump Controller in MAN and set at 40% demand. • ALL Steam Dumps are closed. The operator momentarily places the Steam Dump Control Selector Train A and Control Selector Train B switches to BYPASS INTRLK and then releases them. What is the expected status of the Steam Dump valves following the operator's actions? a. All valves remain closed. b. The valves in group 1 are open and the valves in groups 2 and 3 are closed. c. The valves in group 1 and 2 are open and the valves in group 3 are closed. d. The valves in group 1 and 2 are open and the valves in group 3 are partially open.

b. CORRECT. Below P-12 (2/4 Tavg less than 541 deg F) only the 3-Group 1 dump valves are available for service.

The control room operators are responding to a Steam Generator Tube Rupture. In order to cool down the RCS and establish required subcooling margin, the operators dump steam to the condenser using the intact SGs. Which ONE of the following describes why this method of RCS cooldown is preferred over dumping steam through the PORVs of the intact SGs? a. Minimizes radiological releases. b. Minimizes thermal shock to the reactor vessel. c. Minimizes shrink experienced by the RCS. d. Minimizes RCS subcooling requirements.

a. CORRECT. Dumping steam to the condenser will keep the radioactivity contained within the plant. Prior to the ruptured SG isolation, the leaking RCS would have mixed with the secondary and raised the activity levels of the intact SGs. Distracters are plausible since the SG PORVs may provide greater cooling at maximum rate and are typically less controllable than the fine control provided by the steam dumps.

Given the following plant conditions on Unit 1: • Containment Purge System is operating in the CLEAN-UP MODE. • An external failure alarm on VRS-1101, Upper Containment Normal Range Area Monitor, occurs. Which ONE of the following describes the response of the Containment Ventilation System to the failure alarm? a. Containment ventilation isolation valves 1-VCR-101 through 1-VCR-107 close 1-HV-CIPS-1, Containment Instrument Room Purge Supply Fan, trips b. Containment ventilation isolation valves 1-VCR-101 through 1-VCR-107 close 1-HV-CPS-1/2, Containment Purge Supply Fans 1 and 2, trip 1-HV-CPX-1/2, Containment Purge Exhaust Fans 1 and 2, trip 1-HV-CPR-1, Containment Pressure Relief Fan, trips 1-HV-CIPS-1, Containment Instrument Room Purge Supply Fan, trips c. Containment ventilation isolation valves 1-VCR-201 through 1-VCR-207 close 1-HV-CIPS-1, Containment Instrument Room Purge Supply Fan, trips d. Containment ventilation isolation valves 1-VCR-201 through 1-VCR-207 close 1-HV-CPS-1/2, Containment Purge Supply Fans 1 and 2, trip 1-HV-CPX-1/2, Containment Purge Exhaust Fans 1 and 2, trip 1-HV-CPR-1, Containment Pressure Relief Fan, trips 1-HV-CIPX-1, Containment Instrument Room Purge Exhaust Fan, trips

a. CORRECT. While Operating in the Clean-UP mode the Radiation Monitor switches are unblocked allowing actuations. VRS-1101 closes the Inside Containment Isolation valves and trips the Instrument Room Purge Supply fan only.

Given the following plant conditions on Unit 1: • The unit has experienced a loss of offsite power. • You are an extra operator and are assigned to address the annunciators on Panel 119, Station Auxiliary AB. Which ONE of the following would you address FIRST based on the OHI-4000 Annunciator Priority system? The Annunciator with: a. a Red "C" on the lens. b. a Red lens with a Purple slash. c. a White lens with a slash in the lower right corner. d. an Orange dot on the lens.

b. CORRECT. The red lens is a first priority annunciator. Adding the purple slash means a condition that would not allow the EDG to start or deliver power.

Given the following plant conditions: • Plant is at 75% power. • RCS boron concentration is 480 ppm. • Control Bank D is at 200 steps. • Control Rod Bank Selector is in AUTO. • Turbine load is being raised by 200 MWe. • Ann. 210, Drop 26, ROD CONTROL URGENT FAILURE is LIT. • Rod motion stops. • The AEO reports the URGENT FAILURE light on the 1BD Power Cabinet is LIT. What action is taken for this condition? a. Continue the power escalation using dilution. b. Manually trip the reactor and go to 2-OHP-4023-E-0, Reactor Trip or Safety Injection. c. Stop the power escalation and maintain Tave using Turbine load reduction and/or boration. d. Place rods in MANUAL and reduce reactor power below 50% within 4 hours using Turbine Load Reduction and boration.

c. CORRECT. The power escalation should be stopped to investigate the plant conditions. OHP-4022-012-001, Failure of A Control Bank to Move Directs using a Turbine load reduction or Boron Adjustment to stabilize the plant.

