ML13191A886

Initial Exam 2013-301 Final RO Written Exam

ML13191A886
Person / Time
Site:	Farley
Issue date:	07/10/2013
From:	NRC/RGN-III
To:	Southern Nuclear Operating Co
References
Download: ML13191A886 (302)

Text

ES-401 Site-Specific RO Written Examination Form ES-401-7 lCover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination lApplicant Information Name:Date:Facility/Unit:

Region:I II III IV lReactor Type: W CE BW GE llStart Time:

Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade lof at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

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

______________________________________

Applicant's Signature ResultsExamination Value __________ Points Applicant's Score__________ Points Applicant's Grade__________ Percent

1. Unit 1 is operating at 85% pow er with the following conditions:

• STP-33.0B, Solid Stat e Protection System Tr ain B Operability Test, is in progress.

• The 'B' Reactor Trip Bypass Br eaker has been racked in and closed.

• Control Bank D is at 203 steps.

• The Rod Control Bank Selector Switch is in AUTO.

• PS/446Z, FIRST STG IMPULSE PRESS SEL SWITCH, is in theChannel IV / PT447 position.

Subsequently, the following occurs:

• PT-447, TURB FIRST STG PRESS, fails HIGH.

Which one of the following co mpletes the statements below?

The control rods will (1) .If the reactor is manually tripped at this time, the

'B' Reactor Trip Bypass Breaker (2) light will be LIT.

(1)

(2) insert RED insert GREEN withdraw GREEN withdraw RED A.B.C.D.

2. Unit 1 is in Mode 3 and pr eparing to start the 1C RCP.

Which one of the following completes the statements below per SOP-1.1, Reactor Coolant System?

The 1C RCP oil lift pump handswit ch white light indicates (1) .The 1C RCP breaker closing operation (2) interlocked with a 2 minutetime delay. 1) BOTH the oil lift pre ssure has reached 600 psig AND 2 minutes have elapsed2) is NOT 1) ONLY that the oil lift pressure has reached 600 psig

2) IS

1) ONLY that the oil lift pressure has reached 600 psig2) is NOT

1) BOTH the oil lift pre ssure has reached 600 psig AND 2 minutes have elapsed

2) IS A.B.C.D.

3. Unit 1 was operating at 100%

power when the following occurred:

• The air supply to FCV-122, CHG FLOW REG, actuator has been severed and the valve has repositi oned to its failed position.

Which ONE of the following co mpletes the statement below?

FCV-122 is (1) and RCP seal injection flow will (2) . (1)

(2) CLOSED LOWER CLOSED RISE OPEN LOWER OPEN RISE A.B.C.D.

4. The following conditio ns exist on Unit 1:

• The operating crew is cooling down per UOP-2.2, Shutdown of Unit From Hot Standby to Cold Shutdown.

• MODE 5 has just been entere d and the following conditions exist: - RCS temperature is 195°F.

- BOTH 1A AND 1B RHR pumps are running in the cooldown mode.

Subsequently, PT-402, 1C LOOP RCS PRESS, fails HIGH.

Which one of the fo llowing describes the minimum required action(s), if any, to be performed per AOP-12.0, Residual Heat Removal Malfunction?

No actions are required.

Secure the 1A RHR pump ONLY.

Secure the 1B RHR pump ONLY.Secure BOTH 1A AND 1B RHR pumps.

A.B.C.D.

5. Unit 1 was operating at 100% power when the following conditions occurred:

• A LOCA is in progress.

• The operating crew is performing the actions of EEP-0.0, Reactor Trip or Safety Injection, and is at the step to "Check RCS intact".

• RCS pressure is 475 psig and lowering.

Which one of the following describes the current status of the ECCS system?

SI Accumulator Level

RHR Injection Flow Stable and on-scale Zero Dropping or off-scale low Zero Dropping or off-scale low Rising Stable and on-scale Rising A.B.C.D.

6. Unit 1 has experienced a Reac tor Trip and Safety Injection due to a faulted SG.

The following conditions exist:

• The operating crew is performing EEP-2.0, Faulted Steam Generator Isolation.

• SCMM is in the CETC mode.

• RCS pressure is 1900 psig and rising slowly.

• At the step for verifyin g SI termination criteria, the crew notes that PT-457, PRZR PRESS, has failed LOW.

Which one of the following co mpletes the statements below?

Subcooling margin calculated by A Train ICCMS will (1) .Subcooling margin calculated by B Train ICCMS will (2) . (1)

(2) be affected NOT be affected NOT be affected be affected be affected be affected NOT be affected NOT be affected A.B.C.D.

7. Unit 2 was operating at 2%

power with a plant st artup in progress pe r UOP-1.2, Startup of the Unit from Hot Standby to Minimum Load and the following conditions occurred:

At 1000:

• DG-15-2, 2B S/U XFMR TO 2G 4160V Bus, trips open.

At 1005:

• DF-01-2, 2A S/U XFMR TO 2F 4160V Bus, trips open.

Which one of the following co mpletes the statements below at 1006 with no operator actions taken?

The Reactor Trip breakers will be (1) . DRPI rod bottom lights (2) be LIT.

(1)

(2) OPEN WILL OPEN will NOT CLOSED WILL CLOSED will NOT A.B.C.D.

8. Unit 1 is in Mode 5 and fo rming a pressurizer steam s pace (drawing a bubble) per UOP-1.1, Startup of Unit from Cold Shutdown to Hot Standby.

The vacuum refill procedure will NOT be performed.

The following conditions exist:

• RCS pressure is being maintained at 325-375 psig.

• 1B RCP is running.

• 'A' Train RHR is on service wit h low pressure letdown aligned.

• RCS is in solid plant pressure control.

• Pressurizer temperature is 178°F and slowly rising.

• All PRZR heater s have been energized.

Which one of the following completes the statements below?

Per UOP-1.1, the pressurizer is at saturation conditions when (1) increases.

During this evolution, PRT level will (2) . (1)

(2) charging flow remain constant letdown flow remain constant letdown flow rise charging flow rise A.B.C.D.

9. Unit 2 was operating at 100%

power when a Reactor Trip occurs and the following conditions exist:

• Q2B13PSV8010A, PZR SAFETY, has failed OPEN.* Pressurizer pressure is 1020 psig.

• PRT pressure rises to 55 psig.

• Core Exit Ther mocouples read 560°F.

Which one of the following co mpletes the statements below?

Temperature on TI-469, SAFETY VLVS, will indicate approximately (1) .Pressurizer level will be (2) .Reference provided (1)

(2) 546°F rising 546°F lowering 320°F lowering 320°F rising A.B.C.D.

10. Unit 2 is operating at 100% power when the following occurs:

• A simultaneous dual Unit LOSP occurs:Which one of the following completes the statement below?The (1) CCW pump is being powered by the (2) DG.

(1)

(2) 2A 1-2A 2A 1C 2C 1-2A 2C 1C A.B.C.D.

11. The following conditions exist on Unit 1:

• An RCS leak is in progress.

• RCS pressure is 1600 psig and stable.

• Containment pressure is 3.1 psig and slowly rising.

• The crew has just transitioned to ESP-1.2, Post LOCA Cooldown and Depressurization.

Which one of the following completes the statement below? The minimum SG narrow range water le vel must be greater than (1) to (2) .1) 31%

2) ensure SG tubes are cover ed to promote reflux boiling

1) 48%2) ensure SG tubes are cover ed to promote reflux boiling

1) 31%2) ensure adequate SG inventory to provid e a secondary heat sink

1) 48%

2) ensure adequate SG in ventory to provide a secondary heat sink A.B.C.D.

12. Unit 1 was operating at 100% power when a Reactor Trip and LOSP occurred.

The following conditions exist:

• The 1A PZR HTR GROUP BACKUP handswitch is in AUTO.

• RCS pressure is 2000 psig.

Which one of the following correctly de scribes Pressurizer Heater operation per ESP-0.1, Reactor Trip Response?

The 1A PZR Heaters will have power available, NO actions are requ ired to energize them.The 1A PZR Heaters will have power available AND manual actions on the MCB are required to energize them.

