2010 NRC Flashcards by Nathan Waters

Given the following conditions on Unit 2:
• The crew is responding to a Reactor Trip and has transitioned to 2-OHP-4023-ES-0.1, Reactor Trip Response, from 2-OHP-4023-E-0, Reactor Trip or Safety Injection.
• Following the transition the crew notes the following conditions:
• RCS Pressure is 2100 psig and lowering
• 2-NRV-152, indicating lights show an intermediate position
Which ONE of the following describes the actions the operator should take to address these conditions?
a. Close 2-NMO-152, PORV Block Valve, to stop discharge into the PRT.
b. Open Pressurizer Spray valves to depressurize the RCS and limit the loss of Reactor Coolant.
c. Turn on all Pressurizer Heaters to maintain RCS pressure.
d. Isolate air to containment to fail the Pressurizer PORV closed and stop the RCS mass loss.

a. CORRECT. Block valve is in series with the Pressurizer PORV. Closing the block valve will isolate the leak and prevent further depressurization of the RCS.

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Given the following conditions on Unit 2:
• A reactor trip and safety injection have occurred.
• The crew is responding to a Small Break Loss of Coolant Accident (LOCA).
• All Reactor Coolant Pumps are tripped.
• The crew is depressurizing the Reactor Coolant System (RCS) in accordance with Step 13 of 2-OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization.
• A PORV is being used to depressurize the RCS.
As the depressurization occurs, which one of the following describes the expected trend of pressurizer level and the adverse operating condition that may initially occur as a result?

a. Lowering Pressurizer Level
Uncovering Pressurizer heaters.

b. Rising Pressurizer Level
Water solid conditions in the RCS and Pressurizer.

c. Rising Pressurizer Level
Upper head region voiding may occur.

d. Lowering Pressurizer Level
Upper head region voiding may occur.

c. CORRECT. The caution prior to commencing depressurization in ES-1.2 to refill the Pressurizer, states that a head void may occur as indicated by a rising Pressurizer level as water in transferred from the RCS to the Pressurizer.

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Given the following conditions on Unit 1:
• A Large Break LOCA has occurred.
• Safety Injection has actuated.
Which ONE of the following describes:
1) How the Centrifugal Charging Pump suction swaps to the RWST when a Safety Injection is initiated, and
2) how the CHARGING PUMP SUCTION swapover to the containment sump is completed in accordance with 1-OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation?
Note: VCT valves = QMO-451/QMO-452
RWST valves = IMO-910/IMO-911

a. 1) The VCT valves will start to close AFTER one of the RWST valves have traveled to the full open position.
2) The RWST valves will AUTOMATICALLY close after IMO-340, Charging Pp Suction from East RHR Hx has been opened.

b. 1) The VCT valves will start to close AFTER one of the RWST valves have traveled to the full open position.
2) The RWST valves will be MANUALLY closed after IMO-340, Charging Pp Suction from East RHR Hx has been opened.

c. 1) The VCT valves will start to close AS SOON AS one of the RWST valves start to open.
2) The RWST valves will AUTOMATICALLY close after IMO-340, Charging Pp Suction from East RHR Hx has been opened.

d. 1) The VCT valves will start to close AS SOON AS one of the RWST valves start to open.
2) The RWST valves will be MANUALLY closed after IMO-340, Charging Pp Suction from East RHR Hx has been opened.

b. CORRECT. The valves from the VCT will start to close when the valves from the RWST get fully open and after transfer to the containment sump the valves are placed in the A-Auto position.

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Given the following conditions on Unit 2:
• Unit tripped from 29% power.
• RCP 21 breaker tripped open when the busses swapped.
Which one of the following describes the response of Thot and Tcold in Loop 1?
a. Tcold rises to approximately equal Thot.
b. Thot lowers to approximately equal Tcold.
c. Tcold lowers, Thot remains approximately stable.
d. Thot rises, Tcold remains approximately stable.

b. CORRECT. Loss of RCS flow in 1 loop, reverse flow in that loop will cause Thot to drop (no more forced circulation in that loop) to the Tcold value or slightly below.

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Given the following conditions on Unit 2:
• Reactor Coolant System (RCS) is in mid-loop condition
• The following indications are fluctuating on the running Residual Heat Removal (RHR): amps, flow, and discharge pressure
Which ONE of the following statements is correct regarding the standby RHR pump?
The standby RHR Pump should:
a. NOT be immediately started because air
entrainment could cause a loss of both RHR trains.

b. be immediately started because following a loss of
RHR flow, an RCS pressurization may occur precluding gravity feed makeup.

c. be immediately started because under certain loss of RHR conditions, core uncovery or core voiding can occur within 15 to 20 minutes.
d. NOT be immediately started because starting an idle RHR pump under mid-loop conditions could cause an unacceptable reduction in reactor shutdown margin.

a. CORRECT. The ARG provides a clear guidance which includes industry experience of why operation of an RHR pump operating with air entrapment should be evaluated because it could lead to pump damage. Starting the other RHR pump could transfer the problem to the other pump leading to a complete loss of the system.

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Given the following conditions on Unit 1:
• Unit is operating at 100% power
• 1-NPP-151, PZR Press Channel 1, fails high
• The operator has taken manual control of Pressurizer pressure control and stabilized pressure at 2085 psig.
• Following completion of the procedure for response to a malfunction of a pressurizer pressure instrument, what will be the status of the CVCS/Charging system?
The ______ CCP will be INOPERABLE with the associated emergency leakoff valve deenergized in the ______ position.
a. East; open
b. East; closed
c. West; open
d. West; closed

a. CORRECT. Channel 1 affects QMO-225 for the East CCP. The Emergency leakoff is racked out in the open position to ensure minimum flow in the event of an SI.

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Given the following conditions on Unit 2:
• The operators are implementing 2-OHP-4023-FR-S.1, Response to Nuclear Power Generation / ATWS
• Both Reactor Trip Breakers remain closed.
• The operators have completed steps of 2-OHP-4023-FR-S.1 through opening the MG Set output Breakers to shutdown the reactor.
Which ONE of the following describes the impact and potential consequences of the Reactor Trip Breakers remaining closed?

a. main steam line isolation signal will NOT occur to prevent excessive reactivity during the trip due to rapid RCS cooldown.
b. feedwater isolation signal will NOT actuate to prevent excessive reactor coolant system cooldown from the overfeeding of the steam generators.
c. main generator trip signal will NOT be generated preventing transfer of busses to reserve feed.
d. feedwater flow conservation signal will NOT occur to ensure equal distribution of water to the steam generators.

b. CORRECT. P-4 (Rx Trip Breaker Position) feeds the feedwater isolation signal. Either breaker being open will cause an isolation of flow to the SGs. However with neither breaker open, all FMOs and FRVs will remain open and Main Feedpumps will not trip. This will cause excessive flow to the SGs, which could lead to an overcooling of the RCS.

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Given the following conditions on Unit 1:
• The Unit is operating at 75% power with all systems in automatic.
• Main Turbine DCS is in “MW IN” in preparation for turbine valve testing.
• Condenser vacuum is lowering.
Assuming no action has been taken by the crew, which ONE of the following describes the response of the rod control system to this event?
Control rods will automatically:
a. insert due to the rise in Tavg from the rise in steam flow.
b. insert due to the rise in Tavg from the lowering in steam flow.
c. withdraw due to the drop in Tavg from the rise in steam flow.
d. withdraw due to the a drop in Tavg from the lowering steam flow.

c. CORRECT. With the Main Turbine in “MW IN” control, the turbine valves are allowed to reposition to try to maintain Main Generator Load. As vacuum lowers, the turbine will become less efficient, causing more steam flow for the same MW output. As steam flow rises, RCS Tave will lower below Tref. With rods in AUTO, the rods will withdraw to minimize the Tave-Tref deviation.

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Given the following conditions on Unit 1:
• Unit 1 was operating at 100% power when a condensate system transient caused both Main FW pumps to trip.
• The turbine and reactor failed to trip automatically.
In accordance with the immediate actions of 1-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS, the operators will:
1) Manually trip the Reactor, if it fails to trip insert control rods.
2) Manually actuate AMSAC.
3) Manually trip the Turbine, if it fails to trip, then runback the turbine.
Which ONE of the following describes the bases for these immediate actions in 1-OHP-4023-FR-S.1?
The safeguards systems are designed assuming that the only heat being added to the RCS is from _______ . For an ATWS event with a loss of normal feedwater, a Turbine trip within 30 seconds will ________ .

a. fission product decay and RCP heat;
prevent challenging the Pressurizer PORV’s.

b. fission product decay and RCP heat;
maintain S/G inventory.

c. 5% power;
maintain S/G inventory.

d. 5% power;
prevent challenging the Pressurizer PORV’s

b. CORRECT. Per FR-S.1 Background Document, the assumed heat generation is from decay heat and RCP heat. Turbine is tripped to maintain SG water inventory.

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Given the following conditions on Unit 2:
• A loss of all AC Power occurred due to severe weather conditions and failure of emergency diesel generators to start and supply safeguard buses.
• The operating crew is carrying out actions of 2-OHP-4023-ECA-0.0, Loss of All AC Power.
• The operators are at a point where they are to commence cooldown and depressurization of the steam generators to 190 psig.
Based on these conditions, which ONE of the following statements describes the reason why a secondary depressurization is directed?

a. To prevent a challenge to the Core Cooling Safety Function Status Tree which is being monitored for implementation.
b. To remove stored energy in the steam generators to limit the potential of challenging RCS integrity.
c. To remove available energy in the steam generators and thus minimizing any challenges to the containment structure if a Faulted S/G were to occur.
d. To minimize RCS inventory loss through the RCP seals, which maximizes time to core uncovery.

d. CORRECT. The primary concern is a loss of RCS inventory with no way to recover level. This could lead to core uncovery. The SGs are depressurized to lower RCS temperature and pressure to slow the loss of inventory.

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Given the following conditions on Unit 2:
• Unit has just tripped due to a Loss of Offsite power.
• Both EDGs started and energized the required loads.
• All equipment responded as designed.
The following conditions exist:
• Containment parameters are normal
• Average core exit thermocouple (CET) temperature is stable.
Which ONE of the following combination of RCS pressure and average CET temperature verifies the MINIMUM required subcooling to AVOID Safety Injection per 2-OHP-4023-ES-0.2, Natural Circulation Cooldown?
a. 600 psig, 590oF
b. 500 psig, 460oF
c. 450 psig, 430oF
d. 375 psig, 400oF

d. CORRECT. 2-OHP-4023-ES-0.2 , Foldout Page (FOP) needs >40oF of subcooling, or requires that a SI be actuated. Tsat fo 400 psia (375 psig + 15 psi) is 444.6oF. Based on the conditions provided, 44.6oF of subcooling exists, exceeding the 40oF requirement.