A reactor trip with a safety injection occurred due to a feed line break on SG#22. The crew is performing actions of 2-OHP-4023-E-0, Reactor Trip or Safety Injection. Steam Generator Aux Feedwater Flows were indicating as follows: SG21 SG22 SG23 SG24 Flow FFI-210 FFI-220 FFI-230 FFI-240 Flow 200x103 Pegged High 200x103 200x103 • Ann. 214, Drop 9, TDAFP DISCHARGE FLOW HIGH, has just alarmed. Which ONE of the following responses are correct given these conditions? a. Do NOT trip TDAFP. Manually throttle AFW Flow to ALL 4 SGs until the Discharge Flow High Alarm clears. This alarm indicates that Aux Feed Flow Retention has FAILED. b. Do NOT trip TDAFP. Verify that AFW Flow to ALL 4 SGs has automatically throttled as expected for Aux Feed Flow Retention.This alarm indicates that Aux Feed Flow Retention has ACTUATED. c. Trip TDAFP. This alarm indicates that the feed line break is on the Aux Feed Line. d. Trip TDAFP. This alarm indicates that Aux Feed Flow Retention has FAILED.

b. CORRECT. Upon High AFP flow to a SG (>975 gpm) the flow retention circuit will throttle the AFP valves closed to prevent pump runout. This is an expected alarm given these conditions. The pump should continue to operate after verifying that flow retention is properly operating.

Given the following plant conditions on Unit 2: • Unit is in MODE 4 cooling down on RHR. • RCS Temperature - 340°F. • RCS pressure - 300 psig lowering. • Pressurizer level - 22% lowering. • Containment pressure - 0.1 psig. • VRS-2505, U2 Vent Noble Gas is in Alarm. • VRS-2503, U2 Vent Iodine Monitor radiation levels are trending higher. • SG levels stable at - 42% (21); 40% (22); 43% (23); 40% (24). Which ONE of the following identifies the problem and the associated action? a. A LOCA has occurred on the suction of the RHR pump. Enter 2-OHP-4022-002-015, Mode 4 LOCA. b. LTOP (Low Temperature Over Pressure) actuated and one PORV is stuck open. Enter 2-OHP-4022-002-015, Mode 4 LOCA. c. Letdown line pressure relief valve has failed open. Enter 2-OHP-4022-002-020, Excessive Reactor Coolant Leakage d. A LOCA has occurred in the area of the Regenerative Heat Exchanger. Enter 2-OHP-4022-002-020, Excessive Reactor Coolant Leakage

a. CORRECT. During any LOCA, RCS pressure and inventory will fall. Rising indication on the Aux Bldg. Radiation monitors is indicative of the LOCA outside containment (RHR pump suction).

Unit 2 was operating at 100% power when indications of a lowering CCW Surge tank required entry into 2-OHP-4022-016-001, Malfunction of the CCW System. The Crew has started the West CCW pump, split the East and West Headers aligning the Miscellaneous Services Header to the East Header. The following Surge Tank Level Recorder conditions exist: CLR-410 CLR-411 Reading 18" 48" Trend Lowering Stabl An AEO reports that a CCW leak of approximately 150 gpm has been identified near the South Spent Fuel Pool Heat Exchanger. Which ONE of the following describes the leak location and the required actions? The leak is located on the: a. Miscellaneous Services Header. Trip the Reactor, Stop both CCW Pumps, and Implement 2-OHP-4022-016-004, Loss of CCW along with 2-OHP-4023-E-0, Reactor Trip or Safety Injection. b. East Safeguards Header. Shutdown the East CCW pump and align the Miscellaneous Services Header to the West Safeguards Header. c. West Safeguards Header. Shutdown the West CCW pump and the equipment cooled by the West Header. d. Miscellaneous Services Header. Trip the Reactor, Trip the RCPs, and isolate the Miscellaneous Services Header while performing 2-OHP-4023-E-0, Reactor Trip or Safety Injection.

d. CORRECT. During the initial train split, the Misc. Header is aligned to the East Safeguards Header. The SFP cooling is supplied from the Misc. Header. Based on the Leak rate isolation of the Misc. Header requires that the Reactor be tripped and the RCPs stopped since they lose cooling.

Given the following plant conditions on Unit 2: • Reactor power: 62% and rising. • RCS pressure: 2220 PSIG and lowering. • Auctioneered High Tavg: 560°F and lowering. • Turbine power: 520 MWe and lowering. Based on the above plant indications, what event is occurring and what are the required actions/procedures to address the event? a. A Steamline Break requires a reactor trip and 2-OHP-4023-E-0, Reactor Trip or Safety Injection. b. A Steamline Leak requires a rapid power reduction per 2-OHP-4022-001-006, Rapid Power Reduction Response. c. A Small Break RCS LOCA requires a reactor trip and 2-OHP-4023-E-0, Reactor Trip or Safety Injection. d. An RCS Leak requires implementation of 2-OHP-4022-002-020, Excessive Reactor Coolant Leakage.

a. CORRECT. Reactor power is rising, indicating positive reactivity event. Electric load is lowering, indicating loss of steam to the turbine. Turbine power should be closer to 700 Mw with Tavg closer to 563°F based on this reactor power. Based on this degree of mismatch a reactor trip is required (due to the size of the steam break).