The 1A PZR Heaters will NOT have power available. Manu al actions are required to align power to t hem on the EPB but NO other actions are requi red to energize them.The 1A PZR Heaters will NOT have power available. Manu al actions are required to align power to them on the EPB AND manual actions on t he MCB are required to energize them.

A.B.C.D.

13. Unit 1 has experienced a Loss of Off-site Power and a Large Break LOCA. The following conditions exist:

• ESP-1.3, Transfer to Cold Leg Recirculation, has been completed.

Subsequently, the Shift Supervi sor directs the OATC to pe rform ESP-1.4, Transfer to Simultaneous Cold and Hot Leg Reci rculation and the following occurs:

• Power is lost to the 1G 4160V Bus and will not be rest ored for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.Which one of the following completes the statement below?At the completion of ESP-1.4, the running LHSI pump will be aligned for (1) leg recirculation and the running HHSI pump will be aligned for (2) leg recirculation.

(1)

(2) HOT COLD COLD COLD HOT HOT COLD HOT A.B.C.D.

14. Unit 1 is operating at 100% power when the 1E 4160V bus becomes de-energized dueto an electrical fault.Which one of the following completes the statement below?

Pressurizer heater groups have lost their normal power supply.

1C and 1D ONLY 1A, 1C and 1D ONLY 1B and 1E ONLY 1B, 1D and 1E ONLY A.B.C.D.

15. The following conditions exist on Unit 1:

• A loss of 'A' Train Auxiliary Building 125V DC Bus has occurred.Which one of the following completes the statement below?

Placing the MCB Reactor Trip handswitch in TRIP would

if they were closed.

open ALL reactor trip and bypass breakers ONLY open the 'B' reactor trip breaker and the 'B' reactor trip bypass breaker ONLY open the 'B' reactor trip breaker and the 'A' reactor trip bypass breaker open BOTH reactor trip breakers but NOT open either reactor trip bypass breaker A.B.C.D.

16. The Integrated Plant Computer Display on the following page is provided for evaluation of this question.

Unit 1 has experienced a reactor trip and the fo llowing conditions exist:

• The operating crew is verifying t he immediate operator actions per EEP-0.0, Reactor Trip or Safety Injection.

• MLB-1, 1-1 and 11-1, SAFE TY INJECTION, are NOT LIT.

The STA reports the following indi cations on the Plant Computer:

• PT0455 PRESSURIZER PRE SSURE CHAN 1 is 1841 psig.

• PT0456 PRESSURIZER PRE SSURE CHAN 2 is 1855 psig.

• PT0457 PRESSURIZER PRE SSURE CHAN 3 is 1845 psig.

• PT0444A PRESSURIZER PRESSURE CHAN 4 is 1857 psig.

• PT0445A PRESSURIZER PRESSURE CHAN 5 is 1855 psig.

• PT0464 STEAM HEADER PRESSURE is 6.4 psig. Which one of the following co mpletes the statements below?A Safety Injection (1) required.

MSIV-3370A, B and C (2) OPEN.Reference Provided (1) (2) IS are NOT IS ARE is NOT are NOT is NOT ARE A.B.C.D.

17. Unit 1 is performing a reactor startup per UOP-1.2, Startup Of Unit From Hot Standby To Minimum Load, when the following conditions occurred:

• The OATC pulled Control Bank D to 100 Steps by St ep Demand Counter.* Rod B8 was noted to be indicating 54 Steps by DRPI.

Which one of the following co mpletes the statements below?

Rod B8's position is (1) .Per Tech Spec Bases 3.1.7, Rod Position Indication, (2) is(are) the most reliable indication.

1) exactly 100 steps

2) the group step counters

1) approximately 100 steps2) the group step counters

1) exactly 54 steps2) DRPI

1) approximately 54 steps2) DRPI A.B.C.D.

18. Unit 1 is operating at 100% power.

The following occurs:

• MOV-3052, CCW TO RCP CLRS, closes.

• DD3, CCW FLOW FROM RCP OIL CLRS LO, comes in to alarm.

Which one of the following co mpletes the statements below?

The most limiting components for this event are the RCP (1) .The RCPs will be required to be stopped within approximately (2) . 1) Motor Bearings

2) 2 minutes

1) Motor Bearings

2) 60 minutes

1) Pump Lower Radial Bearings

2) 2 minutes1) Pump Lower Radial Bearings

2) 60 minutes A.B.C.D.

19. Unit 1 has experienced a Reactor Trip and SI due to a LOCA and the following conditions exist:

• The operators have transitioned to EEP-1.0, Loss of Reactor or Secondary Coolant.

• The Core Exit Thermocouples (CETCs) are reading as follows: - TWO CETCs are indicating a SHORT circuit. - THREE CETCs are 1204°F and rising.

- All other CETCs are reading between 950°F and 1150°F and rising.

Which one of the following co mpletes the statements below?

The indication for the SHORT circuited CETCs fail (1) . The (2) CETC is used to evaluate entry into FRP-C.2, Response To DegradedCore Cooling.

(1)

(2) HIGH hottest HIGH 5 th hottest LOW hottest LOW 5 th hottest A.B.C.D.

20. Unit 1 is operating at 100% power with the following conditions:

At 1000:

• Containment Pressure is 0 psig.

• Containment temperature is 100°F.

• MI-3319A, B, C and D, CTMT CLR SUPP AIR MOISTURE, indicate 72 Dewpoint °F.

• Containment Coolers are running in slow speed.

At 1015: A LOCA occurs and the following conditions exist:

• Containment Pressure is 5 psig.

• Containment temperature is 150°F.

• MI-3319A, B, C and D, CTMT CLR SUPP AIR MOISTURE, indicate 130 Dewpoint °F.

Which one of the following comp letes the statements below at 1015?The Containment Cooler discharge will be through the (1) .The Containment Cooler fans will be drawing (2) amps than at 1000.

(1)

(2) ductwork MORE dropout plate MORE ductwork LESS dropout plate LESS A.B.C.D.

21. Unit 1 is operating at 100% power and the following conditions exist:

• AOP-16.0, CVCS Malfunction, has just been exited after a charging flow controller failure.

• FK-122, CHG FLOW, is in MANUAL and has been repaired.

Subsequently, FK-122 is placed in AUTOMATIC and the follo wing conditions exist:

• One 60 gpm orifice is on service.

• Charging flow is stable at 62 gpm.Which one of the following completes the statement below?

If FK-122 were to go to minimu m demand, chargi ng flow would decrease to a minimum flow rate of (1) , which is designed to prevent (2) . 1) 18 gpm 2) flashing downst ream of the letdown orifices

1) 18 gpm 2) overheating of the charging pumps

1) 40 gpm 2) flashing downst ream of the letdown orifices

1) 40 gpm

2) overheating of the charging pumps A.B.C.D.

22. Unit 1 is operating at 1 00% power when a Steam Break occurs on 1B SG and the following conditions exist:

• EE5, CTMT ISO PH B, is in alarm.

• All Phase B automat ic actions have occurred.

Which one of the following co mpletes the statements below?CCW to the RCP Thermal Barrier Heat Exchanger (1) isolated.

Seal Injection (2) isolated.

(1)

(2) IS is NOT is NOT is NOT IS IS is NOT IS A.B.C.D.

23. Unit 1 is in Mode 5 with the following conditions:

• 1B RHR pump is tagged out.

• All SG Wide Range levels are 84%.

• Pzr level is being maintained at 21% on LI-462, PRZR LVL.

• RCS temperature is 155°F.

• RCS pressure is 325 psig.

• All RCP's are secured.

• 1A RHR pump is running in the cooldown lineup.

Subsequently, the following occurs:

• 1A RHR pump trips on over current and cannot be restarted.

• RCS temperature is 175°F and slowly rising.

Which one of the following co mpletes the statements below?