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Given the following conditions on Unit 1:
• Unit is operating at 60% power
• Pressurizer Level Control is in MANUAL
• Pressurizer LEVEL CTRL SELECTOR switch is in the Channel 2-3 position
• CRID 3 power supply fails
Assuming no operator action, which ONE of the following statements describes the effect of this failure on the CVCS and PZR level control system?

a. QRV-251, CCP Disch Flow Control fails OPEN
Letdown Isolates
Actual Pressurizer Level Rises

b. QRV-251, CCP Disch Flow Control fails CLOSED
Letdown Isolates
Actual Pressurizer Level Rises

c. QRV-251, CCP Disch Flow Control fails CLOSED
QRV-200, Charging Header Pressure Control Valve fails OPEN
Actual Pressurizer Level Lowers

d. QRV-251, CCP Disch Flow Control fails OPEN
QRV-200, Charging Header Pressure Control Valve fails CLOSED
Actual Pressurizer Level Lowers

a. CORRECT. Loss CRID 3 causes a Loss of PZR level Channel NLP-153 which will result in an indicated low Pressurizer level. This will cause the PZR Level control to Close QRV-112 and Open QRV-251. (Note that QRV-251 will also fail Open & QRV-112 will Close from Loss of CRID 3 per OHP-4021-082-008 Table 3. i.e. - either NLP-153 failing or loss of power causes same effects)

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Which ONE of the following describes the reason for disabling AMSAC, de-energizing DCS Inverters, and stopping all DC powered Turbine Oil Pumps in 2-OHP-4023-ECA-0.0, Loss of All AC Power ?

a. Allow turbine oil systems to be de-energized and drain to the main turbine lube oil tank.
b. Prevent inadvertent actuation of control systems and auto start of pumps.
c. Extend the DC battery life for N Train and BOP batteries.
d. Limit overheating of cabinets and pump motor overload.

c. CORRECT. ECA-0.0, Step 17 Note states that DC Loads are shed to extend the life of the DC batteries associated with the loads.

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Given the following conditions:
• An accidental spill of the Monitor Tank has occurred in the Aux Building.
• Radiation levels in the area of the spill are 40 mRem per hour at 30 cm.
• Contamination levels based on smear on the floor around the tank are 1.2 x104 dpm/100 cm2 beta-gamma.
Which ONE of the following describes how the area will be posted in accordance with PMI-6010, Radiation Protection Plan?
a. Radiation Area ONLY.
b. Contamination Area ONLY.
c. Radiation Area AND Contamination Area.
d. High Radiation Area AND Contamination Area.

c. CORRECT. This area should be posted as a radiation area (>5 mrem in 1 hour and <100 mrem/hr) and a contamination area (>1000 dpm/100 cm2).

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Given the following condition:
• 1-MRV-213, Unit 1 SG11 PORV, was locally isolated due to excessive leakby.
Which ONE of the following describes the status of the SG11 PORV Radiation Monitoring and the reason?
a. Channel MRA-1601 is Unavailable. Technical Requirement actions are NOT required since all of the other SG PORV monitors are operable.
b. Channel MRA-1601 is Inoperable. Technical Requirements actions are required since PORV isolation renders the radiation monitoring function Inoperable.
c. Channel MRA-1601 is Unavailable. SG11 Radiation Monitoring is still Operable since MRA-1602 is still functioning.
d. Channel MRA-1601 is Operable. Isolation of the SG PORV does NOT affect radiation monitor operability.

b. CORRECT. TRO 8.3.8 requires 1 Channel per loop to be Operable. The SG PORV Monitor is required to be declared Inoperable if the PORV is closed or isolated. Declaring the SG PORV Radiation monitor inoperable requires that Function 2.b Action C be applied.

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Given the following conditions on Unit 2:
• Unit was operating at 100% power when a malfunction of the Control Air system occurs.
• The Control Air header rapidly depressurizes and cannot be restored.
Which ONE of the following describes the correct operator response?
Immediately trip the Reactor and implement:

a. 2-OHP-4023-E-0, Reactor Trip or Safety Injection.
2-OHP-4022-064-002, Loss Of Control Air Recovery, may be performed concurrently after transitioning to 2-OHP-4023-ES-0.1, Reactor Trip Response.

b. 2-OHP-4023-E-0, Reactor Trip or Safety Injection.
2-OHP-4022-064-002, Loss Of Control Air Recovery, is NOT needed since the EOP network may be performed without reliance on Control Air.

c. 2-OHP-4023-E-0, Reactor Trip or Safety Injection.
2-OHP-4022-064-002, Loss Of Control Air Recovery, may NOT be performed until completion of 2-OHP-4023-ES-0.1, Reactor Trip Response.

d. 2-OHP-4022-064-002, Loss Of Control Air Recovery, until restoration of Control Air from any source.
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 Non-Accident (ES-0.1, 0.2 or 0.3) Emergency Procedures after the immediate actions are complete at US discretion.

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Given the following conditions:
• A Main Transformer fire has occurred on Unit 1.
• The Unit 1 Reactor and Main Turbine/Generator have been tripped.
• The Turbine AEO has reported that the Main Transformer deluge system has actuated.
• The Outside Tour AEO has reported that all three fire water pumps are running.
The reported status of the Fire Water System is ____________ for this event. The Main Transformer deluge system ___________ expected to automatically actuate once the Main Generator is tripped, ___________ fire water pumps are expected to be running.
a. abnormal; is; but only 2
b. abnormal; is not; but 3
c. normal; is; and 3
d. abnormal; is not; and only 2

a. CORRECT. For any given actuation of the fire system, the maximum number of pumps running should be 2. All three pumps running are an indication of a piping rupture. The deluge valve will not automatically actuate until the main transformer is de-energized.

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Given the following conditions on Unit 2:
• Reactor Power is at 100% when a fire occurs in the Control Room Cable Vault.
• A large amount of smoke accumulates in the Control Room,
• The Shift Manager determines that the main control room must be evacuated in accordance with 2-OHP-4025-001-001, Emergency Remote Shutdown.
Which ONE of the following describe the operation of the Emergency Diesel Generators (EDGs)?

a. Start both EDGs prior to control room evacuation.
Locally control EDGs from LSI panels as required.

b. Trip both EDG HEAs prior to leaving control room.
Restore EDGs per 2-OHP-4025-R-15, if required

c. Leave EDG control switches as is.
Locally Trip and Isolate EDGs in accordance with 2-OHP-4025-LTI-3, if required.

d. Leave EDG control switches as is.
Locally control EDGs from LSI panels as required.

c. CORRECT. EDGs are left as is. Trip, Isolation, and restoration is performed following evacuation per appendix R procedures.

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Given the following conditions on Unit 1:
• Unit 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.
The following conditions now exist:
• 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 T11 A and T11B to lose power. This will cause the EDG to start and energize T11A and T11B.

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Given the following conditions on Unit 1:
• Reactor power is 100%.
• West CCP is in operation with the East CCP in standby.
• West CCW pump is tagged out for maintenance.
Which ONE of the following describes the immediate operator actions required for a loss of Bus T11D:

a. Trip reactor because the RCP seals will overheat without Component Cooling flow.
b. Trip reactor because there is NO charging flow to replace letdown.
c. Initiate a controlled shutdown because the Charging pump will overheat without Component Cooling flow.
d. Initiate a controlled shutdown because the RCP seals will overheat without charging flow.

a. CORRECT. Per Loss of CCW procedure, Trip Reactor and Then trip RCPs.

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Given the following conditions on Unit 2:
• Reactor is at 100% power.
• All control systems are in normal alignment.
• Letdown flow is aligned with a flow of 120 gpm at QFI-301.
The following parameters are now noted on the CVCS system:
• Seal Return Flows are 3 gpm per RCP
• Charging flow is 137 gpm and rising.
• 2-QTA-160, Regen HX Outlet Temp - Letdown, has lowered 5°F from its steady state value.
• VCT level is 33% and lowering.
• PZR level is 55% and lowering slowly.
• RCS temperature is 574°F and stable.
Which ONE of the following describes the effect on the unit and the action required to address the conditions?

a. RCS leakage is from the letdown line between the orifices and the letdown containment isolation valves. Isolate Letdown.
b. RCS leakage is from the charging line on the RCS side of the regenerative heat exchanger. Isolate Charging and Letdown.
c. RCS leakage is from the letdown line on the CVCS side of the regenerative heat exchanger. Initiate an investigation to determine if the leak is isolable.
d. RCS leakage is from the charging line on the CVCS side of the regenerative heat exchanger. Isolate Charging and Letdown.

b. CORRECT. If Regen Hx Outlet temperature is lowering, then more charging flow is going through the Regen Hx. This means the leak is downstream of the RHX in containment. Since Charging flow increased from a normal value of 132 GPM to 137 GPM with no other changes the leak rate is approximately 5 GPM and will required isolation of the charging header to isolate the leak.

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Given the following conditions:
•	QRV-200, RCP Seal Backpressure Valve, is operating at 30% open.
Assuming QRV-251, Charging Line Flow Control Valve is NOT adjusted, IF QRV-200 fails to 60% open, THEN:
	Charging Pump	RCP Seal	Charging Flow to
	Discharge Press	Injection Flow	Regen Hx
a.		Lowers		Rises		Lowers
b.		Rises		Lowers		Rises
c.		Rises		Rises		Lowers
d.		Lowers		Lowers		Rises

d. CORRECT. QRV-200 will cause a lower backpressure on the CCP discharge and seal injection line, resulting in lower CCP discharge pressure and less flow to the RCP seals. In addition this action will raise charging flow.

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Given the following conditions:
• Unit 2 is in Mode 4 during cooldown per 2-OHP-4021-001-004, Plant Cooldown from Hot Standby to Cold Shutdown
• West RHR Pump and Heat Exchanger are operating, aligned to the cooldown path through injection lines to Cold Legs Loops 2 & 3
• RCS temperature is 300°F and stable
• RCS pressure is 335 psig and stable
The air supply line to IRV-320, West RHR Hx Outlet Valve, breaks, causing a complete loss of Instrument Air to the valve.
Which ONE of the following describes the effect on the plant and the action that could be taken to mitigate the transient?
a. RHR Flow through the West HX will be lost. Throttle open IRV-311 RHR HX Bypass to maintain greater than 3000 gpm RHR flow.
b. RHR Flow through the West HX will be lost. Stop the West RHR pump immediately to prevent overpressurizing letdown.
c. RHR Flow through the West HX will rise. Throttle ICM-111, RHR Discharge to Cold Leg 2 & 3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.
d. RHR Flow through the West HX will rise. Throttle ICM-321, West RHR Injection to Loops 2 & 3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.

d. CORRECT. IRV-320 fails open on loss of air. This will raise RHR flow through the HX. ICM-321 can be throttled closed to reduce total RHR flow and IRV-311 can be throttled open to allow more flow to bypass the HX in order to control RCS cooldown.

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Given the following conditions:
• Unit 2 has experienced a loss of both CCW pumps in MODE 3
• NEITHER Unit 2 CCW pump can be restarted.
• CVCS crosstie from Unit 1 is NOT available.
• 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 Seal injection crosstie.

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Given the following conditions: • The Plant has just completed a Heatup to Normal Operating Temperature and Pressure. • Operators suspect a small leak through Pressurizer Safety Valve SV-45B. What indication combinations are available to help the operator determine if this valve is faulted? a. A significant PRT Temperature Rise (>200oF) The Common Safety Valve Tailpipe Temperature indicator The Common Safety Valve line acoustic monitor b. A significant PRT Temperature Rise (>200oF) The Safety Valve SV-45B Tailpipe Temperature indicator The Safety Valve SV-45B line acoustic monitor c. A slight PRT Temperature Rise (<50oF) The Safety Valve SV-45B Tailpipe Temperature indicator The Common Safety Valve line acoustic monitor d. A slight PRT Temperature Rise (<50oF) The Safety Valve SV-45B Tailpipe Temperature The Safety Valve SV-45B line acoustic monitor

d. CORRECT. For a small leak into the PRT there should not be a significant rise in PRT temperature. PORVs share indicators and safeties have separate indicators for the acoustic line.