You are the Unit Supervisor in Unit 2. The unit was operating at steady state full power when a Loss of Offsite Power occurred. The following indications were observed during the performance of Step 1 of 2-OHP-4023-E-0, Reactor Trip or Safety Injection: • WR Neutron flux is 10% and slowly lowering. • BYA and RTB are closed. • RTA and BYB are open. • ALL Auxiliary Feedwater Pumps are Running. The above indications remained constant after the RO actuated the manual reactor trip breaker switches. Which ONE of the following describes the action required and the associated reason? a. Go to 2-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS. WR Neutron flux at 10% indicates that the heat load from the RCS may be higher than that assumed for the safeguard systems protection during an accident. b. Continue in 2-OHP-4023-E-0, Reactor Trip or Safety Injection. The WR Neutron flux lowering with the AFW Pumps running confirm that the RCS heat load is low enough to meet the safeguard systems design. c. Go to 2-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS. The Reactor is Shutdown but since Rod Position Indication is NOT available, additional actions may be required to ensure that Shutdown Margin can be maintained. d. Go to 2-OHP-4023-FR-S.2, Response to Loss of Core Shutdown. The Reactor is Shutdown but since Rod Position Indication is NOT available, additional actions may be required to ensure that Shutdown Margin can be maintained.

a. CORRECT. FR-S.1 is used since WR Flux is not less than 5% and this may exceed the assumptions used for the safeguards.

Unit 2 was at 60% power when the following conditions are noted: • Ann. 219, Drop 29, CRID 3 INVERTER ABNORMAL is alarming. • Ann. 219, Drop 30, CRID 4 INVERTER ABNORMAL is alarming. • SG Feed Flows are Stable. A scan of the NIS drawers indicates the following: • Source Range N-31 and N-32 at zero. • Source Range N-21 and N-23 near full scale. • Intermediate Range N-35 and N-36 near full scale. • Power Range N-41, N-42 and N-44 at mid-scale. • Power Range N-43 at zero. Which ONE of the following describes this failure and what actions/procedures are required? a. CRID 3 & 4 have transferred to the Alternate Power Supply. Battery 2AB has failed, enter 2-OHP-4022-082-002A, Loss of Power to 250VDC Bus 2AB ONLY. b. CRID 4 has transferred to the Alternate Power Supply. CRID 3 has Failed to Transfer. Refer to 2-OHP-4021-082-008, Operation of CRID Power Supplies. c. CRID 3 has transferred to the Alternate Power Supply. CRID 4 has Failed to Transfer. Refer to 2-OHP-4021-082-008, Operation of CRID Power Supplies. d. CRID 3 & 4 have failed to transfer to the Alternate Power Supply. Initiate a Reactor Trip and enter 2-OHP-4023-E-0, Reactor Trip or Safety Injection.

b. CORRECT. The Failure of N-43 indicates that CRID 3 is lost. N-23 is powered from Unit 1.

Unit 2 was performing a normal reactor shutdown to enter a refueling outage. Power was stabilized and a manual reactor trip was performed to shutdown the reactor in accordance with 2-OHP-4021-001-003 Power Reduction. The Turbine Driven AFW pump automatically started following the reactor trip due to SG levels. Which ONE of the following describes the MOST LIMITING NRC reportability requirement, if any, for this event? Note: PMP-7030-001-001 Prompt NRC Notification is attached. a. 1 hour notification b. 4 hour notification c. 8 hour notification d. No reportability required

c. CORRECT. 8 hour report is the most limiting notification required per 50.72(b)(3)(iv)(B).

The plant was at 100% power when a Control Room fire caused a Loss of Offsite Power and forced the evacuation of the Control Room. A cooldown at 14°F/hr is in progress. The operator has been dispatched to initiate Process Monitoring from the LSI Panels using 1-OHP-4025-LS-1-1. Startup LSI Panels and Process Monitoring. The current conditions exist: • RCS Thot is 400°F. • RCS Tcold is 365°F. • RCS pressure is 1700 psig. Which ONE of the following actions should the Local SRO direct the local operator at SG LSI panels to perform? Note: 1-OHP-4025-LS-1-1, Startup LSI Panels and Process Monitoring is attached. a. No action is required since the 50°F subcooling margin is satisfied. b. No action is required since RCS Tcold is < 390°F. c. Depressurize RCS to minimize excess subcooling above the 50°F requirement. d. Cool down RCS Thot to establish a minimum of 220°F subcooling margin.

d. CORRECT. The required subcooling margin is 220°F when on natural circulation. Subcooling is based on RCS pressure and Hot Leg temps.