Per AOP-12.0, Residual Heat Removal System Malfunction, the preferred method to re-establish core cooling is to establish (1) .Core cooling is monitored using (2) . (1)

(2) feed and bleed RCS co ld leg temperatures a secondary heat sink RCS cold leg temperatures feed and bleed CETCs a secondary heat sink CETCs A.B.C.D.

24. Unit 1 is operating at 100% power when the following occurs:

• A leak develops in the CCW system.

• CCW Surge Tank level is slowly lowering.

• AA4 and AB4, CCW SRG TK LVL A(B) TRN HI-LO, are in alarm.

• AA5, CCW SRG TK LVL A TRN LO-LO, has come into alarm.

Which one of the following co mpletes the statements below?CCW system automatic isolations are designed to occur at (1) in theCCW Surge Tank.

Using the NORMAL source of mak eup water, the op erator will open (2) to make up to the CCW Surge Tank.

(1)

(2) 35 inches MOV-3031A, MKUP TO CCW FROM RMW 20 inches MOV-3031A, MKUP TO CCW FROM RMW 35 inches MOV-3030A, MKUP TO CCW FROM DW STOR TK 20 inches MOV-3030A, MKUP TO CCW FROM DW STOR TK A.B.C.D.

25. Unit 2 is operating at 50% power when a simultaneous Dual Unit LOSP occurs.

• 4160V Bus 2G remains de-energized due to the DG not starting for that emergency bus.

Three (3) minutes after the LOSP, a Large Break LOCA occurs on Unit 2.

• Containment pressure peaked at 29 psig and is trending down.Which one of the following completes the statement below?The (1) Containment Spray pump is currently running and is powered from the (2) DG. (1) (2) 2A 1C 2B 1C 2A 1-2A 2B 1-2A A.B.C.D.

26. Unit 2 plant conditions are as follows:* Containment Main Purge system is running.* Containment radiation levels are rising.Subsequently, R-24A, CTMT PURGE, loses control power.

Which one of the following co mpletes the statements below?

Radiation levels (1) stop rising in the Main Exhaust Plenum.

CTMT Main Purge supp ly and exhaust fans (2) trip.

(1)

(2) WILL will NOT will NOT will NOT WILL WILL will NOT WILL A.B.C.D.

27. Unit 1 is operating at 100% power and the following conditions exist:

• A blended make-up to the Spent Fuel Pool (SFP) is occurring.

• A calibration error results in FT-168, PRI WATER MKUP FLOW, providing a flow input to the Reac tor Makeup System that is less than the actual flowrate.

Which one of the following co mpletes the statements below?

The blended flow makeup resulted in a (1) of the SFP.

Per Tech Spec 3.7.14, Fuel Stor age Pool Boron Concentration, theMINIMUM required SFP bo ron concentration is (2) ppm. (1)

(2) boration 2000 boration 2200 dilution 2000 dilution 2200 A.B.C.D.

28. Unit 1 is operating at 100% power with t he following conditions:

• Reactor power is now 100.5% and slowly rising.

• Tavg is 570.5°F and slowly lowering.

• Pressurizer pressure is 2210 psig and slowly lowering.

• Turbine load is 890 MWe and lowering.

• SG pressures are 720 psig and slowly lowering.

• Containment pressure is 2.1 psig and slowly rising.

Which one of the following co mpletes the statements below?

The event in progress is a (1) line break.

Per AOP-14.0, Secondary Syst em Leakage, the operators are required to (2) . (1) (2) steam reduce turbine load steam trip the reactor feed reduce turbine load feed trip the reactor A.B.C.D.

29. Unit 1 is operating at 100%. A fuel shuffle is being performed in the Unit 1 SFP.

At 1000:

• EH2, SFP LVL HI-LO, is in alarm.

• SFP level is 153' 3" and stable. At 1015: The SRO in charge of refueling repor ts that a fuel assembly has been dropped.

• FH5, SFP AREA RE-25 A OR B HI RAD, is in alarm.

• R-25A & B, SPENT FUEL BLDG EXH, reads off scale high. Which one of the following co mpletes the statements below?

The operating crew is required to enter (1) .The crew is required to dispatch personnel to (2) per the applicable AOP.

Procedure titles are as follows: AOP-30.0, Refueling Accident AOP-49.3, Spent F uel Pool Emergency (1) (2) AOP-49.3 make up to the SFP using the RWST AOP-30.0 make up to the SFP using the RWST AOP-49.3 ensure all SFP hat ches and doors are closed AOP-30.0 ensure all SFP hatches and doors are closed A.B.C.D.

30. Unit 1 is performing the actions of AOP-2.0, Steam Generator Tube Leakage, due to a tube leak on the 1A SG. T he following conditions exist:

• RCS pressure is currently bein g reduced to minimize break flow.

The following parameters are observed:

• SG pressures are:

1A SG 1B SG 1C SG 948 psig 905 psig 900 psig - RCS pressure is 916 psig. - The highest reading non-upperhead CETC is 518°F. - PRZR level is 43%.

• BOTH Subcooled Margin Monitors are malfunctioning.

Which one of the following co mpletes the statements below?

The current value of subcooling is approximately (1) .The RCS pressure reduction (2) required to be stopped.Reference Provided (1)

(2) 22°F IS 22°F is NOT 18°F IS 18°F is NOT A.B.C.D.

31. Unit 1 has experienced a tube rupture on the 1C SG.

The operating crew is at the step in EEP-3.0, Steam Generator Tube Rupture, to "Check SI termination criteria."

• The following plant conditions are observed: - RCS Subcooling is 22°F and slowly rising.

- RCS pressure is 950 psig and slowly rising. - Pressurizer level is 45% and slowly rising.

- AFW flow is 450 gpm.

- 1A SG NR level is 29% and slowly rising.

- 1B SG NR level is 26% and slowly rising.

- 1C SG NR level is 65% and rising rapidly.

Which one of the following co mpletes the statements below?

SI termination criteria (1) been met.Per EEP-3.0, SI termination is nece ssary to prevent overfilling the (2) . (1)

(2) has NOT Steam Generator has NOT Pressurizer HAS Steam Generator HAS Pressurizer A.B.C.D.

32. Concerning R-70A/B/C, 1A/1B/1C SG TUBE LEAK DET, on Unit 1:

Which one of the following co mpletes the statements below?

The R-70s are located (1) of the MSIVs.

A minimum reactor power level that the R-70s can accurately estimate a SG leak rate is (2) . (1)

(2) upstream 25% downstream 25% upstream 10% downstream 10%A.B.C.D.

33. Which one of the following coincidences will cause an anticipated transientwithout trip (ATWT) mitigation system actuation circuitry (AMSAC) Main TurbineTrip? (1) Turbine impulse pressure channels > 40%

AND (2) SG NR levels < 10% for > 25 seconds.

(1)

(2) 1 of 2 2 of 3 2 of 2 2 of 3 1 of 2 1 of 3 2 of 2 1 of 3 A.B.C.D.

34. Unit 1 is operating at 40% power when PR-4029, CONDENSER PRESSURE indicates as follows:

• PT0501 and PT0502 are 6 ps ia and rising rapidly.

Subsequently, Condenser pressure stabilizes at 12 psia.

Which one of the following co mpletes the statements below?

The Steam Dump (1) controller is enabled.

The Steam Dumps are (2) . (1) (2) Plant Trip CLOSED Plant Trip OPEN Loss of Load CLOSED Loss of Load OPEN A.B.C.D.

35. Unit 1 is operating at 4% power. The following conditions exist:

• 1A SGFP is running.

• All SG NR levels ar e in the programmed band.

• FCV-479/489/499,1A/1B/1C SG FW BYP FLOW, controllersare in MANUAL and 35% open.

Subsequently, the 1A SGFP trips.

Which one of the following co mpletes the statements below?

MOV-3232A/B/C, MAIN FW TO 1A/1B/C SG, will (1) .FCV-479/489/499,1A/1B/1C SG FEED FLOW B YPASS FCVs, will (2) . (1)

(2) remain OPEN remain OPEN remain OPEN CLOSE CLOSE remain OPEN CLOSE CLOSE A.B.C.D.