Given the following conditions: • Containment pressure instrument Channel #1, 2-PPP-303, declared inoperable. • Required actions per 2-OHP-4022-013-011, Containment Instrumentation Malfunction, have been completed. • Required Technical Specification Actions have been taken for Channel #1, 2-PPP-303. Which ONE of the following describes the SI and CTS, and Containment Isolation Phase A (CIA) and B (CIB) response to a subsequent failure of CRID 4 power supply. SI CTS CIA CIB ACTUATES ACTUATES ACTUATES ACTUATES a. YES NO YES NO b. YES YES YES YES c. NO YES NO YES d. NO NO NO NO

d. CORRECT - The CTS Actuation Bistable is placed in the BYPASSED condition to preventinadvertent actuation. This changes the remaining channel coincidence to 2/3 instead of the previous 2/4. Only 3 channels (Channels 2, 3, & 4) feed the SI Actuation. The bistable for the CTS actuation is placed in the BYPASS condition, making the CTS a 2/3 coincidence for the remaining channels (2, 3, and 4). CRID 4 failure will NOT meet the 2/3 co-incidence for either the SI or CIA. CTS/CIB still required 2/3 to actuate, therefore only one channel will not cause the CTS/CIB. The logic between CTS and CIB is always the same which prevents the student from eliminating distractors based on obvious distractors; this same logic applies to SI and CIA. This forces the student to have a full understanding of the question in order to answer the question correctly.

Given the following conditions: • Unit 1 and Unit 2 are operating at 100% Power. • North Spent Fuel Pit pump and cooler in service • Spent Fuel Pit at Normal Level • Spent Fuel Pit contains 2015 spent fuel assemblies • Refueling Water Purification aligned to purify Unit 1 RWST The 1-CRV-445, CCW from North SFP Hx, control air supply line ruptures causing valve to fail. Which ONE of the following conditions describes the impact on the Spent Fuel Pool Cooling system and the actions needed to address this condition? a. SFP Temperature will rise. Place the South SFP Pump and Hx in service. b. SFP Temperature will rise. Manually control SFP temperature using 1-CRV-445 bypass valve. c. SFP Temperature will lower. Place the South SFP Pump and Hx in service. d. SFP Temperature will lower. Isolate 1-CRV-445 and manually control SFP temperature using 1-CRV-445 bypass valve.

a. CORRECT. 1-CRV-445 fails closed on loss of air. Due to the heat load in the SFP, the SFP temperature will rise. Impacts the ability of Unit 1 CCW system to provide SFP Cooling. Unit 2 SFP cooling loop will need to be placed in service.

Given the following conditions on Unit 1: • A Small Break LOCA has occurred. • RCS Wide Range Pressure lowered to 1350 psig and is stable. • Containment pressure has remained less than 2.8 psig. • 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 depressurize the RCS. b. All Pressurizer PORVs to depressurize the RCS. c. The PZR Aux Spray Valve to spray down the PZR steam space. d. PZR Normal Spray Control valve(s) to spray down the PZR steam space.

d. CORRECT. This is the "normal" method used to depressurize the RCS in ES-1.2.

The plant was in Mode 1. Reactor trip and safety injection have occurred. Due to high Aux Building radiation levels, the crew has entered 2-OHP-4023-ECA-1.2, LOCA Outside Containment. Actions have been taken in an attempt to isolate the break. Given the following plant conditions: • PZR level is off-scale low • SI pump flow is 0 GPM • RCS pressure is 1700 psig and rising. • Aux Building Radiation Monitors are in alarm Which ONE of the following describes the status of the leak based on the requirements of 2-OHP-4023-ECA-1.2? a. The leak is isolated based on SI flow of 0 GPM b. The leak is isolated based on RCS pressure rising. c. The leak is NOT isolated based on PZR level indication not rising. d. The leak is NOT isolated based on Aux Building radiation monitor indication.

b. CORRECT. RCS pressure is the required parameter for determination of isolation Incorrect.

The control room operators are responding to a red path on the Heat Sink CSF. While attempting to restore feed flow to a SG in accordance with OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink, conditions degrade to the point that RCS bleed-and-feed must be established. Under these conditions, RCS bleed-and-feed must be established expeditiously to: a. prevent a loss of secondary heat sink. b. minimize core uncovery and prevent inadequate core cooling. c. prevent an overpressurization challenge to the reactor vessel. d. prevent a rapid RCS overpressurization, followed by a rapid RCS depressurization due to RCP seal failure.

b. CORRECT. If the operator cannot restore feedwater flow to the SGs, conditions will degrade to the point where RCS bleed and feed must be established to minimize core uncovery and prevent inadequate core cooling.

Given the following conditions: • Unit 2 Reactor Tripped due to a loss of offsite power • The crew is implementing 2-OHP-4023-ES-0.2, Natural Circulation Cooldown Which ONE fo the following describes the reason for maintaining subcooling greater than 90°F if ALL CRD fans are running OR greater than 220°F if less than ALL CRD fans are running during the cooldown. a. To collapse any voids formed in the CRD housings. b. To prevent possible void formation in the upper head. c. To prevent degradation of reactor coolant pump seals due to steam. d. To ensure adequate subcooling due to possible degradation of core exit T/Cs accuracy.

b. CORRECT. 2-OHP-4023-ES-0-2, Natural Circulation Cooldown requires an RCS subcooling of 220°F in the event CRDM fans are NOT running to preclude void formation in the upper head. Normal natural circulation RCS subcooling is 90°F.

Given the following conditions: • 1-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation, has just been entered • Refueling Water Storage Tank (RWST) level is 5.5% Which ONE of the following procedure actions is performed FIRST? a. Initiate makeup to the U-1 RWST from the Boric Acid Blender. b. Start one reactor coolant pump. c. Initiate makeup to the U-1 RWST from the U-2 RWST. d. Secure all ECCS and CTS pumps taking a suction from the RWST.

d. CORRECT. The Foldout Page has actions to secure RHR & CTS when level is < 11% and CCP & SI pumps when level is <7%. This would be the first action taken upon entering the procedure.

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: • Containment pressure is 3 psig. • The crew has taken action to minimize the plant cooldown. • Steam Generator AFW flow indicates 25x103 pph each SG. • T-hots are slowly lowering. • The following alarms are received: • Ann. 213 Drop 5, STEAM GEN #1 WATER LEVEL LOW-LOW • Ann. 213 Drop 35, STEAM GEN #2 WATER LEVEL LOW-LOW • Ann. 214 Drop 5, STEAM GEN #3 WATER LEVEL LOW-LOW • Ann. 214 Drop 35, STEAM GEN #4 WATER LEVEL LOW-LOW Which ONE of the following actions is required in accordance with 2-OHP-4023-ECA-2.1? a. Adjust AFW flow to 60x103 pph on each Steam Generator. The minimum NR level, per 2-OHP-4023-ECA-2.1, is 28%. b. Adjust AFW flow to 60x103 pph on each Steam Generator. The minimum NR level , per 2-OHP-4023-ECA-2.1, is 50%. c. Maintain AFW flow at its current value. If T-hot starts to rise, raise AFW flow to stabilize RCS temperature. d. Maintain AFW flow at its current value. If SG levels continue to lower, raise AFW flow to maintain SG levels >13% to prevent a transition to 2-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink.

c. CORRECT. AFW Flow is maintained at a minimum amount since the levels are low and T-hot is lowering.

Given the following conditions on Unit 2: • A LOCA occurred 60 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 if the CTS pump trips. c. Do NOT place RHR spray in service because the RHR pump suction is NOTaligned to the RWST. d. Do NOT place RHR spray in service because ONLY one RHR pump is operating.

a. CORRECT. RHR has injected for 50 minutes. (A LOCA occurred on Unit 2 sixty minutes ago.)

Chemistry had confirmed two leaking fuel rods on Unit 1 when a Small Break LOCA occurred 12 hours ago. The following conditions exist on Unit 1: • All Red and Orange Paths have been addressed. • Containment pressure is 1.0 psig. • Containment air temperature is 215ºF. • Lower Containment high range area monitors, (VRA-1310/1410) are reading 10 R/HR • 1-OHP-4023-FR-Z.3, Response to High Containment Radiation Level, is entered. In accordance with 1-OHP-4023-FR-Z.3, which ONE of the following must be verified? a. Both Containment Recirculation Fans (CEQ) are running. b. Upper and Lower Containment Ventilation Fans (CUV/CLV) are running. c. Containment Ventilation Isolation has occurred. d. Control Room Ventilation System is in ISOLATE.

c. CORRECT. 1-OHP-4023-FR-Z.3 requires the crew to verify Containment Ventilation Isolation.

Given the following conditions: • Unit 1 in Mode 4 cooling down to Mode 5 • Pressurizer level is 80% • 11PHC Pressurizer heater groups are in MANUAL and ON • Reactor coolant pumps (RCP) #13 and #14 are running • An electrical fault results in the loss of RCP Bus 1D and T11D Five minutes later it is reported the pressurizer outflow cannot be verified. Which of the following actions will reinitiate and then maintain a continuous pressurizer outflow? a. Verify pressurizer heaters from 11PHC output current and close NRV-164 Loop 4 PZR Spray Control valve b. Energize pressurizer heaters from 11PHA and close NRV-163 Loop 3 PZR Spray Control valve c. Raise the demand on NRV-164 Loop 4 PZR Spray Control valve d. Adjust charging and letdown to raise pressurizer level to 85%

b. CORRECT. The loss of RCP Bus 1D causes the loss of RCP #3. The loss of T11D causes the loss of PZR Heaters 11PHC. The operator should energize the 11PHA heaters and close the PZR Spray valve associated with the tripped RCP.

Unit 2 was operating at 100% power when an instrument failure caused Pressurizer Level to rise to 73%. The Crew has restored the level control instrumentation and has stabilized charging and letdown. The following conditions exist at 1300 hours: • 2-QRV-160 and 2-QRV-161, Letdown Orifice Valves are Open • Letdown Hx Outlet Flow QFI -301 118 gpm • Charging Header Flow QFI -200 120 gpm • Total seal flow to RCPs QFI -210 to 240 32 gpm • Pressurizer Level NLP-151 to 153 73% Assuming no operator actions, design RCP seal return flows, and pressurizer volume of 75 gallons/% at what time will the Pressurizer Level be returned to the 100% program value? a. 1348 hrs. b. 1406 hrs. c. 1522 hrs. d. 1618 hrs.

c. CORRECT. Unit 2 100% PZR level is 54.1%. With charging flow at 120 gpm and letdown at 118 with 12 gpm from the seals, a net of 10 gpm is being removed from the RCS system. Based on this and a conversion of ~75 gallons/% level in the PZR (either unit), level will reach the program level setpoint of 54.1% at 1522 hrs (T+141.75 minutes - 18.9%=1417.5 gallons).

Given the following conditions: • Pressurizer Pressure Channel #1 has failed and has been placed in the tripped condition. • Reactor trip breaker testing was taking place at 75% power. • Pressurizer Pressure Channel #2 has spiked low causing an inadvertent Safety Injection Actuation and a reactor trip on Unit 1. • Pressurizer Pressure Channel #2 has returned to a normal reading. The following conditions currently exist: • Reactor trip breaker A: OPEN • Reactor trip bypass breaker A: OPEN • Reactor trip breaker B: OPEN • Reactor trip bypass breaker B: CLOSED Which ONE of the following describes the impact (if any) this condition will have on restoring the plant to stable conditions? a. The Train B Safety Injection signal will NOT be able to be reset. Train B equipment will have to be placed in Pull-to-Lockout to stop it. b. The Train B Safety Injection signal will reset but Auto Safety Injection Actuation will NOT be blocked. c. The Safety Injection signal will NOT be able to be reset on either train. Safeguards equipment will have to be placed in Pull-to-Lockout to stop it. d. The Safety Injection signal will reset on both trains. Auto Safety Injection Actuation will be blocked.

b. CORRECT. The SI reset and P-4 block features are train specific. With a failure of Train B reactor Trip Bypass Breaker to open a P-4 signal is not generated on Train B. Since the cause of the SI was a pressure channel spike the SI signal is NOT preventing Train B from being reset. The SI will reset but the Auto SI blocking function of P-4 will NOT function on Train B.