Step 2 of 2-OHP-4023-FR-C.3, Response to Saturated Core Conditions, checks if the RHR system has been placed in service in the shutdown cooling mode. Which ONE of the following describes the basis for this step? a. To ensure an ORANGE or RED condition in Core Cooling will not arise while performing this procedure. b. If a problem exisits with RHR shutdown cooling, then a transition to 2-OHP-4022-017-001, Loss of RHR Cooling, is required to address the problem. c. If RHR is in service in the shutdown cooling mode, then a transition to 2-OHP-4021-017-001, Operation of the RHR System, is required to verify RHR is properly aligned. d. If a problem exisits with RHR shutdown cooling, then a transition to the steps within this procedure is required to identify and isolate the affected train.

b. CORRECT. The FRPs do NOT address the RHR Shutdown cooling malfunctions, the check is made to direct a transition if RHR is in Shutdown Cooling.

A Unit 2 power ascension was in progress following a maintenance outage. While raising power from 90% to 100% power, several alarms were received. Given the following plant conditions: • Ann. 210: • Drop 18, POWER RANGE FLUX DEVIATION - LIT. • Drop 21, NUCLEAR INSTN SYSTEM TILT CMPTR ALARM - LIT. • Drop 29, ROD SEQUENCE VIOLATION - LIT. • IRPI for Control Bank D rod D4 - 0 steps. • IPRI for rest of Control Bank D - 225 Steps. • Bank Demand for Control Bank D - 225 steps. D Flux N-41 N-42 N-43 N-44 Indication 3.1 3.1 1.2 3. Given the attached section of 2-OHP-4022-012-005, Dropped or Misaligned Rod, which ONE of the following is the required action? a. Declare Rod Position Indication for Rod D4 Inoperable and perform flux map as required per Technical Specification 3.1.7 b. Declare Rod Position Indication for Rod D4 Inoperable and initiate PMP-4030-EIS-001 Event -Initiated Surveillance Testing. c. Initiate a Plant Shutdown per 2-OHP-4021-001-003, Power Reduction. d. Initiate a power reduction. Stabilize the plant at <75% power to begin repairs and recovery. A Unit 2 power ascension was in progress following a maintenance outage. While raising power from 90% to 100% power, several alarms were received. Given the following plant conditions: • Ann. 210: • Drop 18, POWER RANGE FLUX DEVIATION - LIT. • Drop 21, NUCLEAR INSTN SYSTEM TILT CMPTR ALARM - LIT. • Drop 29, ROD SEQUENCE VIOLATION - LIT. • IRPI for Control Bank D rod D4 - 0 steps. • IPRI for rest of Control Bank D - 225 Steps. • Bank Demand for Control Bank D - 225 steps. D F lux N-41 N-42 N-43 N-44 Indication 3.1 3.1 1.2 3. Given the attached section of 2-OHP-4022-012-005, Dropped or Misaligned Rod, which ONE of the following is the required action? a. Declare Rod Position Indication for Rod D4 Inoperable and perform flux map as required per Technical Specification 3.1.7 b. Declare Rod Position Indication for Rod D4 Inoperable and initiate PMP-4030-EIS-001 Event -Initiated Surveillance Testing. c. Initiate a Plant Shutdown per 2-OHP-4021-001-003, Power Reduction. d. Initiate a power reduction. Stabilize the plant at <75% power to begin repairs and recovery.

c. CORRECT. The indications show that rod D4 has dropped and become misaligned. D Flux 2-NRI-14 and the flux deviation alarm indicate that it is an actual misalignment and NOT a RPI failure. The procedure directs that a Plant Shutdown initiated per 02-OHP-4022-012-005 since it has been stable for less than 48 Hours.