36. Unit 1 is at 70% power with the following conditions:

• R-15A, SJAE EXH, is in alarm.

• R-15B, TURB BLDG VNTL , is in alarm.

• AOP-2.0, Steam Generator Tube Leakage, is in progress.

• The Turbine Building SO has placed the SJAE Filtration System in service.Which one of the following completes the statement below?After the SJAE Filtration system is placed in service, the reading on R-15B will (1) and the SJAE Filtration system will (2) .

(1)

(2) decrease be aligned in a recirc alignment

remain the same be aligned in a recirc alignment decrease discharge to the Turbine Bldg roof remain the same discharge to the Turbine Bldg roof A.B.C.D.

37. Unit 1 is operating at 100% when a LOSP occurred. The following conditions exist:

• The Emergency Diesel Generators failed to energize the ESF busses.

• The operating crew is conducting a secondary depressurization per ECP-0.0, Loss Of All AC Power.

• SG pressures are as follows:

- 1A SG: 245 psig and lowering

- 1B SG: 247 psig and lowering

- 1C SG: 244 psig and lowering Which one of the following co mpletes the statements below?

Per ECP-0.0, this secondary pre ssure reduction is required to (1) .The reason the secondary pressure r eduction is requir ed to be stopped at the SG pressure specified in ECP-0.0 is to prevent (2) . (1)

(2) be STOPPED injection of accumulator nitrogen into the RCS CONTINUE injection of accumulator nitrogen into the RCS be STOPPED a challenge to the Integrity Critical Safety Function CONTINUE a challenge to the Integrity Critical Safety Function A.B.C.D.

38. Unit 1 is operating at 100% power when the 1B SGFP trips.

Which one of the following completes t he statements below for the 1B SGFP?The HIGH PRESS. GOV. VA LVE CLOSED light is (1) . The LOW PRESS. GOV. VALVE CLOSED light is (2) .

(1)

(2) LIT LIT LIT NOT lit NOT lit LIT NOT lit NOT lit A.B.C.D.

39. Unit 2 is operating at 100% power when a SG tube leak occurs.Which one of the following completes the statement below?

When R-23A, SGBD HX OUTLET , alarms, it will cause

, to automatically close.HV-7614A/B/C, 2A/B/C SGBD ISO FCV-1152, SGB INLET STOP VALVE RCV-023B, SGBD DISCH TO ENVIRONMENT HV-7697A/B, 7698A

/B and 7699A/B, 2A/B/C SGBD ISO A.B.C.D.

40. Unit 1 is operating at 33% power and the following conditions exist:

• 1A and 1B Condensat e pumps are running.

• 1C Condensate pump is in OFF with a CAUTION TAG that says, "EMERGENCY USE ONLY."

• 1A SGFP is running.

Subsequently, the 1B Condensate pump tr ips and the following conditions are observed:

KB4, SGFP SUCTION PRESS LOW, come s into alarm and the operating crew observes the following on PR 4039, SGFP SUCT PRESS:

Time 0 sec 10 sec 20 sec 30 sec 40 sec300 psig 275 psig 265 psig 270 psig 285 psigAt time 20 seconds, the 1C condensat e pump was started.

Which one of the following co mpletes the statements below?At time 30 seconds, the 1A SGFP (1) be tripped.

The operating crew is required to (2) .1) will NOT2) rapidly reduce Turbi ne load using AOP-17.1, Rapi d Turbine Power Reduction1) will NOT2) check SGFP suction pressure stabilizes

1) WILL2) trip the Reactor and enter EEP-0.0, Reactor Trip or Safety Injection.

1) WILL

2) trip the Main Turbine and enter AOP-3.0, Turbine Trip Below P-9 Setpoint.

A.B.C.D.

41. Unit 1 is in Mode 3 with the following conditions:

• 1A MDAFW pump was star ted per UOP-1.2, Startup of Unit From Cold Shutdown to Hot Standby.

• There are no other AFW pumps running.

• All SG NR levels are 65%.

Subsequently, power is lost to the 1A Startup Transformer.

Which one of the following co mpletes the statements below?The TDAFW Pump (1) be running.

Total design AFW flow rate w ill be approximately (2) gpm. (1)

(2) will NOT 350 will NOT 700 WILL 700 WILL 1050 A.B.C.D.

42. Unit 1 was operating at 100% power when a Reactor Trip occurred and the following conditions exist:

• ESP-0.1, Reactor Trip Re sponse, has just been entered.

• Pressurizer level is 12% and slowly lowering.

• SG NR levels are 40% and slowly rising.

• Tavg is 534°F and slowly lowering.

• RCS pressure is 2050 psig and slow ly lowering.

Which one of the following actions will be performed FIRST as required by ESP-0.1 to address the cooldown?

Minimize total AFW flow.

Emergency borate the RCS.Close all MSIVs and MSIV Bypass Valves.Manually initiate SI and return to EEP-0.0, Reactor Trip or Safety Injection.

A.B.C.D.

43. Unit 2 is operating at 100% power. The following conditions exist:

• SGBD is on service. * #1 WMT release is in progress.

• The service water pond level ha s dropped to 179 feet, 10 inches.

Which one of the following combinations predicts the plant re sponse to the change in pond level?

1) SW Dilution Flow on FR-4107, SW DILUTION FLOW, will lower;

2) RCV-023B, SGBD DISCH TO ENVIRONMENT, will automatically close.

1) SW Dilution Flow on FR-4107, SW DILUTION FLOW, will lower;

2) RCV-018, WMT DIS CH TO ENVIRONMENT, w ill automatically close.

1) SW Pressure on PI

-3001A & B, SW TO CCW HX HDR PRESS, will lower;

2) PCV-562 and 563, TR N B (A) DILUTION BYPASS PCV, will fully open.

1) SW Pressure on PI

-3001A & B, SW TO CCW HX HDR PRESS, will lower;

2) MOV-538 and 539, SW B (A) HDR EMERG RECIRC TO POND, will fully open.

A.B.C.D.

44. The following conditions exist on Unit 2:

• DG02-2, 2G 4160 V bus tie to 2L 4160 V bus, has tripped opened. Which one of the following completes the statement below?The has lost Service Water cooling.2C Instrument Air Compressor 2C Reactor Coolant Pump Motor Air Cooler 2C Component Cooling Water Heat Exchanger Steam Generator Bl owdown Heat Exchanger A.B.C.D.

45. Unit 1 has experienced a Reactor trip with the following conditions:* A Loss of All AC has occurred.* ECP-0.0, Loss of All AC Power, is in progress.

Which one of the following co mpletes the statements below?The 1B Aux Building DC bus voltage will (1) . Per ECP-0.0, there may no t be enough DC capacity to start a DG and sequence needed loads if power is not restored to the 125V DC batte ry chargers on each train within a MINIMUM of (2) .1) drop slowly at first; then later drop rapidly as t he battery nears exhaustion

2) 30 min 1) drop slowly at first; then later drop rapidly as t he battery nears exhaustion

2) 90 min

1) drop at a const ant, linear rate the entire time the battery discharges

2) 30 min

1) drop at a const ant, linear rate the entire time the battery discharges

2) 90 min A.B.C.D.

46. A loss of all AC power has occurred on Unit 1 and the following conditions exist:

• VA2, 1B DG GEN FAULT TRIP, has come into alarm.

• The crew has completed the step in ECP-0.0, Loss Of All AC Power, to verify breaker s for major loads OPEN.

• A Safety Injection occurs on Unit 1 at this time.

Which one of the following co mpletes the statements below?

The 2C DG will be started from the EPB in (1) using the START pushbutton. All ESF loads will (2) .1) Mode 2 2) automatically start

1) Mode 2 2) have to be manually aligned

1) Mode 1 2) automatically start

1) Mode 1 2) have to be manually aligned A.B.C.D.

47. Unit 1 is operating at 100% power with the following conditions:

• A problem with 1B DG starting air system has occurred.

• The B Air receiver has been tagged out.Which one of the following completes the statement below?A MINIMUM of (1) psig must be available in the remaining air receiver to ensure five (5) start attempts are available.