Given the following conditions on Unit 1: • A reactor startup is in progress. • The reactor is critical in the source range. • N41 Power Range channel has failed and been removed from service with all bistables placed in the trip condition. • A loss of power to the CRID 2 bus occurs. Which ONE of the following actions will occur? a. Reactor trips and N32 Source Range channel is de-energized. N31 Source Range channel is still in operation. b. The reactor is critical and BOTH source range channels are de-energized. c. The reactor is critical and N32 Source Range channel is de-energized. N31 Source Range channel is still in operation. d. Reactor trips and BOTH source range channels are de-energized.

d. CORRECT. A loss of CRID 2 causes a loss of power to N42. This loss also causes a loss of power to RPS channel 2. This will cause a trip condition for Power range trips for channel 2. Since N41 is already removed from service its bistable are in the tripped condition. This meets the 2/4 logic to cause a reactor trip. Additionally the signal for 2/4 power range channels above P-10 will cause the SR channels to deenergize.

Given the following conditions on Unit 2: • Reactor power is 3% • East Main Feedwater Pump is in service • Both MDAFW Pumps have been stopped with control switches in NEUTRAL Which ONE of the following signals will cause an automatic start of the MDAFW Pumps? a. AMSAC East Main Feedwater Pump Trip b. Safety Injection East Main Feedwater Pump Trip c. Safety Injection Blackout Sequence d. Blackout Sequence Steam Generator Low Level of 26% on 1 of 4 SGs

c. CORRECT. Safety Injection and Blackout will start AFW Pps in Neutral or AUTO.

During a power ascension, with reactor power at 48%, Control Bank C - Group 1 rod B-8 drops. Prior to the drop it was at 230 steps. While restoring the rod, a control rod urgent failure alarm occurs. Which one of the following explains why the alarm actuated? a. All other Bank C - Group 1 rod lift coil disconnect switches are open. b. All Bank C - Group 2 rod lift coil disconnect switches are open. c. The step counter of the pulse to analog (P/A) converter was not reset to 0. d. Group C rod moving with group D rods withdrawn.

b. CORRECT. Since the dropped rod is completely inserted, the lift coil disconnect switches for all operable rods within the affected bank are opened. An Urgent failure will occur when the misaligned rod begins to move. This is caused by the non-movement of the group without the misaligned rod.

Which set of the following describes the response of the reactor protection system to a controlling feedwater flow instrument failing low with no operator action from 100% power conditions? 1) Turbine/Reactor trip on Low-Low level in associated steam generator 2) Turbine/Reactor trip on High-High level in associated steam generator 3) Feedwater Conservation 4) Feedwater Isolation a. 2 AND 3 b. 1 AND 3 c. 2 AND 4 d. 1 AND 4

c. CORRECT. The FW valves will open when the FW flow instrument fails low, causing actual level to rise to the High-High SG setpoint causing a Turbine trip and FW Isolation.

Given the following conditions: • Unit 2 is in Mode 3 at rated temperature and pressure awaiting a startup. • Lower Containment Cooling NESW supply is throttled to all ventilation units. • A power failure causes a loss of 4 of the 8 Lower Containment Vent units. • Average containment temperature rises from 100°F to 119°F. • Charging Flow Control is in MANUAL • Assume RCS Pressure and Temperature remain Constant. Which ONE of the following describes the change in indicated Pressurizer level due to the rise in Containment temperature? Density lowering in the _______ leg causes indicated pressurizer level to read ________ than actual level. a. reference; higher b. reference; lower c. variable; higher d. variable; lower

a. 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 rises the density & weight of the reference leg lowers. This means that the level in the pressurizer will indicate higher for the same initial actual level.

Given the following conditions on Unit 1: • Unit is in Mode 1 at 100% power • A control air leak has resulted in isolation of glycol to containment. • WIN team states that it will take approximately 6 hours to repair the control air leak. Which ONE of the following describes the operating implications of the loss of glycol to containment. a. Immediately declare the Ice Bed Inoperable. Monitor Ice Bed temperatures to ensure they remain < 27°F every 4 hours. b. Monitor Ice Bed temperatures to ensure they remain at an acceptable range. Enter appropriate Tech Spec actions if Ice Bed temperature rises to > 27°F c. Start all Unit 1 Air Handling Units (AHUs). Monitor Ice Bed Temperature locally once per hour until glycol system is restored. d. Maximize Containment cooling. If glycol cannot be restored within ONE hour declare the Ice Bed Inoperable.

b. CORRECT. Monitoring temperatures to ensure that they remain <27 is all that is required. The loss of glycol alone does not require Tech Spec entry

Given the following conditions on Unit 1: • A Large Break LOCA has occurred 10 minutes ago • 600 VAC buses 11C and 11D de-energized on the trip • All other systems function as desired. • Containment pressure is 5.0 psig and rising Which ONE of the following describes the current status of the CTS Pump discharge valves? a. ALL CTS Pump Discharge valves are OPEN b. IMO-210, East CTS Pump Discharge is CLOSED IMO-211, East CTS Pump Discharge is CLOSED Both West CTS Pump Discharge valves are OPEN c. IMO-220, West CTS Pump Discharge is CLOSED IMO-221, West CTS Pump Discharge is CLOSED Both East CTS Pump Discharge valves are OPEN d. IMO-211, East CTS Pump Discharge is CLOSED IMO-210, East CTS Pump Discharge is OPEN IMO-221, West CTS Pump Discharge is CLOSED IMO-220, West CTS Pump Discharge is OPEN

b. CORRECT. The Loss of 600VAC Bus 11D will cause the discharge valves for the East CTS pump to lose power will prevent them from opening.

Given the following conditions in Unit 1: • Unit is in Mode 5 • Containment pressure is negative 0.3 psig. • RP has requested that Containment Purge be placed in service. 1-OHP-4021-028-005, Operation Of The Containment Purge System, provides a defined sequence of operation due to a concern with Containment Pressure. Which ONE of the following describes this sequence and the reason for the concern while starting up the Containment Purge System? a. Start one Purge Supply fan then open the supply fan valve since Technical Specifications require Containment pressure to be < 0 PSIG at all times. b. Prior to starting the fans, open the Upper Containment Purge Supply valves to prevent Ice Condenser doors from opening when initiating containment purge. c. Prior to starting the fans, open the Lower Containment exhaust fan valves to prevent Ice Condenser doors from buckling when initiating containment purge. d. Start one Purge Exhaust fan then open the exhaust fan valve to prevent a positive pressure from adversely affecting the radiation monitor operations.

b. CORRECT. A low pressure in upper containment with respect to lower containment will cause the Ice Condenser Doors to open. 1-OHP-4021-028-005 Attachment 1 step 4.7.4 is performed to raise/equalize upper containment pressure.

Given the following condition on Unit 2: • Unit is operating at 100% power when an inadvertent Steam Line Isolation occurred. • Immediately following the isolation, and resultant plant response, the operators note that all the SG PORVs failed to open. One minute following the Steam Line Isolation, which ONE of the following describes the maximum expected SG pressures? a. 1025 psig b. 1040 psig c. 1065 psig d. 1085 psig

c. CORRECT. Following the Steam Line Isolation a Rx trip would be expected due to OTDT. SG pressures would initially surge opening most of the safeties but as the RCS cooled down pressures would stabilize on the lowest safety valve setpoint (1065 psig) due to the reduction in Reactor Power and Decay heat during the initial 30 seconds of the event.

Given the following conditions: • Reactor power is 50% with a negative moderator temperature coefficient • Control Rods are in manual. If steam flow is raised by 5 percent, which ONE of the following statements best describes how reactor power will respond to the change? Reactor power will: a. decrease to a new lower value. b. increase temporarily, then return to its initial value. c. increase to a new higher value. d. decrease temporarily, then return to its initial value.

c. CORRECT. The increased Steam flow will cause RCS temperature to lower and add positive reactivity due to the negative MTC

Which ONE of the following power supply failures would allow the steam dump system to continue to operate? a. CRID II b. CRID III c. 250 VDC Bus VDAB d. 250 VDC Bus VDCD

b. CORRECT. CRID III does Not supply power to the Steam Dumps or relays.

Unit 1 has just completed a Heatup and is preparing for a Reactor Startup. You have been directed to open the FW pump Emergency Leak Offs (ELOs). Which ONE of the following describes the indications/systems that you should check prior to opening these valves and why? a. Ensure that the FW pump is reset to allow the ELOs to be opened. b. Ensure that the FW pump oil system is operating and has been warmed to minimize the effects of cold seal water on FW pump bearings. c. Ensure that the FW pump oil system is operating to prevent damage to the FW pumps due to condensate flow spinning the pumps. d. Ensure that the FW pump has been removed from turning gear to prevent damage to the turning gear motor.

c. CORRECT. Placing flow though the FW pumps (opening recirculation valves) causes the turbine and pump to rotate at > 100 rpm and so the oil system is required for bearing protection.

The Unit 2 FW Pump Discharge Header Pressure Transmitter 2-FPC-250A slowly drifts LOW during normal plant operation. This will cause the MFP Speed Control System to generate an indicated FW Delta-P signal ____(1)_____ than required, causing the main feed pump(s) to ______(2)_____. Note: Assume FPC-250A is not identified as failed by DCS. (1) (2) a. larger speed up b. larger slow down c. smaller speed up d. smaller slow down

c. CORRECT. The Main FW Pump Speed control compares the UPC-102A/B (highest) steam header pressure to the FW pump Discharge pressure FPC-250A/B (lowest). The speed control attempts to maintain the Main FW Pump speed such that the FW header to Steam Header DP is on Program. When the FW Discharge Pressure drifts Low, it will appear that a smaller DP exists which will raise FW pump speed to try to raise FW pump Discharge header pressure.

Which ONE of the following describes the functional relationship with respect to controlling steam generator (SG) levels between the Main Feedwater Pumps (MFPs) and the Main Feedwater Regulating Valves (MFRVs) when the unit is ramping from 50% to 100% power? a. The MFPs maintain a variable differential pressure across the MFRVs, while the MFRVs throttle to maintain a constant SG water level. b. The MFPs maintain a constant differential pressure across the MFRVs, while the MFRVs throttle to maintain a variable SG water level. c. The MFPs maintain a variable differential pressure across the MFRVs, while the MFRVs throttle to maintain a variable SG water level. d. The MFPs maintain a constant differential pressure across the MFRVs, while the MFRVs throttle to maintain a constant SG water level.

a. CORRECT. The design of the SGWLC system is to maintain a constant level in the SGs at all power levels. The MFW control system however varies the programming to maintain an optimum DP across the MFRVs.

Given the following conditions on Unit 2: • Unit is in Mode 1. • The TDAFW Pump is tagged out of service. • A Loss of Feedwater causes a reactor trip. • Coincident with the trip, T21D Differential trip actuates. Which ONE of the following describes the Auxiliary Feedwater alignment and approximate flow rates? a. 1 and 4 SGs being fed at 120 x 103 pph each b. 1 and 4 SGs being fed at 240 x 103 pph each c. 2 and 3 SGs being fed at 120 x 103 pph each d. ALL SGs being fed at 120 x 103 pph each

a. CORRECT. The T21D Differential causes a loss of T21D Bus. With a loss of T21D, Only the West MDAFW Pump is available. Capacity is ~240 x 103 pph, and it is aligned to automatically feed 1 and 4 SGs.