Chemistry had just confirmed two leaking fuel rods on Unit 1 when a LOCA occured. Given the following plant conditions: • 1-OHP-4023-E-0 Reactor Trip or Safety lnjection is complete. • 1-OHP-4023-E-1 Loss of Reactor or Secondary Coolant is complete. • 1-OHP-4023-ES-1.2 Post LOCA Cooldown and Depressurization is in effect. • All Red and Orange Paths have been addressed. • Lower Containment high range area monitors, (VRA-1310/1410) are reading 10R/hr. • Pressurizer level is 0%. The SM directs you to implement the highest priority Yellow Path procedure. Note: • 1-OHP-4023-FR-I.2, Response to Low Pressurizer Level • 1-OHP-4023-FR-Z.3, Response to High Containment Radiation Level Which ONE of the following describes the proper procedural implementation based on these conditions? a. Go to 1-OHP-4023-FR-I.2, and exit 1-OHP-4023-ES-1.2 b. Stay in 1-OHP-4023-ES-1.2, and implement 1-OHP-4023-FR-I.2, concurrently c. Go to 1-OHP-4023-FR-Z.3, and exit 1-OHP-4023-ES-1.2 d. Stay in 1-OHP-4023-ES-1.2, and implement 1-OHP-4023-FR-Z.3, concurrently

a. INCORRECT. Z.3 has a higher priority than I.2 b. INCORRECT. Z.3 has a higher priority than I.2 c. INCORRECT. Yellow path procedures are implemented concurrently. d. INCORRECT. Yellow path procedures are implemented concurrently. Z.3 has a higher priority than I.2 * Key has no right answer

A Loss of Off-Site Power has occurred on Unit 2. The crew is performing the actions of 2-OHP-4023-ES-0.2, Natural Circulation Cooldown. Given the following plant conditions: • RCS temperature is 527°F and trending down at approximately 20°F per hour. • RCS pressure is 1850 psig and trending down slowly. • East CCP is operating. • 2-QRV-251, Charging Flow Control Valve, is fully open. • 2-QRV-200, Charging Header Pressure Control Valve, is fully open. • Pressurizer level is 4% and trending down slowly. • High Steam Flow and Low Pressurizer Pressure SI signals are BLOCKED. Which ONE of the following describes the correct action(s) for these conditions? a. Actuate Safety Injection and return to 2-OHP-4023-E-0, Reactor Trip or Safety Injection. b. Transition to 2-OHP-4023-ES-0.3, Natural Circulation Cooldown with a Steam Void in the Vessel. c. Throttle closed the steam dumps to allow RCS temperature to stabilize IAW 2-OHP-4023-ES-0.2, Natural Circulation Cooldown. d. Operate SI pumps as necessary to maintain RCS inventory and avoid overfilling the pressurizer IAW 2-OHP-4023-ES-0.2, Natural Circulation Cooldown.

a. CORRECT. 02-OHP-4023-ES-0.2 Foldout page directs this action when pressurizer level can NOT be maintained at >8%

Given the following plant conditions on Unit 2: • Reactor power is 12%. • The controlling Pressurizer (PRZ) Pressure Channel slowly fails high. • The RO takes manual control of PRZ Pressure Master Controller and Lowers demand. • All PRZ heaters have energized. • RCS pressure is 2075 psig and slowly lowering. • You notice that 2-NRV-163 (PRZ spray) is failed OPEN. • When the RO places 2-NRV-163 in manual it will NOT close. Which ONE of the following is the proper sequence of actions to stop the pressure reduction? a. Trip RCP #23. The RCP trip will NOT cause a reactor trip at this power. Dispatch an AEO to locally isolate Spray Valve 2-NRV-163. b. Reduce Power to 8% so the RCP trips will NOT cause a reactor trip. Trip RCPs #23 and #24. Dispatch an AEO to locally isolate Spray Valve 2-NRV-163. c. Trip RCP #23. The reactor will trip when the RCP is tripped. Go to 2-OHP-4023-E-0, Reactor Trip Or Safety Injection. d. Manually trip the reactor. Go to 2-OHP-4023-E-0, Reactor Trip Or Safety Injection. Trip RCPs #23 and #24.

d. CORRECT. Three loop operation is not allowed. The RCP#3 & 4 must be stopped to stop the spray flow. Therefore the Reactor must be manually tripped and then the RCPs tripped.

Given the following plant conditions on Unit 1: • The unit 1 was at 90% power. • A pressurizer PORV was found to be leaking. • The associated PORV block valve was shut 15 minutes later. For Information • 1-OHP-4021-002-006, Pressurizer Relief Tank Operation • Attachment 3: Draining The Pressurizer Relief Tank • Attachment 4: Feed And Bleed Of PRT To Reduce Pressure Or Temperature • Attachment 7: Adjusting Pressure In The PRT • Attachment 10: Raising Level in the PRT Which ONE of the following represents the expected status of the PRT and the actions that must be taken to restore it to normal limits? a. PRT Temperature = 100°F, Level = 35%, and Pressure = 14 psig Open the Vent to depressurize per Attachment 7 and then perform Attachment 10 to raise level. b. PRT Temperature = 130°F, Level = 81%, and Pressure = 8 psig Open the Vent to depressurize per Attachment 7 and then perform Attachment 4 to lower temperature. c. PRT Temperature = 280°F, Level = 82%, and Pressure = 34 psig Perform Attachment 4 to lower pressure and temperature. d. PRT Temperature = 240°F, Level = 95%, and Pressure = 3 psig Reduce level per Attachment 3 and then perform Attachment 4 to lower temperature.

b. CORRECT. PRT temperature is normally at Containment Temperature of ~100-110°F with level 80-84% and pressure of ~ 2-3 psig. The leaking PORV would have elevated PRT temperature & Pressure. The tank needs to be vented and a Feed & Bleed would be performed to cool the tank.