1B DG's required minimum time to reach rated speed and voltage is (2) seconds after receiving an em ergency start signal. (1) (2) 200 7 200 12 350 7 350 12 A.B.C.D.

48. Unit 1 was operating at 100% power when the following occurred:

• A complete loss of instrument air caused an automatic Reactor Trip.

The following conditions exist:

• All AFW pumps are running.

• All SG NR Levels are 25% and rising.

• The Shift Supervisor has directed AFW flow to be reduced.

Per AOP-6.0, Loss of Inst rument Air, which one of the following methods belowwill be successful in reducing AFW flow? Valve nomenclature:

• HV-3228A / B / C, TD AFWP TO 1A/1B/1C SG

• MOV-3764A / D / F, MDAF WP TO 1A/1B/1C SG ISO

• MOV-3350A / B / C, AFW TO 1A/1B/1C SG STOP VLV Place BOTH MDAFW pump MCB hand s witches in the STOP position and release them.Throttle HV-3228A / B / C on the MCB.

Close MOV-3764A / D / F on the BOP.

Close MOV-3350A / B / C on the MCB.

A.B.C.D.

49. There is a fire in the Control Room and the following conditions exist for Unit 1:

• FNP-1-AOP-28.2, Fire In The Control Room, has been entered.

Which one of the following co mpletes the statements below?

During the conduct of AOP-28.2, the Diesel Ge nerators are required to be placed in (1) and the output breakers (2) automatically closewhen the DGs are started after a Loss of Offsite Power.

(1)

(2) MODE 3 WILL MODE 3 will NOT MODE 4 WILL MODE 4 will NOT A.B.C.D.

50. Unit 1 is operating at 100% power with the following conditions:

• The 1A Waste Gas Compressor is running and aligned to #7 WGDT.

• R-13, WGC SUCT, alarms.

Subsequently, the #7 WGDT relief va lve lifts and fails to reseat.

Which one of the following co mpletes the statements below?

R-22, VENT STACK GAS, (1) trend up.

1. 7 WGDT relief valve (2) be manually isolated.

(1)

(2) WILL CANNOT WILL CAN will NOT CANNOT will NOT CAN A.B.C.D.

51. Unit 1 is operating at 100% when the following condition occurs:

• R-19, SGBD SAMPLE, fails HIGH.

Which one of the following co mpletes the statements below?

(1) valves will automatically close.

Per SOP-45.0, Radiation Monitori ng System, the actions required toallow the Shift Chemist to obtai n a sample of the SGs is to (2) .1)

HV-3328, HV-3329 AND HV-3330, STEAM GEN 1A/1B/1C SAMPLE ISO, 2) pull the INSTRUMENT power fuses for R-19

1) HV-3328, HV-3329 AND HV-3330, STEAM GEN 1A/1B/1C SAMPLE ISO, 2) place R-19 Operations Selector Switch to the RESET position

1) HV-3179A, 318 0A, AND 3181A, STEAM GEN 1A/1B/1C LOWERBLOWDOWN,

2) pull the INSTRUMENT power fuses for R-19

1) HV-3179A, 318 0A, AND 3181A, STEAM GEN 1A/1B/1C LOWER BLOWDOWN, 2) place R-19 Operations Selector Switch to the RESET position A.B.C.D.

52. Unit 1 is operating at 100% power.

Which one of the following meets the MINIMUM reactor coolant leakage detectionsystem(s) that must be in operation and OPERABLE to prevent entering a REQUIRED ACTION STATEMENT of Tech Spec 3.4.15, RCS Leakage Detection Instrumentation?

• R CTMT PARTICULATE

• R CTMT GAS

• Containment Air Cooler Condensate Level Monitoring System (CACCLMS)

R-11 ONLY ()R-11 AND R-12The CACCLMS ONLYR-12 AND the CACCLMS A.B.C.D.

53. Unit 1 has been operating at 100% power and the Gross Failed Fuel Detector (GFFD)has been steady at 2000 cpm during the entire fuel cycle.

At 1000:

• FG5, GFFD SYS TRBL, has just come into alarm.

At 1015:

• A Reactor Trip and Safety Injection occurs.

Which one of the following co mpletes the statements below?The minimum GFFD reading that woul d cause FG5 to come into alarm is (1) above background.

At 1020 , flow through the GFFD (2) be isolated. (1) (2) 1 X 10 4 cpm will NOT 1 X 10 4 cpm WILL 1 X 10 5 cpm will NOT 1 X 10 5 cpm WILL A.B.C.D.

54. Unit 1 is stable in Mode 3 following a Reactor Trip when the following conditions occur:

• Power has been lost to 4 160V AC buses G, J, and L. Which one of the following lists the valves that the OATC can close from the MCB to isolate a Service Water ruptur e in the Turbine Building? Q1P16V514, SW TO TURB BLDG ISO B TRN Q1P16V515, SW TO TURB BLDG ISO A TRN Q1P16V516, SW TO TURB BLDG ISO A TRN Q1P16V517, SW TO TURB BLDG ISO B TRN MOVs 514 and 517MOVs 514 and 516 MOVs 515 and 517 MOVs 515 and 516 A.B.C.D.

55. Unit 1 is operating at 100% power with the following conditions:

• 1A Containment Cooler is isolated per SOP-12.1, Cont ainment Air Cooling System.

• The following valves are closed with power available: - MOV-3019A, SW TO 1A CTMT CLR AND CTMT FPS - MOV-3441A, SW FROM 1A CTMT CLR - MOV-3024A, EMERG SW FROM 1A CTMT CLR

• MOV-3023A, 1A CTMT CLR SW DISCH, is OPEN.

Subsequently, a steam break occurs and c ontainment pressure rises to 5 psig.Which one of the following completes the statement below?

1A Containment Cooler se rvice water flow will be

.0 gpm approximately 600 gpm approximately 800 gpm approximately 2000 gpm A.B.C.D.

56. Unit 1 is at 100% power with the following conditions:

• 1B DG is running for STP-80.1, Diesel Generator 1B Operability Test.

• The 1B DG is currently loaded to 1 MW.

• DG01, 1B S/U XFMR TO 1G 4160 V BUS, breaker is CLOSED.

Subsequently, the following occurs:

• WE2, 1F, 4KV BUS OV-OR-UV OR LOSS OF DC, and VE2, 1G, 4KV BUS OV-OR-UV OR LOSS OF DC, come into alarm.

• 1F and 1G 4160V bus volt ages are reading 3825 volts.

• The crew has entered AOP-5.2, Degraded Grid.

Per AOP-5.2, which one of the follo wing completes the statements below?The 1B DG (1) .The reason for the above action is (2) .1) is required to be secur ed and aligned for AUTO START

2) because this places the 1B DG is in the most reliable condition

1) is required to be secur ed and aligned for AUTO START2) because the LOSP Sequencer will not run LOSP loads if DG01 opens with 1B DG output breaker aligned to 1G 4160V bus

1) load is required to be raised to FULL LOAD and DG01 opened

2) to ensure adequate voltage fo r safety related equipment

1) load is required to be raised to FULL LOAD and DG01 opened

2) to prevent extended low lo ad operation which would result in the buildup of combustion products in the engine exhausts A.B.C.D.

57. The following conditions exist on Unit 1:* A rupture in the Instrument Air system has occurred.* Instrument Air header pressure is 65 psig and lowering slowly.

Which one of the following co mpletes the statements below?

V-902, AIR DRYER AU TO BYP, will be (1) . V-904, NON-ESSENTIAL IA HDR AUTO ISO, will be (2) . (1)

(2) OPEN OPEN OPEN CLOSED CLOSED OPEN CLOSED CLOSED A.B.C.D.

58. The following plant conditions exist on Unit 1:

• The Reactor has been tripped du e to loss of Instrument Air.

• The operating crew is performing t he actions of ESP-0.1, Reactor Trip Response.

• SG Atmospheric Relief Valves (ARVs) are aligned per SOP-62.0,Emergency Air System.