The operator incorrectly opens the breaker labeled "7.5 KVA Static Inverter Channel IV" on 250 VDC distribution panel "MCAB". The operator realizes the mistake and immediately recloses the breaker. Which ONE of the following describes the effect of these actions, if any? a. The alternate power source to the CRID Inverter will be lost when the breaker is reclosed. The CRID will transfer to the 120 VAC from the Regulating Transformer. b. The alternate power source to the CRID Inverter will be lost. No automatic action will occur when the breaker is reclosed. The auto transfer lockout must be reset at the inverter. c. The normal power source to the CRID Inverter will be lost so it will auto transfer to the alternate source. When the breaker is reclosed, it will auto transfer to the normal source. d. The normal power source to the CRID Inverter will be lost so it will auto transfer to the alternate source. When the breaker is reclosed, the auto transfer lockout must be reset at the inverter.

c. CORRECT. The static transfer switch provides a virtual zero time transfer to the alternate source in case of inverter failure. Thirty seconds after the static switch transfer event ceases and all system parameters are normal, the static switch automatically re-transfers the load to the inverter, without power interruption.

Which ONE of the following describes the effect on a closed Circulating Water pump breaker if DC control power is lost to the breaker? a. The breaker immediately trips open and cannot be reclosed until control power is restored. b. The breaker can be tripped from the Control Room but automatic trip functions are not operable. c. Automatic trips are not operable and tripping the breaker from the Control Room is not possible. d. Automatic breaker trips are operable but tripping the breaker from the Control Room is not possible.

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

Given the following conditions on Unit 1: • Bus T11B normal supply breaker has opened. • DG1AB Diesel Generator has started and is tied to the bus. • Ann. 118, Drop 53, DG1AB TRIPS DISABLED is LIT Which ONE of the following conditions will automatically trip the diesel generator? a. Engine Speed of 590 rpm b. CO2 actuating in the EDG Room c. Main Bearing Temperature 198° F d. Low Lube Oil Pressure 23 psig

a. CORRECT. EDG is in Emergency Mode so Overspeed Trip is the only one available, 590 rpm is 114.7% of Normal 514 rpm - Trip at 110%.

Which ONE of the following describes a condition that would cause a load conservation signal to be generated for the DG1AB. a. Train A SI and Train A Load Shed b. Train B SI and Train B CTS c. Train A SI and Train A CTS d. Train B SI with a Loss of Offsite Power

d. CORRECT. An SI with a LOOP (and subsequent Load Shed) will generate a Load Conservation signal. Train B is associated with DG1AB.

Which ONE of the following lists the two conditions that will independently cause termination of a liquid release by closing 12-RRV-285, Liquid Waste Disposal Effluent Discharge Header Shutoff Valve, and/or tripping the operating monitor tank pump? a. Low circulating water flow High radiation sensed in the release header b. Low circulating water flow High radiation sensed in the circulating water flow c. Low release header radiation monitor sample flow High radiation sensed in the release header d. High release header radiation monitor sample flow High radiation sensed in the circulating water flow

c. CORRECT. RRS-1001 High alarm sensed on the actual release line will energize R18-AUX & R18-AUX1 which closes RRV-285 and trips the monitor tank pumps. In addition, either high or low sample flow (less than 20% or greater than 90%) will energize R18-AUX & R18-AUX1.

Which ONE of the following lists the Unit 1 Control Room Ventilation system damper alignment for operation during a high alarm on ERS-7401, U1 Control Room Area Radiation Monitor? 1-HV-ACR-DA-1/1A 1-HV-ACR-DA-2 1-HV-ACR-DA-2A 1-HV-ACR-DA-3 Outside air to CR Outside air to CR PRZN Outside air to CR PRZN CR air to PRZN a. OPEN CLOSED PARTIAL OPEN OPEN b. CLOSED PARTIAL OPEN CLOSED OPEN c. OPEN PARTIAL OPEN CLOSED CLOSED d. CLOSED CLOSED PARTIAL OPEN CLOSED

b. CORRECT. On an ERS-7401 high alarm: Damper 1/1A will be closed; Damper 2 will be partially open; Damper 3 opens.

Given the following 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 power supply Which ONE of the following describes the response of the CCW system for the given conditions and the subsequent operator actions required? a. No automatic actions will occur since the West CCW pump is running. Notify RP of the failed CRS-4301, East CCW HX Radiation Monitor. b. No automatic actions will occur since the CRS-4401, West CCW HX Radiation Monitor is still functioning. Split the CCW Trains with Misc Header on the West Train and isolate the East Train. c. 2-CMO-420, West CCW HX Outlet, opens and 2-CMO-410, East CCW HX Outlet, closes. Remove CRS-4301, East CCW HX Radiation Monitor, from service and re-align CCW flow through the West CCW Hx ONLY. d. 2-CRV-412, CCW Surge Tank Vent Valve, will automatically close. Notify RP to remove CRS-4301, East CCW HX Radiation Monitor, from service, then reopen 2-CRV-412.

d. CORRECT. An EXTERNAL FAIL in either CRS-4301 (East CCW Header) OR CRS-4401 (West CCW Header) will close the CCW Vent valve.

The operators are attempting to energize emergency bus T21A from DG2AB during a loss of all AC power. Which ONE of the following states the equipment switch that would NOT be placed in PULL TO LOCKOUT and the reason why? a. ESW pump to ensure diesel cooling. b. CCW pump to ensure cooling to vital loads. c. MDAFW pump to ensure an adequate heat sink is maintained. d. CTS pump to ensure containment integrity is not challenged.

a. CORRECT. Since the EDG is the probably source of power, ESW to the EDG needs to be available immediately to maintain the EDG cooled once it starts and is loaded.

Given the following plant conditions: • Unit 2 is in MODE 4 at 280°F. • The Unit 2 East CCW HX is in service with 5000 GPM of ESW Flow. • The Unit 2 EAST RHR train in service with 5000 GPM of CCW Flow. • All four ESW pumps are in service with crossties open. • The DG2CD is running at a constant 3200 KW for a Surveillance test. Which ONE of the following describes the impact on the listed parameters if the Unit 2 East ESW pump trips? Note: Assume no operator action. U2 East CCW DG2CD Jacket Water (JW) Heat Exchanger Heat Exchanger CCW Outlet Temperature JW Outlet Temperature a. RISES RELATIVELY CONSTANT b. RISES RISES c. RELATIVELY CONSTANT RELATIVELY CONSTANT d. RELATIVELY CONSTANT RISES

a. CORRECT. ESW Flow through the CCW HX is maintained by manually throttling ESW through the CCW HX. When the East ESW pump Trips this will lower flow through the HX (since the flow is manually controlled). The Flow through the DG is maintained constant by the automatic temperature control valve.

Procedure 2-OHP-4022-064-001, Control Air Malfunction, has been entered based upon a 100 psi Control Air Pressure alarm. Which ONE of the following is the correct sequence of events that will occur automatically as air pressures lower? a. 95 psig at PPS-10 (20) Standby PAC starts 90 psig CAS wet receiver pressure CAC starts 85 psig at PPS-11 (21) Plant air header isolates b. 98 psig at PPS-10 (20) Standby PAC starts 97 psig CAS wet receiver pressure CAC starts 95 psig Control Air pressure Plant air header isolates c. 97 psig CAS wet receiver pressure CAC starts 95 psig at PPS-10 (20) Standby PAC starts 90 psig at PPS-11 (21) Plant air header isolates d. 95 psig CAS wet receiver pressure CAC starts 90 psig at PPS-11 (21) Plant air header isolates 85 psig at PPS-10 (20) Standby PAC starts

a. 95 psig at PPS-10 or 20 Stand-by PAC starts 90 psig CAS wet receiver pressure Associated CAC auto-starts 85 psig at PPS-11(21) Plant air header isolates 80 psig Control Air Pressure Manual Reactor Trip

Given the following conditions on Unit 1: • Plant heatup is in progress with RCS temperature at 420°F. • Fire system engineer reports that the fire door at the entrance to the positive displacement charging pump and the centrifugal charging pump rooms is nonfunctional. • Additionally the engineer reports that no other fire system impairments exist. Which ONE of the following is the minimum required action? a. Establish an hourly fire watch patrol within 1 hour. b. Enter Tech Spec 3.0.3 due to no operable charging pumps. c. Verify, by inspection, the operability of the manual fire fighting equipment within 1 hour. d. Close the door and establish a continuous fire watch on at least one side of the fire door within 15 minutes.

a. CORRECT. Hourly fire watch is required since no other impairments exist. Equipment in the area is still operable. Action A.1.1 and A.1.2

Given the following conditions on Unit 1: Reactor has tripped. RO notes that two large sections of the containment isolation panel valves have just gone closed. Which ONE of the following sets of conditions would indicate that this is due to a spurious Phase A signal? CNTMT STEAM LINE SG PZR PRESSURE DELTA-P PRESSURE PRESSURE a. 2.9 psig 80 psid 1005 psig 2000 psig b. 0.2 psig 20 psid 450 psig 1810 psig c. 0.5 psig 15 psid 600 psig 1900 psig d. 1.7 psig 110 psid 800 psig 1850 psig

c. CORRECT. All parameters listed are within values to prevent an SI/Phase A actuation.

A licensed individual is planning to undergo some medical evaluations and a test utilizing radioisotopes. It has been determined that this test will not affect judgment or fitness for duty in any way. Which ONE of the following describes the procedural requirements for these conditions? The licensed individual: a. Does not need to report this condition as a potentially disqualifying condition since it is not a fitness for duty issue. b. Must report this situation to the fitness-for-duty liaison for independent verification that it is not a fitness for duty issue, prior to assuming license duties. c. Must notify the Plant Manager who will evaluate the condition, prior to assuming license duties. d. Must notify the Ops Training Manager of a potential disqualifying medical condition

d. CORRECT. The described condition does not affect judgment nor is it a fitness for duty issue. Since a medical test that utilizes radioisotopes would impact an individual’s ability to enter and exit the auxiliary building, it limits the individual’s ability to perform licensed duties. It is therefore reportable to the Ops Training Manager using Data Sheet 1.

Given the following conditions on Unit 2: • Reactor power is stable at 100%. • A Reactor Operator and the Unit Supervisor are in the Control Room. • A high vibration alarm is received on the Heater Drain pump requiring someone to go behind the panel to check the indications. Which ONE of the following describes the procedurally accepted method of checking the indications? a. The Unit Supervisor can go behind the panel to check the vibration. b. The Reactor Operator can go behind the panel to check the vibration. c. Both the Reactor Operator and the Unit Supervisor are allowed to go behind the panel to check the vibration as long as all controls are in automatic. d. Neither the Reactor Operator or the Unit Supervisor can go behind the panels. They must get another operator to check the vibration.

a. CORRECT. The Unit Supervisor must be in the Control Room but may go behind the panels. The RO must remain in the view of the panels.

Conservative Decision making states that : "WHEN faced with unexpected or uncertain conditions, THEN personnel must promptly identify a transition point at which efforts to keep the unit on-line or on schedule are no longer conservative, nor reasonable." Once this point is reached : a. the reactor must be tripped immediately. b. senior management must be notified to determine course of action. c. actions must be taken to place the unit in a safe condition without hesitation. d. the NRC must be notified and actions taken to address the problem within one hour.

c. CORRECT. OHI-4000, Att. 5, Step 3.4 states "WHEN faced with unexpected or uncertain conditions, THEN personnel must promptly identify a transition point at which efforts to keep the unit on-line or on schedule are no longer conservative, nor reasonable. Once this point is reached, actions to place the unit in a safe condition by reducing power, tripping the reactor, or suspending core alterations must be taken without hesitation.