Unit 1 was operating at a steady state power level of 48%. At 13:35 the following alarm is received: • Ann. 121, Drop 49, MAIN XFMR HYDRAN FAILURE OR H2 CONC HI. The H2 concentration had previously been stable at 245 ppm. The following Hydran readings have been taken: Time Reading • 14:05 257 ppm • 14:35 265 ppm • 15:05 274 ppm • 15:35 283 ppm • 16:05 292 ppm • 16:35 311 pp Based on CURRENT plant conditions, which ONE of the following actions should be directed by the Unit Supervisor? Note: 1-OHP-4024-121, Drops 48 and 49, 1-OHP-4021-081-001, Data Sheet 1, are attached. a. Continue monitoring, and notify Engineering of the abnormal readings. b. Reduce load by 10% in accordance with 1-OHP-4022-001-006, Rapid Power Reduction. c. Reduce load by 30% in accordance with 1-OHP-4022-001-006, Rapid Power Reduction. d. Trip the reactor and enter 1-OHP-4023-E-0, Reactor Trip or Safety Injection

b. CORRECT. Rate of rise is less than 50 ppm/hr but greater than 20 ppm/hr. Action is to perform a 10% load reduction then evaluate in another 3 hours.

Unit 2 was operating at 100% power The following alarms are received: • Ann. 204: • Drop 88, WEST CCW SURGE TANK LVL HI OR LOW. • Drop 98, EAST CCW SURGE TANK LVL HI OR LOW. • Ann. 207: • Drop 7, RCP #1 THERMAL BARRIER CLG WTR D/P HIGH. • Drop 8, RCP #1 THERMAL BARRIER CLG WTR TEMP HIGH. • Drop 9, RCP #1 THERMAL BARRIER DP LOW. • Ann. 238: • Drop 10, R-17A EAST CCW HEADER HIGH RADIATION. • Ann. 139: • 2-CRS-4301, East CCW Header High Radiation Which ONE of the following statements is the required action and why? The required actions are to verify CCW vent (2-CRV-412) shut, notify RP of high activity, and: a. Enter 2-OHP-4022-002-001, Malfunction of a Reactor Coolant Pump, to address the #1 RCP seal failure. b. Enter 2-OHP-4022-016-001, Malfunction of the CCW System, and monitor RCP Bearing temperatures since CCW lines in containment have ruptured. c. Enter 2-OHP-4022-016-003, CCW In-Leakage, to close the RCP thermal barrier valves (2-CCM-453 and 454) since the #21 RCP thermal barrier has failed. d. Enter 2-OHP-4022-016-003, CCW In-Leakage, to remove letdown from service and place excess letdown in service since the letdown heat exchanger has failed.

c. CORRECT. Panel 207 Drops 7, 8, & 9 indicate a failure of the RCP Thermal Barrier. These alarms along with the others (Surge tank level and radiation) indicate the need to close the CCW from RCP Thermal Barrier Valves as per 02-OHP-4022-016-003 steps 1 & 2.

The plant is in MODE 6. Fuel movement was suspended for repairs to the Spent Fuel Bridge Crane. Repairs to the Spent Fuel Bridge Crane are complete. • Source Range Channels N31 and N32 are INOPERABLE. • Source Range Channels N21 and N23 are OPERABLE. • The West RHR pump has just been placed in service due to the failure of the East RHR pump seal. • The Reactor Cavity Water Level is 644' 6". The refueling team has established communications with the control room, and has requested permission to move the next fuel bundle from the fuel building to the core. Are administrative conditions met to recommence fuel movement? a. Yes, but only if the Reactor Cavity Water Level is raised to greater than 644' 9" b. No, the East RHR pump must be restored to OPERABLE. c. No, Source Range Channel N31 or N32 must be restored to OPERABLE. d. Yes, provided that the operator selects N21 or N23 on the front of the Audio Count Rate Channel drawer.

c. CORRECT. For refueling to begin, 2 SR channels are required, one with an audible count rate indication. This comes only from N31 or N32. The gamma metrics (N21 or N23) may be used for the other channel. One RHR pump must be operating and level must be > 23' or 644' 1.5". The conditions for refueling would be met provided that the audible count rate is selected to N31.