Subsequently, the operator applies 18 psig to the valve actuator for PCV-3371A, 1A MS ATMOS REL VLV.

Which one of the following co mpletes the statements below?

PCV-3371A (1) open.If PCV-3371A were fully open, (2) .1) IS

2) a High Steam Flow - Lo Lo Tavg Main Steam Isolation may occur

1) IS 2) Technical Specification c ooldown limits may be exceeded1) is NOT 2) a High Steam Flow - Lo Lo Tavg Main Steam Isolation may occur1) is NOT

2) Technical Specification c ooldown limits may be exceeded A.B.C.D.

59. Unit 1 has experienced a large break LOCA and the following conditions exist:

• PT-953, CTMT PRESS has reached the PHASE B setpoint.Which one of the following completes the statement below?A MINIMUM of (1) OR (2) PHASE B CTMT ISO CS ACTUATION handswitch(es) is(are) required to actuate a Phase B isolation.

1) 1 additional Containment pre ssure channel reac hing 16.2 psig

2) TWO 1) 2 additional Containment pre ssure channels r eaching 16.2 psig

2) ONE

1) 1 additional Containment pr essure channel r eaching 27 psig

2) TWO

1) 2 additional Containment pre ssure channels r eaching 27 psig

2) ONE A.B.C.D.

60. The electronic log is malfunctioning. The control room has shifted to manual logs and the following entries have been made: 1000 Q1E21V061A, HHSI to 1C RCS loop CL iso, as left position; 1.5 turns OPEN. 1012 Started 1B CCW Pump.

At 1030:

• The OATC recognizes that an e rror was made on the 1000 log entry.

• Q1E21V061A should have been logged as throttled to 1.25 turns OPEN.

Per SOP-0.11, Watch Stati on Tours and Operator Logs, the OATC is required to correct the 1000 log entry by wh ich one of the following methods?

• Circle the incorrect entry in red.

• Enter the correct information ne xt to the incorrect information and record the date and initial.

• Circle the incorrect entry in red.

• At 1030 make a log entry with t he correct informat ion and designate it as a LATE ENTRY.

• Draw a single line through the incorrect entry.

• Enter the correct information ne xt to the incorrect information and record the date and initial.

• Draw a single line through the incorrect entry.

• At 1030 make a log entry with the correct informati on and designate it as a LATE ENTRY.

A.B.C.D.

61. Both Units are operating at 100% power with the following conditions:

• A non-licensed Fire Protection Administrator who is qualified as a Shift Communicator is on shift.

Which one of the following co mpletes the statements below?

Per EIP-0.0, Emergency Organization, a minimum of (1) licensed Plant Operators is required to staff the shift.The maximum number of hours that a Pl ant Operator may work in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is (2) per NMP-AD-016-003, Scheduling and Calculating Work Hours.

(1)

(2) 3 12 3 16 4 12 4 16 A.B.C.D.

62. Unit 1 is operating at 100% power when the following occurs:

• STP-4.1, 1A Charging Pump Quarterly Inservice Test, is in progress.

• 1A Charging pump failed to start when the handswitch was taken to START. Which one of the following are the required actions per SOP-0.0, General Instructions to Operations Personnel?

Obtain Shif t Manager's permission, THEN take the handswitch to START a second time.Take the handswitch to ST ART a second time, THEN write a Condition Report to document the action.

Write a condition report doc umenting the event and contact Maintenance.

The System Operator will rackout and perform a visu al inspection of the circuit breaker and write a condition report.

A.B.C.D.

63. Per Tech Specs Bases 3.3.1, Reactor Trip System (RTS) Instrumentation, what is thebasis of the Pressurizer Water Level - High Reactor trip?

Protects the pressurizer safety valves against water relief.

Provides the primary protection fo r preventing RCS over pressurization.

Protects against loss of pr essure control due to spra y nozzle being submerged.

Provides protection against exceeding containment design pressure in the event of a LOCA.A.B.C.D.

64. Unit 1 is in Mode 6 for a refueling outage.

• Two Plant Operators ar e required to enter a room that is posted as a Locked High Radiation Area (LHRA) to perform work.

Which one of the following co mpletes the statements below? The radiation level at which this posting is required is (1) . The LHRA key is obtained from (2) . 1) > 100 mrem/hr

2) Health Physics Supervision

1) > 100 mrem/hr

2) the Shift Suppor t Supervisor (SSS)

1) > 1000 mrem/hr

2) Health Physics Supervision

1) > 1000 mrem/hr

2) the Shift Suppor t Supervisor (SSS)

A.B.C.D.

65. Which one of the following completes the statements below for entry into the Dry Cask Storage Radiation Controlled Area (RCA)?Per AP-42, Access Cont rol, the operator (1) required to log in on the normalAuxiliary Building Access Control System (ACS) terminal prior to entering the Dry Cask Storage Area RCA.

Upon exiting from the Dry Ca sk RCA, the operator is requi red to perform a 2 minute frisk and also use the (2) .1) IS 2) Primary Access Point (P AP) exit portal monitors1) IS 2) Auxiliary Building RCA exit portal monitors 1) is NOT

2) Primary Access Point (P AP) exit portal monitors1) is NOT

2) Auxiliary Building RCA exit portal monitors A.B.C.D.

66. Which one of the following completes the statement below?

An employee who is a fully documented r adiation worker and DO ES NOT declare her pregnancy has an annual FNP Admi nistrative TEDE limit of

.450 mRem 500 mRem 2000 mRem 5000 mRem A.B.C.D.

67. Unit 1 is performing the actions of EEP-3.0, Steam Generator Tube Rupture, due to a tube rupture in the 1B SG.

• The 1B SG Narrow range level is 36% and rising.

Which one of the following co mpletes the statements below? The 1B SG narrow range level (1) adequate to begin the initial RCS cooldown. The operational implication of having sufficient level in the 1B SG prior to the cooldown is to (2) .1) is NOT

2) ensure a secon dary side heat sink1) is NOT 2) prevent SG depressurizati on during the RCS cooldown

1) IS 2) ensure a secon dary side heat sink

1) IS

2) prevent SG depressurizati on during the RCS cooldown A.B.C.D.

68. FRP-Z.1, Response to High Containment Pressure, has the following caution:

IF ECP-1.1, Loss of Emer gency Coolant Recirculation, is in effect, THEN Containment Spray should be operated as directed in ECP-1.1.

Which one of the following descr ibes the bases for giving priority to ECP-1.1?

ECP-1.1 directs t he operation of the Containm ent Spray (CS) pumps to ensure

.RWST level is conserved adequate NPSH for the RHR pumps is available the maximum available Containment heat removal systems are running automatic swapover of the CS pumps to the Containment sump is prevented A.B.C.D.

69. Unit 1 is operating at 100% power and the following conditions exist: * #1 Waste Monitor Tank (WMT) release is in progress.

• The Unit 1 Rad Side SO is at the RCA exit preparing to enter the portal monitors.

Subsequently, the plant emergency alar ms sounds and an announcement is made declaring a Site Area Emergency.

Which one of the following co mpletes the statements below? The Rad Side SO will go to t he designated assembly area (1) . The designated assembly area for the Rad Side SO is the (2) .1) after securing the #1 WMT release

2) Operations Support Center (OSC)

1) after securing the #1 WMT release

2) Control Room

1) immediately

2) Operations Support Center (OSC)

1) immediately

2) Control Room A.B.C.D.

70. The crew has transitioned to ECP-1.2, LOCA Outside Containment.* Step 2 of ECP-1.2 is in progress and the first flow path has been isolated.

The following conditions exist:* Aux Building radiation levels are rising slowly.* Safety Injection flow is stable.* Aux Building sump levels are rising slowly.

• PI-402 and 403, RCS 1C/1A L OOP RCS NR PRESS, are rising.

Which one of the following completes the statements below per ECP-1.2?

The first flow path t hat was isolated was (1) injection.The intersystem LOCA (2) been isolated.

(1)

(2)

RCP seal HAS RCP seal has NOT RHR cold leg HAS RHR cold leg has NOT A.B.C.D.