In accordance with administrative procedures, which ONE of the following conditions would permit troubleshooting to be performed on Technical Specification equipment? a. Equipment is operating. b. Equipment is declared inoperable. c. Equipment is logged operable but degraded. d. Equipment is NOT operating and redundant train is in operation.

b. CORRECT. Troubleshooting activities are allowed if the equipment is declared inoperable.

A maintenance visual inspection requires momentarily placing a 'B' Train pump control switch in PULL-TO-LOCKOUT. 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 its normal 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.

Given the following conditions on Unit 2: • A LOCA has occurred • The TSC has been fully staffed and activated • An individual is needed for lifesaving activities during which 30 Rem of TEDE exposure is expected to be received. Which ONE of the following is correct concerning this lifesaving activity? a. The individual is required to be a volunteer and the Site Emergency Coordinator is required to approve the exposure. b. The individual is required to be a volunteer and the Operations Shift Manager is required to approve the exposure. c. The individual is NOT required to be a volunteer and the Site Emergency Coordinator is required to approve the exposure. d. The individual is NOT required to be a volunteer and the Operations Shift Manager is required to approve the exposure.

a. CORRECT. Once in a lifetime doses in excess of 25 REM require a person to be a volunteer. Any extension above 10CFR20 limits requires SEC approval.

Given the following conditions: • Units 1 and 2 are at 100% power. • Unit 2 has experienced several fuel pin failures. • A leak must be repaired on a pipe at the end of the Aux. Bldg. 601 ft. elev. Pipe tunnel. • The general area dose rate in the location of the repair is 600 mrem/hr. • In order to reach the location of the repair the worker must transit through a 6 Rem/hr high radiation area for 2 minutes and return via the same path. • The worker currently has an accumulated annual dose of 400 mrem. Which ONE of the following is the maximum allowable time that the worker can participate in the repairs and NOT exceed the TEDE Administrative Dose Limit? a. 70 minutes b. 120 minutes c. 140 minutes d. 160 minutes

b. CORRECT. The candidate should determine that the ADL is 2000 mrem. Transient exposure is 400 mrem (6000mrem/hr x 4/60hr). (transit to and from the job). (Current) 400 mrem + (transit) 400 mrem = 800 mrem ADL of 2000 mrem - 800 mrem = 1200 mrem allowable before reaching ADL. 1200 mrem /600 mrem/hr = 2 hours

Given the following conditions on Unit 1: • Containment Purge System is operating in the VENTILATION MODE. • A HIGH alarm on VRS-1505, Auxiliary Building Ventilation Noble Gas Activity Monitor, occurs (unplanned). Which ONE of the following describes the required operator response for the Containment Ventilation System to the failure alarm? a. Stop the Containment Purge and consult with Radiation Protection prior to restarting the system. b. Continue the Purge as long as VRS-1101, Containment Normal Range Area Radiation Monitor still indicating as expected. c. Verify the following: 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. d. Verify the following: 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. When the Containment Purge system is operating in the Ventilation Mode, the automatic isolation signals are blocked. The procedure requires the Purge to be stopped and radiation protection concurrence prior to restarting the system.

Given the following conditions on Unit 1: • A reactor trip occurred from 15% power. • Safety Injection was NOT actuated and was NOT required. • 1-OHP-4023-E-0, Reactor Trip or Safety Injection, has been performed, and a transition to 1-OHP-4023-ES-0.1, Reactor Trip Response, has been made. The following conditions exist: • Tavg is STABLE at 547°F • Pressurizer level is 11% and lowering slowly • RCS subcooling is 32°F and lowering slowly • All NR SG levels are 28 - 30%; • AFW flows indicate 0 klb/hr • Containment pressure is 0.7 psig and rising slowly Which ONE of the following describes the appropriate actions for these conditions? a. Actuate SI and return to 1-OHP-4023-E-0 step 1. b. Go to 1-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink. c. Go to 1-OHP-4023-FR-I.2, Response to Low Pressurizer Level. d. Manually start ECCS pumps and continue with 1-OHP-4023-ES-0.1.

a. CORRECT. The foldout page directs an SI and return to 1-OHP-4023-E-0 if subcooling is < 40°F

Given the following conditions on Unit 2: • A Safety Injection (SI) has ocurred. • The Immediate Action steps of 2-OHP-4023-E-0, Reactor Trip Or Safety Injection, have just been completed. The following steam generator conditions are noted: • SG 21 pressure is 740 psig and lowering slowly. • SG 22 pressure is 450 psig and lowering rapidly • SG 23 pressure is 735 psig and lowering slowly. • SG 24 pressure is 745 psig and lowering slowly. • Main Steam header pressure is 700 psig and lowering slowly. Which ONE of the following actions should be promptly performed to mitigate the event? a. Transition to 2-OHP-4023-E-1, Loss of Reactor or Secondary Coolant. b. Transition to 2-OHP-4023-E-2, Faulted Steam Generator Isolation, c. Close all the SG stop valves and continue with 2-OHP-4023-E-0, Reactor Trip Or Safety Injection. d. Close SG 22 Stop Valve and verify steam supply available to the Turbine Driven Auxiliary Feedwater Pump (TDAFP)

c. CORRECT. Steamlines should be isolated for any number of reasons, such as Automatic steamline isolation failure, RCS temperature lowering (procedural steps)and personnel protection. OHI-4023 allows prudent actions to trip the SG Stop Valves when it is apparent that a steam line leak has occured for personnel protection and in response to automatic action failures.

Given the following conditions on Unit 2: • Unit was at 90% power • Control Rods were in AUTO • CBD began to step out with no mismatch signal. • Rods were taken to MANUAL and rod motion ceased. The following conditions now exist: • CBD Bank Demand position is now 222 Steps. • Group 2 Bank D RPIs ALL indicate 222 Steps. • Group 1 Bank D RPIs indicate as follows: • Rod D4: 205 Steps. • Rod D12: 222 Steps • Rod M12: 207 Steps • Rod M4: 222 Steps Reactor Engineering has determined that all CBD rods are free to move and has provided the following information: • R is 1.041 • CFQ is 2.335 • K(Z) is .95 (at 10 feet) • FQW (Z) 2.174 (at 10 feet) Which ONE of the following identifies the Technical Specification Action Condition(s) that must be entered? Reference Provided a. 3.1.4.A Only b. 3.1.4.B Only c. 3.1.4.A and 3.1.4.B Only d. 3.1.4.B and 3.1.4.D Only

d. CORRECT. Both D4 and M12 are unacceptably misaligned. D4 is 17 steps off and M12 is 15 steps off.

Given the following conditions on Unit 1: • A power escalation is in progress • At approximately 35% power the Main Turbine Trips The following conditions now exist: • All 4 Turbine Stop Valve Status Lights are lit • All RPIs have lost power • Train A Reactor Trip Breaker is Open • Train B Reactor Trip Breaker is Closed • WR Log Power = 7% • WR Startup Rate = 0.0 DPM and stable Which ONE of the following actions is required? a. Implement 1-OHP-4023-E-0, Reactor Trip or Safety Injection. Following completion of Immediate Actions, transition to 1-OHP-4023-ES-0.1, Reactor Trip Response. b. Implement 1-OHP-4023-E-0, Reactor Trip or Safety Injection. During verification of Reactor Trip, transition to Implement 1-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS, and manually insert control rods. c. Implement 1-OHP-4022-001-002, Loss of Load (Load Rejection). When directed go to 1-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS. d. Implement 1-OHP-4022-001-002, Loss of Load (Load Rejection). Upon Turbine Trip Verification, go to 1-OHP-4023-E.0, Reactor Trip or Safety Injection.

b. CORRECT. Due to flux being greater than 5% and not lowering, E-0, Step 1 RNO will require a transition to FR-S.1. Since Reactor trip cannot be verified (flux NOT lowering), manual control rod insertion will be required.

Given the following conditions on Unit 2: • Reactor power was 100% power when an RCS leak developed. • The Unit Supervisor is implementing 2-OHP 4022-002-020, Excessive RCS Leakage. The following conditions now exist: • Letdown flow is isolated. • East and West Charging pumps are operating. • Charging flow is 180 gpm. • Pressurizer level is 51% and constant. • VCT makeup is in Auto. • VCT level is 22% and lowering. • Containment pressure is 0.5 psig and constant. Which ONE of the following describes the required operator action and why (assume all control systems function as designed)? a. Verify that CCP suction automatically aligns to the RWST at 14.0% VCT level and perform a controlled rapid shutdown per 2-OHP-4022-001-006 Rapid Power Reduction Response, to maintain RCS Tavg-Tref. b. Verify VCT auto makeup begins at 14.0 % and then restore 75 gpm letdown to ensure proper regen heat exchanger warming of the charging flow. c. Verify that CCP suction automatically aligns to the RWST at 7.0% VCT level and perform a controlled rapid shutdown per 2-OHP-4022-001-006 Rapid Power Reduction Response since RCS leakage is greater than the Technical Specification Limit. d. Trip the reactor and transition to 2-OHP-4023-E-0, Reactor Trip or Safety Injection since VCT level can NOT be maintained with VCT auto makeup.

d. CORRECT. Leakage in excess of VCT makeup will lead to eventual loss of CCP suction. This would be mitigated by the refueling water sequence swapover to the RWST suction source but this would result in excessive boration of the RCS. Lowering level in excess of auto makeup capability require a RX trip per 2-OHP-4022-002-020.

Given the following conditions on Unit 2: • Unit was operating at 100% power • 2-OHP-4022-016-001, Malfunction of the CCW System is being implemented due to indications of a lowering CCW Surge tank level. • The Crew has started the West CCW pump, split the East and West Headers, and aligned the Miscellaneous Services Header to the East Header. • An AEO has reported that a CCW leak of approximately 150 gpm has been identified in the Aux Building 609' elevation, flowing toward the passenger elevator. The following Surge Tank Level Recorder conditions exist: Train 'A‘ Train 'B' CLR-410 CLR-411 Reading 48“ 18" Trend Stable Lowering The leak is located on the_____(1)______ Header. The Unit Supervisor will direct the crew to ____(2)______. a. (1) Miscellaneous Services (2) 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. (1) East Safeguards (2) Shutdown the East CCW pump and align the Miscellaneous Services Header to the West Safeguards Header in accordance with 2-OHP-4022-016-001, Malfunction of the CCW System. c. (1) West Safeguards (2) Shutdown the West CCW pump and the equipment cooled by the West Header in accordance with 2-OHP-4022-016-001, Malfunction of the CCW System. d. (1) Miscellaneous Services (2) Trip the Reactor, Trip the RCPs, and isolate the Miscellaneous Services Header while performing 2-OHP-4023-E-0, Reactor Trip or Safety Injection.

c. CORRECT. Initial train separation places the Miscellaneous Header on the East Train. The conditions presented indicate that the leak is on the West Safeguards Header which can be isolated from the East Header amd Miscellaneous Header.

Given the following conditions on Unit 2: • A reactor startup is in progress with the reactor just critical. • The operator has just stopped moving control rods. • Intermediate Range Power slowly rises above 2 x 10-10 amps. • ONE source range (SR) nuclear instrumentation channel (N-31) fails LOW. • Remaining power indications stabilize. Which ONE of the following actions, if any, is required for compliance with Technical Specifications? Reference Provided a. No action required, source range not required to be operable. b. Trip the reactor and enter 2-OHP-4023-E-0, Reactor Trip or Safety Injection. c. Conduct a reactor shutdown and restore both SR channels to operability prior to next startup. d. Suspend all operations involving positive reactivity changes until both SR channels are restored to operability.

a. CORRECT. TS 3.3.1 (Instrumentation) establishes that above P-6, the SR NI are not required by TS and will shortly be de-energized by procedure. Since there are no TS implications, the startup may proceed.