Given the following plant conditions on Unit 1: • The unit is operating at 100% power. • A Containment Pressure Relief is in progress. • VRS-1201 Upper Containment Normal Range Monitor failed due to a power supply failure. • The Containment Pressure Relief is stopped. Which ONE of the following describes the restrictions placed on Containment Purge/Pressure Relief Operations? Note: Technical Specifications 3.3.6, Technical Requirements Manual 8.3.8, and the associated Bases are attached. a. Containment Purge/Pressure Relief operations may NOT be performed until VRS-1201 is restored to operable status. b. Containment Purge/Pressure Relief operations may continue under administrative controls provided that VRS-1201 is restored to operable status prior to entering Mode 4 following the next refueling outage. c. Containment Purge/Pressure Relief operations may continue under administrative controls for up to 7 days, provided that area surveys of upper Containment are performed at least once every 24 Hrs. d. Containment Purge/Pressure Relief operations may continue for up to 30 days, provided that grab samples of the containment atmosphere are taken and analyzed at least once every 12 Hrs.

b. CORRECT. Condition A of TS 3.3.6 allows Purge Operations if 2/3 channels per train are operable. The failed channel must be fixed during the next refueling outage.

Given the following plant conditions on Unit 2: • The unit is at 50% power. • East and West Main Feed Pumps (MFPs) are running. • North and South Condensate Booster Pumps (CBPs) are running. • Middle Condensate Booster Pump (CBP) is in Auto. The following alarm is received in the Main Control Room: • Ann. 216, Drop 82, CNDST BOOSTER PUMP MOTOR OVERHEATED. While addressing the alarms, the following events occur: • Ann. 216, Drop 72, CNDST BOOSTER PUMP MOTOR OVERLOAD TRIP - LIT. • Ann. 216, Drop 73, CNDST BOOSTER PUMP DISCH PRESSURE LOW - LIT. • Ann. 215, Drop 41, FEEDPUMP SUCTION HEADER PRESSURE LOW alarmed for approximately 3 seconds then cleared. The following breaker indicating light conditions exist: • North CBP: Red • Middle CBP: Green • South CBP: Green Procedurally, the Unit Supervisor will direct the BOP to ______(1)________, and locally have an operator _______(2)___________. a. 1) trip one Main Feedwater Pump 2) close the South CBP recirculation valve manual isolation. b. 1) start the Middle CBP 2) check the position of 2-CRV-224, Low Pressure Heater Bypass Valve c. 1) start the Middle CBP 2) verify CBP recirculation valve manual isolation valves are throttled. d. 1) trip one Main Feedwater Pump 2) open 2-CRV-224, Low Pressure Heater Bypass Valve

b. CORRECT. The Middle CBP should have started in Auto when Ann 216, Drop 73 alarmed. The Feedpumps do not trip until 180 psig for greater than 5 seconds. Ann. 215 Drop 41 is set at 188 psig. Momentary alarm on Ann. 215, Drop 41 may open CRV-224. If pressure remains above 188 psig. Valve should be closed.

Unit 1 has experienced a rupture of the NESW piping inside containment. NLI-340/341 "FLOOD LEVEL" lights are Lit. Why is safe plant recovery NOT assured for a design basis large break LOCA when "FLOOD LEVEL" lights are Lit? a. Operation of critical ECCS components needed for safe recovery is endangered by submersion. Actions within 1-OHP-4023-FR-Z-2, Response to Containment Flooding may be required. b. Operation of the CTS pumps is endangered by excess debris fouling the containment suction strainers. Actions within 1-OHP-4023-ECA-1.3 Sump Blockage Control Room Procedure may be required. c. Operation of the hydrogen recombiner system is compromised by the high water level. Actions within 1-OHP-4023-FR-Z-2, Response to Containment Flooding may be required. d. Operation of the RHR system is compromised by high suction pressure. Actions within 1-OHP-4023-ECA-1.1 Loss of emergency Coolant Recirculation may be required.

a. CORRECT. Containment design basis flood level takes into account the entire water contents of the RCS, RWST, Ice condenser ice bed melt, and SI accumulators, plus the added mass of a LOCA and a steam line or feedline break inside containment. NESW and CCW may be major contributors to exceeding "flood" level and causing a loss of equipment required for long term cooling.