71. Unit 1 was operating at 100% power when a Reactor Trip and SI occurred due to a steam line break in containment.

The following conditions exist:

• The operating crew is performing the actions of EEP-2.0, Faulted Steam Generator Isolation.

• The maximum total AFW flow rate that can be achieved is 350 GPM.

• Containment pressure is 6 psig and falling.

• SG Narrow range levels are:

- 1A - Off Scale Low

- 1B - 32% and decreasing slowly

- 1C - 34% and decreasing slowlyWhich one of the following completes the statement below?

Secondary heat sink (1) adequate because (2) .1) IS

2) 1B and 1C SG leve ls are sufficient1) is NOT

2) Neither SG leve ls nor AFW flow c apability is sufficient

1) IS 2) AFW flow capability is sufficient1) is NOT

2) 1B and 1C SG levels are sufficient but AFW flow capability is NOT sufficient A.B.C.D.

72. Unit 2 has experienced a large steam break inside containment and the followingconditions exist:

• 1A SG Wide Range level indicates 0%.* AFW flow to 1B and 1C SG is currently 500 gpm.* FRP-P.1, Response To Imminent Pr essurized Thermal Shock, has been entered on a RED Path.* RCS cold leg temperature c ontinues to decrease slowly.* RCS pressure is 1500 psig and stable.

Which one of the following completes the statements below concerning the mitigation strategy of FRP-P.1?The RCS cooldown must (1) .An RCS pressure reduction (2) required.

(1)

(2) continue is NOT continue IS be stopped is NOT be stopped IS A.B.C.D.

73. A Dual Unit LOSP with a LOCA on Unit 1 has occurred and the following conditions exist:

• EEP-1.0, Loss of Reactor or Secondary Coolant, is in progress.

At 1000:

• WA2, 1-2A DG GEN FAULT TRIP, comes into alarm.

At 1015:

• The following alarms are received: - CF3, 1A OR 1B RHR PUMP OVERLOAD TRIP - CH2, RWST LVL A TRN LO

- CH3, RWST LVL B TRN LO Which one of the following states:

1) the status of Unit 1 emergency recirculation capability and 2) the action(s) t hat the applicable pr ocedure(s) direct?

1) One train ONLY of emergency recirculation capability has been lost.

2) Transfer to Cold Leg AND Containment Spray recirc ulation at this time.

1) One train ONLY of emergency recirculation capability has been lost.

2) Transfer to Cold Leg recirculation ONLY.

1) Both trains of emergency recirculation capability have been lost.

2) Minimize HHSI flow to the minimu m required to remove decay heat while attempting to restore at least one train of emergency recirculation. 1) Both trains of emergency recirculation capability have been lost.

2) Secure HHSI pumps while attempting to rest ore at least one train of emergency recirculation.

A.B.C.D.

74. The crew is responding to a Steam Line Break on Unit 1.

• Due to equipment failures , ECP-2.1, Uncont rolled Depressurization of All Steam Generator s, has been entered.

• All SG Narrow Range levels are 25% and lowering.Which one of the following completes the statement below?Per ECP-2.1, AFW flow will be adjusted to (1) .0 gpm to each SG to prevent excessive cooldown at least 20 gpm to each SG to prevent dryout of the SGs at least 20 gpm total AFW flow to minimize thermal stress to the SGs at least 395 gpm total AFW flow to main tain adequate heat sink A.B.C.D.

75. Which one of the following is the first Major Action Category in FRP-Z.2, Response To Containment Flooding, and reason for th is in accordance with the background document?Identify unexpected sources of water in t he sump since flooding could damage critical plant equipment.Evaluate the ECCS system status to determine a strategy to transition to simultaneous cold and ho t leg recirculation.

Have chemistry evaluate sump level, chem istry, and activity level to determine astrategy to transfer exce ss water out of containment.

Notify the TSC of sump chemistry, and acti vity level to determ ine potential changes in the planned transition to simultaneous cold and hot leg recirculation

.A.B.C.D.

REFERENCES UNIT 110/18/12 9:25:43FNP-1-AOP-2.0STEAM GENERATOR TUBE LEAKAGEVersion 35.0StepAction/Expected ResponseResponse Not Obtained 38ProcedureStepsMainPage 35 of 39 34.2[CA] WHENone of the following conditions occur,THEN stop the RCS pressure reduction.

34.2 33.2[ ]RCS pressure is less than affected SG pressure, ANDpressurizer level greater than 15%.OR[ ]Pressurizer level greater than 63%.

OR[ ]SUBCOOLED MARGIN MONITOR indication less than 16°F subcooled in CETC mode.34.3Verify both normal pressurizer spray valves -CLOSED.34.3Perform the following.

33.3 1A(1B) LOOP34.3.1Deenergize rod control system.

33.3.1[ ]PK 444C[ ]PK 444D [ ]Open both RX TRIP BKRs.

OR[ ]Open 1A and 1B MG SET SUPP BKRs.34.3.2Stop associated 1A & 1B RCPs.

33.3.234.3.3IF any rod bottom light not lit, THEN 33.3.3emergency borate RCS using FNP-1-AOP-27.0, EMERGENCY BORATION. 34.4Verify auxiliary spray valve - CLOSED.34.4Isolate auxiliary spray line.

33.4RCS PRZRAUX SPRAYCHG PMPS TO REGENERATIVE HX[ ]Q1E21HV8145[ ]Q1E21MOV8107 closed[ ]Q1E21MOV8108 closed34.5Verify both PRZR PORVs - CLOSED.34.5Close PRZR PORV ISO for any open PRZR PORV.

33.5Step 38 continued on next page ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:Date:Facility/Unit:

Region:I II III IV Reactor Type: W CE BW GE Start Time:Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade

of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items

if given in conjunction with the RO exam; SRO-only exams given alone require a final grade

of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

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

______________________________________

Applicant's Signature ResultsRO/SRO-Only/Total Examination Values

/ / PointsApplicant's Scores

/ / PointsApplicant's Grade

/ / Percent

1. 001AA2.01 001Unit 1 is operating at 85% power with the following conditions:

• STP-33.0B, Solid State Protection System Train B Operability Test, is in progress.

• The 'B' Reactor Trip Bypass Breaker has been racked in and closed.

• Control Bank D is at 203 steps.

• The Rod Control Bank Selector Switch is in AUTO.

• PS/446Z, FIRST STG IMPULSE PRESS SEL SWITCH, is in theChannel IV / PT447 position. Subsequently, the following occurs:

• PT-447, TURB FIRST STG PRESS, fails HIGH.

Which one of the following completes the statements below?The control rods will (1) .If the reactor is manually tripped at this time, the

'B' Reactor Trip BypassBreaker (2) light will be LIT.

(1)

(2) insert RED insert GREEN withdraw GREEN withdraw RED A.B.C.D.

PT-447 failing low will cause rods to insert continuously and failing high will cause rodsto withdraw.Per FSD-A181007, Figure 2 Sheet 2, a manual reactor trip will open the 'B' ReactorTrip Bypass Breaker.Distracter AnalysisA. Incorrect. First part is incorrect (See C.1). Plausible if the applicant reversesthe system response so that the rod control system would insertrods to suppress the indicated power rise due to the failure insteadof trying to adjust actual reactor power and Tavg to match thefailed indication. Second part is incorrect (See C.2). Plausible if the applicantbelieves that during the performance of STP-33.0B, the 'B' ReactorTrip Bypass Breaker will not open since that train is being tested.B. Incorrect. First part is incorrect (See A.1). Second part is correct (See C.2).

C. Correct. First part is correct. Impulse pressure, PT-477 is used to calculateTref. Since PT-447 failed high, Tref fails to the 100% power Trefand the rod control system will step rods out in order to raise current Tavg to match Tref. Impulse pressure is also used todetermine turbine power as compared to reactor power in the rodcontrol circuitry. When PT-447 fails high, the rapid rate of change of impulse power as compared to reactor power will also causerods to step out while impulse pressure is changing. This scenario has been run on desktop simulator and the rods willstep out (IC 058). Second part is correct. FSD A18007, Figure 2 sheet 2, shows thata manual trip actuation will open the 'B' Reactor Trip BypassBreaker. D. Incorrect. First part is correct (See C.1).