Given the following conditions on Unit 2: • A SGTR has occurred coincident with a Loss of Offsite Power • 2-OHP-4023-E-3, Steam Generator Tube Rupture, is being performed. • The Unit Supervisor is at Step 41, Select Appropriate Post-SGTR Cooldown Procedure. The following conditions now exist: • RCS Wide Range Pressure is 800 psig and stable. • RCS Incore Thermocouples are 475oF and slowly lowering • RCS T-hots are 470 and slowly lowering • RCS T-colds are 445 and stable • SG Pressures are 387 psig and stable. • PZR level is 25% and slowly rising. Which ONE of the following describes the status of natural circulation and the appropriate procedural transition for the Unit Supervisor? a. Natural Circulation exists. Transition to 2-OHP-4023-ES-3.1, Post-SGTR Cooldown Using Backfill. b. Natural Circulation does NOT exist. Transition to 2-OHP-4023-ES-3.2, Post-SGTR Cooldown Using Blowdown c. Natural Circulation does NOT exist. Transition to 2-OHP-4023-ES-3.3, Post-SGTR Cooldown Using Steam Dump. d. Natural Circulation exists. Transition to 2-OHP-4023-ECA-3.1, SGTR With Loss Of Reactor Coolant - Subcooled Recovery Desired.

a. CORRECT. Based on the conditions Natural Circ exists: T/Cs, T-hot, and SG Press. (Stable or Lowering); RCS cold leg temperatures are at saturation Temp for SG Press; Subcooling is 51oF (>40oF). 2-OHP-4023-ES-3.1, Post-SGTR Cooldown Using Backfill, is an appropriate transition form E-3.

Given the following conditions on Unit 2: • Steam Gen 1/2/3/4 Steam Line Flow High Alarms - LIT • Steam Gen 1/2/3/4 SF>FWF Flow Mismatch Alarms - LIT • RCS Tavg is 561oF and lowering • Turbine load is lowering • Rods are stepping out • Steam flows are 3.6 x 106 lbm/hr and stable. • FW flows are 2.1 x 106 lbm/hr and rising. Which ONE of the following correctly describes the cause and required action to be taken for the above conditions? a. A steam line break exists. Direct the operators to perform a Reactor Trip and Main Steamline Isolation. b. A feed line break exists. Direct the operators to perform a Reactor Trip and Main Feedwater Isolation. c. Feedwater Pump Delta-P is too Low. Direct the operator to raise FW Pump Speed and FW pump flow. d. MPC-253 has failed LOW. Direct the operators to perform actions for failed First Stage Turbine Impulse Pressure Transmitter.

a. CORRECT. Based on the conditions presented a steam line break has occurred. Steam flow is indicating at the 97 to 98% power range. Tavg is 13oF Low for 98% power. A reactor trip and Steam Line isolation is warranted.

Given the following conditions: • A Gas Decay Tank release is in progress through the Unit 1 plant vent. • A High Alarm occurs on VRS-2505, Unit 2 Vent Low Range Noble Gas Radiation Monitor. • RP determines that VRS-2505 is INOPERABLE due to a failed high channel. The effects of the VRS-2505 High Alarm are that 12-RRV-306, Vent Stack Release Valve, ___________. Additionally, which ONE of the following is required by PMP-6010-OSD-001, Off-site Dose Calculation Manual? Reference Provided a. will automatically close Actions 6 and 9 b. must be manually closed Actions 6 and 9 c. will automatically close Action 9 Only d. must be manually closed Action 6 Only

a. CORRECT. A high alarm on VRS-2505 will automatically close 12-RRV-306. PMP-6010-OSD-001, Att. 3.4, Action 6 requires grab samples to be taken at least once per shift and analyzed for gross activity within 24 hours for continued effluent release through the vent header.

Given the following conditions on Unit 2: • Unit is implementing 2-OHP-4023-ECA-0.0, Loss of all AC Power • Power has just been restored from the U2 AB EDG • Cooling flow to the U2 AB EDG is being checked per Step 32 • ONLY the normal ESW supply to the U2 AB EDG is open and providing flow. Which ONE of the following describes the actions the Unit Supervisor should direct and the reason for those actions? a. Open the Alternate Supply to the AB EDG Maximizes flow through the EDG to compensate for maximum loading. b. Leave the Alternate Supply to the AB EDG closed Prevents a loss of ESW cooling to both trains of equipment due to silt and mud build-up in the component's heat exchangers. c. Open the Alternate Supply to the AB EDG Ensures adequate flow in the event of loss of the normal supply path. d. Leave the Alternate Supply to the AB EDG closed Limits the amount of flow to the EDG to ensure that other components in the ESW train receive adequate flow to support safety functions.

b. CORRECT. Step 32 Note states "The alternate ESW cooling supply to the EDGs should remain isolated unless an EDG is running AND the normal ESW supply is NOT available." This Note was added to prevent a loss of ESW cooling from occurring to both trains of equipment due to slit and mud build-up in the component's heat exchangers if the ESW system trains are cross-tied via the alternate cooling supplies to the EDGs.

Given the following conditions on Unit 1: • Unit is preparing for a reactor start up following a refueling outage. • Tavg is 515°F with a heatup in progress. • During the outage testing was performed per the Containment Leak Rate Testing Program • At 0200, a Station Engineer reports that a mistake had been made in analyzing the required Containment Leak Rate Test results that were conducted prior to exceeding 200°F. • The initial calculated Type A leakage had been recorded as ACCEPTABLE • Re-calculation indicates that the Type A leakage is actually UNACCEPTABLE • The re-calculated values have been verified and reviewed by the Shift Manager Which ONE of the following actions, if any, is required by Technical Specifications in response to this situation? Reference Provided a. Continue with the heatup. Do not enter Mode 2 until the leak test is re-performed b. Enter Tech Spec 3.6.1. Remain in MODE 3 until performance of a risk assessment in accordance with Tech Spec 3.0.4. c. Enter Tech Spec 3.0.3. Be in MODE 5 within 37 hours. d. Enter Tech Spec 3.6.1. Be in MODE 5 within 37 hours.

d. CORRECT. TS 3.6.1 required the Containment to be OPERABLE in Modes 1 - 4. If the Containment is INOPERABLE, then return to OPERABLE status in 1 hour or be in Mode 5 in the following 36 hours (37 hours to Mode 5)

Given the following conditions on Unit 1: • A LOCA occurred 30 minutes ago • RCS pressure is 125 psig • RCS Core Exit TCs read 380°F • RCS Cold Leg temperatures are 250°F • 1N SI Pump is running providing 650 gpm flow • 1E RHR Pump is running providing 3000 gpm flow What is the appropriate action taken in response to the above conditions? Entry into 1-OHP-4023-FR-P.1, Response to Pressurized Thermal Shock Condition, is: a. made but NO actions are implemented before returning to procedure in effect. b. made and cooldown will continue within a limit of 50°F in any 60 minute period. c. made and a RCS temperature soak for a ONE hour period will be completed. d. NOT required since RCS pressure is below 300 psig.

a. CORRECT. Entry into FR-P.1 is required due to the Orange Path with RCS at <285°F. The first step of P.1 checks RCS pressure at greater than 300 psig. Since Pressure is less than 300 psig and RHR flow is >400 gpm, no actions are performed and the operator is directed back to the procedure & step in effect.

Given the following conditions on Unit 1: • Unit is responding to a LOCA using 1-OHP-4023-E-1, Loss of Reactor or Secondary Coolant. • The Unit Supervisor is at Step 11, Initiate Evaluation Of Plant Status. The following plant conditions now exist: • RWST Level is 42% and lowering • Containment pressure is 0.3 psig and stable • Containment Recirc Sump Minimum Recirc Level Lights - NOT LIT • East RHR Pump Compartment Sump Annunciator - LIT • East RHR Pump Discharge pressure is 600 psig • Aux Building area radiation monitors are in alarm Which ONE of the following procedures should the Unit Supervisor transition into from 1-OHP-4023-E-1? a. 1-OHP-4023-ECA-1.2, LOCA Outside Containment b. 1-OHP-4023-ECA-1.3, Sump Blockage Control Room Procedure c. 1-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation d. 1-OHP-4023-ES-1.3, Transfer To Cold Leg Recirculation

a. CORRECT. Based on RHR Pump Compartment Sump and Aux Building Rad alarms being in (along with the AEO report of the leak), transition to ECA-1.2, LOCA Outside Containment, would be required.

Given the following conditions on Unit 1: •The Unit Supervisor entered 1-OHP-4022-016-001, Malfunction Of The CCW System, 5 minutes ago. The following plant conditions now exist: •Reactor Power is 8% •All RCP Motor Bearing Lube Oil CCW flow low annunciators are lit. •All flows and temperatures are STABLE as follows: RCP RCP RCP RCP #11 #12 #13 #14 •Upper Motor Bearing temps 189oF 190oF 205oF 191oF •Motor Bearing Lube Oil CCW flow 4 gpm 4 gpm 3 gpm 3 gpm Which ONE of the following describes the sequence of actions the Unit Supervisor will provide to the panel operators? a. 1) Perform a rapid shutdown per 1-OHP-4021-001-006, Rapid Power Reduction 2) Trip #13 RCP immediately after the reactor is shutdown. 3) Close NRV-163, Loop 13 PZR Spray Control b. 1) Trip the reactor 2) Go to 1-OHP-4023-E-0, Reactor Trip or Safety Injection 3) Trip #13 RCP after Reactor Trip is verified 4) Close NRV-163, Loop 13 PZR Spray Control c. 1) Trip the reactor 2) Go to 1-OHP-4023-E-0, Reactor Trip or Safety Injection 3) Trip all RCPs after Reactor Trip is verified 4) Close NRV-163 and NRV-164, Loop 13 and 14 PZR Spray Control d. 1) Perform a rapid shutdown per 1-OHP-4021-001-006, Rapid Power Reduction 2) Trip all RCPs after the reactor is shutdown. 3) Close NRV-163 and NRV-164, Loop 13 and 14 PZR Spray Control

b. CORRECT. 1-OHP-4022-016-001, Caution prior to Step 1 states that the affected RCP must be removed from service by performing Step 17 if motor bearing temperatures exceed 200oF. Step 17 and associated note states that the order is to Trip the Reactor, go to E-0, verify reactor trip, then trip the AFFECTED RCP.

Given the following conditions: • The refueling crew has just set a new fuel assembly next to an irradiated fuel assembly in the Unit 1 core. • The manipulator crane operator observes the refueling cavity level lowering rapidly. • The SRO-CA and Control Room begin implementing 1-OHP-4022-002-006, Loss of Refueling Water Level during Refueling Operations. The following conditions exist Twenty Minutes Later: • The Transfer Tube Gate Valve has been closed • The Weir Gate could NOT be closed • RCS level is 614' and lowering • RWST to RHR makeup is in progress Based on these conditions, which ONE of the following describes the correct actions for the Control Room SRO? a. Go to 12-OHP-4022-018-001, Loss of SFP Cooling. b. Go to 1-OHP-4022-017-001, Loss of RHR Cooling c. Direct the SRO-CA to verify integrity of the Refueling Cavity Seal d. Direct the SRO-CA to check for misalignment of the Refueling Cavity Drains

b. CORRECT. If RCS Level can NOT be maintained >614' then an RCS leak is in Progress and 1-OHP-4022-017-001 Loss of RHR Cooling Must be Initiated.