At 15:50, Unit 2 experienced a LOCA. After the recirculation sump suction valves were opened in 2-OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation, neither RHR pump would start. You have proceeded to 2-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation and are currently at Step 13, "Check if SI Can Be Terminated." At 17:30, the following plant conditions exist: • RCS pressure is 300 psig and slowly lowering. • NR RVLIS is 76% and slowly lowering. • WR RVLIS is 23% and slowly lowering. • CETC Average 345oF. • RWST level is 13% and lowering. • Containment pressure is 6.5 psig and stable. • All Centrifugal Charging and Safety Injection pumps are running. • Both Containment Spray Pumps are currently running with their suction aligned for recirculation. Which ONE of the following actions is required? Note: 2-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation section is attached. a. Do NOT stop or throttle ECCS Pumps, continue attempts to makeup to the RCS. b. Terminate SI. Stop both SI pumps and 1 CCP. Isolate BIT injection and restore normal charging flowpath. c. Stop applicable pumps and throttle BIT injection as required to obtain about 310 gpm of injection flow. d. Stop applicable pumps and throttle BIT injection as required to obtain about 410 gpm of injection flow.

c. CORRECT. Based on these conditions (NR RVLIS<67%, RCPs tripped on Phase B, with subcooling < 86oF) ECCS flow should be reduced to the minimum required per Figure1 as IAW Step 13.b RNO. The time after trip is 100 minutes which is equal to ~307 gpm.

At 0600, the following conditions are noted: • Unit 1 is shutdown, preparing for refueling. • Initial RCS temperature was 175°F. • Initial RCS pressure was 100 PSIG. • Normal Cooldown Alignments. • Subsequently, RHR is lost and the RCS heats up at 4°F/minute. Which ONE of the following correctly identifies the initial MODE and MODE at 0640? Initial MODE MODE at 0640 a. Mode 5 Mode 4 b. Mode 6 Mode 5 c. Mode 6 Mode 3 d. Mode 5 Mode 3

a. CORRECT. With RCS Pressure at 100 PSIG, the RX Vessel head is still tensioned and with temperature is <200°F the plant is in mode 5. At 4°F/min for 40 minutes, temperature will rise 160°F final temperature will be 335°F which is Mode 4.

A LOCA that resulted in significant core damage occurred at 1600 hours. Containment Pressure and Radiation levels were recorded as follows: Radiation Pressure Time (R/Hr) (psig • 1630 950,000 6.2 • 1700 950,000 5.6 • 1730 950,000 5.2 • 1800 950,000 4.5 • 1830 90,000 4. At 1835 hours, while performing Emergency Operating Procedures, a step is encountered which states "Check PRZ level - GREATER THAN 20% [24% ADVERSE]". Which ONE of the following describes the required Pressurizer level and why? a. 20% because adverse values are no longer required because of the limited integrated dose and pressure reduction. b. 24% because adverse values must be used until the integrated dose has been evaluated for lasting effects. c. 24% because adverse containment exists due to the current containment radiation dose rate. d. 24% because adverse containment exists due to the current containment pressure.

b. CORRECT. Adverse containment values are required to be used when containment pressure is >5 psig or >105 R/Hr. When pressure lowers to <5 psig normal values may be used as long as the integrated dose is <106 R. At the levels specified here the integrated dose is above 106 R (9.5R/Hr x105 for 90 minutes) and so the instruments must be evaluated for lasting effects of the radiation.

While preparing a release permit for a waste monitor tank, it is determined that RFS-1010 (Liquid Waste Effluent Sample Flow) switch failed HIGH. Repairs will take at least 3 days. Which ONE of the following actions is required by PMP-6010-OSD-001, Off-site Dose Calculation Manual, regarding the liquid waste release? Note: PMP-6010-OSD-001, Off-site Dose Calculation Manual, Attachment 3.2 is attached. The release: a. may NOT be approved until the flow monitor is restored to OPERABLE. b. may be approved since the effluent monitor (RRS-1001) is OPERABLE. c. may be approved for up to 30 days provided the flow rate is estimated at least once per 4 hours during the actual release. d. may be approved after at least two independent samples are analyzed and at least two qualified persons independently verify the discharge valve lineup.

d. CORRECT. A release may continue if 2 samples are taken and the flowpath is dual verified.

Which ONE of the following sets of duties MAY be delegated by the Site Emergency Coordinator? 1) Notification of Plant Personnel. 2) Performing mitigating actions as required by the EOPs. 3) Classification of the Emergency. 4) Performing initial Offsite Dose Assessment. 5) Approving Protective Action Recommendations. 6) Directing the notification of Off-Site Officials. a. 1, 2, 4 b. 1, 3, 5 c. 2, 4, 6 d. 3, 5, 6

a. CORRECT. Notification of PLANT personnel, Direction of Mitigating actions, and Performance of Dose Assessment activities may be performed by others.

Which ONE of the following describes who has the primary responsibility to assume the role as the Consultant and Liaison to the Fire Brigade leader in the event of a fire emergency on Unit 2? a. Shift Manager b. Shift Technical Advisor c. Unit 1 Senior Reactor Operator d. Work Control Center Senior Reactor Operator

d. CORRECT. The WCC SRO or designee acts as the liaison.

2008 NRC Flashcards

(100 cards)