Given the following conditions on Unit 2: • Unit is in MODE 4 and preparing to startup following a refueling outage. • During the outage, all 5 Main Steam Safety Valves (MSSVs) associated with SG21 were removed and inspected. • The maintenance was satisfactory and the MSSVs will function if required. • The Inservice Testing of the MSSV setpoints has NOT been performed. • A risk assessment has NOT been performed. • Without reliance on SR 3.0.3, which ONE of the following describes if the reactor startup can proceed to MODE 3 with the MSSVs in this condition? Reference Provided a. No. The TS ACTIONS must be immediately entered and all portions of the post testing must be completed before entering MODE 3. b. No. An alternate method of setpoint verification must be used and MSSV OPERABILITY must be demonstrated before entering MODE 3. c. Yes. However, when the unit reaches MODE 2, the test must be at least started within 24 hours after entering MODE 2. d. Yes. However, when test conditions can be established, the test must be completed prior to MODE 2.

d. CORRECT. SR 3.7.1.1 NOTE states that the Surveillance Requirements are only required to be performed in MODES 1 and 2. Per Example 1.4-5, this note allows entry into Mode 3 to perform the surveillance.

Given the following conditions on Unit 2: • A Main Generator fire has been confirmed by an AEO who reported that flames are coming out of the Unit 2 Main Generator Shaft. After directing a Reactor/Turbine Trip and entering 2-OHP-4023-E-0, Reactor Trip Response, which ONE of the following would you direct the AEO to perform first? Dispatch an AEO to: a. Shut down the Main Generator Seal Oil System per 2-OHP-4021-080-002 Operation Of Shaft Seal Oil System b. Depressurize and Purge the Main Generator per 2-OHP-4022-053-002 Emergency Degassing Of The Electrical Generator c. Start All Fire Pumps per 12-OHP-4021-066-001, Fire Protection System (Water) Operation. d. Shut down the Generator Condition Monitor at 2-GCM-AARP-11, Generator Condition Monitor Auto Alarm Remote Panel

b. CORRECT. The major concern is the fire. The source of the fire is the H2 gas from the generator. The first action must be to eliminate the source of the H2 by degassing the main generator.

Given the following conditions on Unit 2: • A plant heatup is in progress following a forced shutdown • Reactor coolant average temperature at 450°F when the following Auxiliary Feedwater (AFW) trains become inoperable: • 0100 on July 7, TDAFW train is declared inoperable due to 2-MCM-221 steam supply valve being inoperable. • 1830 on July 8, 2E AFW train is declared inoperable. • 1900 on July 8, 2-MCM-221 steam supply valve is restored to OPERABLE status. Including any extensions permitted by TS, and without re-entering a Technical Specification condition requiring a plant shutdown, the 2E AFW train must be restored to OPERABLE status by __________________. Reference Provided a. 0100 on July 10. b. 0100 on July 17. c. 1830 on July 11. d. 1830 on July 12.

c. CORRECT. 2E Pump 72 hour clock started when the 2E pump became INOPERABLE. The clock does not get reset with the OPERABILITY of the TDAFP

Given the following conditions on Unit 1: • The plant is operating at 6% power preparing for Turbine roll. • NLP-151, PZR Level Channel 1 failed 4 hours ago. The bistables have been tripped and all actions are complete as per 1-OHP-4022-013-010, Pressurizer Level Instrument Malfunction. • PZR level is currently 40% on the remaining PZR Level channels. Which ONE of the following describes the effects on the plant if NLP-153, PZR Level Channel 3 fails low and the affect on Unit Supervisor's decision to trip bistables for the Channel 3 failure? Note: Assume NO operator actions. Reference Provided a. Letdown will Isolate and heaters will de-energize. Bistables may be tripped without causing a reactor trip. Power must remain less than 10%. b. Letdown will remain in service and heaters will de-energize. Bistables should NOT be tripped since a reactor trip will be generated. Power must remain less than 10%. c. Letdown will remain in service and heaters will de-energize. Bistables may be tripped without causing a reactor trip. Power must be reduced to less than 5%. d. Letdown will Isolate and heaters will de-energize. Bistables should NOT be tripped since a reactor trip will be generated. Power must be reduced to less than 5%.

a. CORRECT. With Channel 1 NLP-151 in the tripped condition the High level Rx Trip signal will be made up for 1 channel (1/2 coincidence on remaining channels). The level control selector switch for the pressurizer is in the 2/3 position with channel 3 NLP-153 as the controlling channel. When it fails low letdown will isolate and the heaters will de-energize. When the bistables are tripped, a reactor trip signal will be generated but it is blocked by P-7 (Reactor and Turbine power both below 10%). Plant startup can NOT continue. Power must be maintained below 10% (P-7).

Given the following conditions on Unit 1: • Unit is in Mode 6 • Refueling cavity is filled to 24.2 feet above flange • Core reload is in progress • RCS temperature is 82°F • 1W RHR is OOS due to an Oil Leak • 1E RHR train is in operation A leak has been reported on the 1E RHR pump mechanical seal heat exchanger. To repair the leak, the RHR pump must be stopped. Maintenance estimates it will take 40 minutes to complete repairs. 1) How does this affect the ability to continue core reload? 2) What is the basis for having one RHR loop in operation in this condition? a. 1. Core reload must be stopped. 2. Provides for adequate RCS mixing and control of reactor coolant temperature. b. 1. Core reload must be stopped. 2. Ensures that a core Keff of less than or equal to 0.95 is maintained during fuel handling operations. c. 1. Core reload may continue provided no operations are permitted that would dilute the refueling cavity boron concentration. 2. Provide for adequate RCS mixing and control of reactor coolant temperature. d. 1. Core reload may continue provided no operations are permitted that would dilute the refueling cavity boron concentration. 2. Ensures that a core Keff of less than or equal to 0.95 is maintained during fuel handling operations.

a. CORRECT. Because both pumps are inoperable, core loading must stop. If the pump were not inoperable, core loading is allowed to continue without RHR for up to one hour provided no change in Boron Concentration. This allows loading near the edges of the core where flow may interfere with setting fuel assemblies. The indication and control of temperature is one of the bases for this LCO. Maintaining core Keff =0.95 is the bases of LCO 3.9.1 which is boron concentration during refueling operations. The confusion is that the RHR bases is for MIXING of the borated water to prevent potential criticality.

Given the following conditions on Unit 2: • A normal plant startup and power escalation to 100% was initiated following a refueling outage. • The reactor achieved 100% rated power (3468 MWth) approximately 4 hours ago. • MTI has informed the Unit Supervisor the Blowdown flow instrument for SG21 is reading approximately 100 gpm higher than actual blowdown flow. The Unit Supervisor should: a. direct the control room operator to raise power slightly since actual thermal power is less than PPC calculated thermal power of 3468 MWth. b. direct an immediate power reduction to ensure the 1 hour average does not exceed 3468 MWth. c. order that no reactor power adjustments be made for the next 1 hour so an accurate 1-hour power average is obtained. d. order that no reactor power adjustments be made for the next 4 hours and then make adjustments to power as required.

b. CORRECT. The blowdown flow error will result in PPC derived thermal power to be nonconservatively low. This may result in actual thermal power already being greater than 3468 MWth. Power must be reduced to ensure plant is operating that less than 3468 MWth actual power.

Given the following conditions in Unit 2: • The unit was stable at 95% power with Rod Control in AUTO • Control Bank D is at 220 steps with AFD at -6.5% • A HIGH failure of Power Range channel N41 occurred • The reactor operator responded by placing Rod control in MANUAL 15 seconds after the event. Which ONE of the following may require prompt crew actions to ensure continued compliance with Technical Specifications? Reference Provided a. Axial Flux Difference (AFD) b. Quadrant Power Tilt Ratio (QPTR) c. Rod Insertion Limits (RIL) d. Shutdown Margin (SDM)

a. CORRECT. AFD is close to the -7.2 limit of TDB 2-Figure 13.1, for 95% power. Based on the The operator response time and the maximum rod speed of 72 steps/minute, the rods would travel approximately 18 steps into the core. Rod motion would place the AFD outside the target band which is has a 15 minute requirement to restore AFD within the band or be below 90% in the next 15 minutes. TS 3.2.3, Axial Flux Difference (AFD)

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) 1600 420,000 6.2 1630 420,000 6.2 1700 350,000 5.8 1730 280,000 5.3 1800 260,000 4.8 1830 120,000 4.6 1900 90,000 4.3 1930 90,000 4.0 2000 90,000 3.9 At 2000 hours, while performing Emergency Operating Procedures, a step is encountered which states "Check PZR level - GREATER THAN 20% [24% ADVERSE]". Which ONE of the following describes the required Pressurizer level and why? a. 20% because the Containment Radiation levels are no longer above the Adverse setpoint requirement. b. 24% because adverse values must be used until evaluated for lasting effects because the integrated dose limit has been exceeded. c. 24% because adverse containment exists due to the current containment radiation dose rate. d. 20% because the Containment Pressure is no longer above the Adverse setpoint Requirement

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. The integrated dose is (1,015,000 R) which is greater than 106 R, so adverse containment values must be used.

Which ONE of the following responses correctly reflects the bases for Reactor Coolant Specific Activity in Technical Specifications? a. The short-lived radioactive isotope fission products will have decayed prior to any fuel movement. b. Limitations on specific activity in the RCS reduces corrosion product activation and subsequent RCS integrity challenge. c. During a LOCA the dose will NOT exceed the 10CFR20 limits at the site boundary. d. During a steam generator tube rupture the release of activity through the atmospheric relief valves will NOT exceed 10CFR100 limits.

d. CORRECT. The maximum dose to the whole body and the thyroid that an individual at the site boundary can receive for 2 hours during an accident is specified in 10 CFR 100 (Ref. 1). The limits on specific activity ensure that the doses are held to a small fraction of the 10 CFR 100 limits during analyzed transients and accidents. The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident.

Given the following conditions on Unit 2: • A LOCA has occurred • The STA is monitoring the Critical Safety Functions and notes the following indications: • WR log power 0% • WR startup rate Negative • Containment Pressure 13 psig • CETC’s 5 highest 760°F • RVLIS Narrow Range 76% • Pressurizer Level 0% • RCS Pressure 480 psig • AFW Flow 300 x103 pph SG21 SG22 SG23 SG24 • NR SG Levels 12% 15% 16% 12% Given the conditions described above, to which ONE of the following procedures should the SRO transition? a. 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling b. 2-OHP-4023-FR-I.2, Response to Low Pressurizer Level c. 2-OHP-4023-FR-Z.1, Response to High Containment Pressure d. 2-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink

c. CORRECT. Containment Pressure of >12 psig is a RED path requiring 2-OHP-4023-FR-Z-1.

Given the following conditions on Unit 2: • The unit has tripped and experienced a safety injection. • While performing 2-OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization, an ORANGE path condition was noted for the Core Cooling Critical Safety Function. • 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling, was entered. While performing steps of this procedure, the Shift Technical Advisor reports the following: • RED path condition exists for Core Cooling Critical Safety Function. • RED path condition exists for Containment Critical Safety Function. • NO other abnormal conditions were noted. Based on these plant conditions, which ONE of the following is the appropriate action for the Unit Supervisor to take? a. Complete actions of 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling, and then transition to 2-OHP-4023-FR-Z.1, Response to Containment High Pressure. b. Complete actions of 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling, and then transition to 2-OHP-4023-FR-C.1. c. Stop performing 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling, and immediately transition to 2-OHP-4023-FR-Z.1, Response to Containment High Pressure. d. Stop performing 02-OHP-4023-FR-C.2, Response to Degraded Core Cooling, and immediately transition to 2-OHP-4023-FR-C.1, Response to Inadequate Core Cooling.

d. CORRECT. Per OHI-4023, Rules of usage, when a Red path is encountered, immediately initiate the FRP. Because FR-C. 1 is a higher priority than FR-Z. 1, the US should proceed to FR-C. 1 vs. FR-Z. 1.

2010 NRC Flashcards

(100 cards)