Second part is incorrect (See A.2).

K/A: 001AA2.01 Continuous Rod Withdrawal - Ability to determine andinterpret the following as they apply to the Continuous RodWithdrawal : Reactor tripped breaker indicatorImportance Rating: 4.2 4.2 Technical

Reference:

FSD-A181007, Reactor Protection System, Ver 18 FNP-1-EEP-0.0, Reactor Trip or Safety Injection, Ver 44 FNP-1-AOP-100, Instrumentation Malfunction, Ver 12References provided: None Learning Objective:

EVALUATE plant conditions and DETERMINE if entry intoAOP-100, Instrument Malfunction is required.(OPS-52521Q02) ANALYZE plant conditions and DETERMINE the successfulcompletion of any step in (1) EEP-0, Reactor Trip or Safety Injection [...]. (OPS-52530A07)Question History: WATTS BAR MAY 09 K/A match: This question requires the applicant to determine that acontinuous rod withdrawal is occurring due to the failureof PT-447. After the reactor is tripped, the applicant is thenrequired to interpret the reactor trip bypass breakerindication as to whether or not it is open.SRO justification: N/A

2. 003A3.05 002Unit 1 is in Mode 3 and preparing to start the 1C RCP.Which one of the following completes the statements below per SOP-1.1, ReactorCoolant System?The 1C RCP oil lift pump handswitch white light indicates (1) .The 1C RCP breaker closing operation (2) interlocked with a 2 minutetime delay. 1) BOTH the oil lift pressure has reached 600 psig AND 2 minutes have elapsed2) is NOT1) ONLY that the oil lift pressure has reached 600 psig2) IS1) ONLY that the oil lift pressure has reached 600 psig2) is NOT1) BOTH the oil lift pressure has reached 600 psig AND 2 minutes have elapsed2) IS A.B.C.D.

SOP-1.1 3.6 DO NOT attempt to start a RCP unless its oil lift pump has been delivering oilto the upper thrust shoes for at least two minutes. Observe the oil lift pumpsindicating lights to verify correct oil pump motor operation and oil pressure. The oil liftpumps should run at least 1 minute after the RCP's are started.

An interlock willprevent starting a RCP until 600 psig oil pressure is established. Note prior to step 4.3.11 - The oil lift pump must be operated for at least 2 minutesprior to starting the RCP.

4.3.18 Verify that the Oil Lift Pump for RCP 1C has run for at least two minutes,and is producing adequate pressure (white light ON).Distracter AnalysisA. Incorrect. First part is incorrect (See C.1). Plausible since the procedurerequires both conditions to be met before starting the RCP. Theapplicant could have the misconception that the light is associatedwith both the time and oil pressure. Second part is correct (See C.2). B. Incorrect. First part is correct (See C.1) Second part is incorrect (See C.2). Plausible if the applicant thinksthat the procedural requirement of allowing the oil lift pump to runfor at least two minutes is to satisfy a closing time delay interlock.C. Correct. First part is correct. RCP oil pressure must be a minimum of 600psig for the white light to come on. This may occur in as little as 1minute or as long as 6 minutes, but the white light is not timedependent, it is only pressure dependent. Second part is correct. The RCP breaker interlock is with the oilpressure switch not the 2 minute procedural requirement.

D. Incorrect. First part is incorrect (See A.1). Second part is incorrect (B.2).

K/A: 003A3.05 Reactor Coolant Pump System (RCPS)

- Ability to monitorautomatic operation of the RCPS, including: RCP lube oiland bearing lift pumpsImportance Rating: 2.7* 2.6 Technical

Reference:

FNP-1-SOP-1.1, Reactor Coolant System, Ver 47.2References provided: NoneLearning Objective: RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the ReactorCoolant Pumps, to include the following (OPS-40301D02):

• Oil lift systemQuestion History: MOD VOGTLE 12 K/A match: The only AUTO features of the RCP LO and Brg lift pump isthe white light will come on when the pressure reaches 600psig and then the permissive will clear at 600 psig allowingthe RCP breaker to be closed. Applicant must be able tomonitor the white indicating light for the RCP oil lift pump which indicates the discharge pressure of that pumpis > 600 psig and have knowledge that when the light comeson, the permissive automatically allows the RCP circuitbreaker being capable of closing due to oil pressure and is time. SRO justification: N/A

3. 004K3.08 003Unit 1 was operating at 100% power when the following occurred:

• The air supply to FCV-122, CHG FLOW REG, actuator has been severed and the valve has repositioned to its failed position.Which ONE of the following completes the statement below?FCV-122 is (1) and RCP seal injection flow will (2) . (1)

(2) CLOSED LOWER CLOSED RISE OPEN LOWER OPEN RISE A.B.C.D.

AOP-6.0 Table 1:Component No. Name Failed PositionQ1E21V347 (1-CVC-FCV-122) CHG FLOW REG OPENARP-1.4, DC4 - SEAL WTR INJ FLTR HI PProbable Cause.2. High seal injection flow rate.

3. Chg Flow Q1E21FCV122. (Also labeled Q1E21V347) failed closed.Distracter AnalysisA. Incorrect. First part is incorrect (See C.1). Plausible if the applicant doesrecall the fail position of FCV-122. Second part is correct (See C.2). Logical connection to the firstpart if the applicant believes that FCV-122 is upstream of the sealinjection line and the closure of FCV-122 would stop seal injectionflow.B. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See C.2). Logical connection to the firstpart if the applicant thought that FCV-122 failed closed since itwould be the correct seal injection response for this condition.C. Correct First part is correct. FCV-122 fails open. Second part is correct. The closure of FCV-122 will cause highseal injection flow and thus high filter DP due to all of the chargingpump discharge flow being directed to the seal injection filter (SeeARP-1.4, DC4 above). Conversely if the air line fails on FCV-122,the valve will fail open and a majority of the charging pump discharge flow will be directed to the normal charging path causing

the seal injection flow to go down. (See P&ID D-175039, SH 6,Chemical and Vol Control System for system flow.)D. Incorrect. First part is correct (See C.1). Second part is incorrect (See C.2). Logical connection to the firstpart if the applicant assumes that more charging flow equates tomore seal injection flow.

K/A: 004K3.08 Chemical and Volume Control System (CVCS) -

Knowledgeof the effect that a loss or malfunction of the CVCS will haveon the following: RCP seal injectionImportance Rating: 3.6 3.8 Technical

Reference:

P&ID D-175039, SH 6, Chemical and Vol Control System Ver 10 FNP-1-AOP-6.0, Loss of Instrument Air, Ver 40 FNP-1-ARP-1.4, DC4, Ver 53References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the Chemicaland Volume Control System, to include the componentsfound on Figure 3, Chemical and Volume Control Systemand Figure 4, RCP-Seal Injection System (OPS-40301F02).Question History: SURRY 09 1ST AUDIT K/A match: The CVCS malfunction is that the air line to FCV-122 hasbeen severed. Applicant must know how this CVCSmalfunction

affects seal injection flow

.SRO justification: N/A

4. 005A2.02 004The following conditions exist on Unit 1:

• The operating crew is cooling down per UOP-2.2, Shutdown of Unit From Hot Standby to Cold Shutdown.

• MODE 5 has just been entered and the following conditions exist: - RCS temperature is 195°F.

- BOTH 1A AND 1B RHR pumps are running in the cooldown mode.

Subsequently, PT-402, 1C LOOP RCS PRESS, fails HIGH.Which one of the following describes the minimum required action(s), if any, to beperformed per AOP-12.0, Residual Heat Removal Malfunction?No actions are required.Secure the 1A RHR pump ONLY.Secure the 1B RHR pump ONLY.Secure BOTH 1A AND 1B RHR pumps.

A.B.C.D.

FSD-181002:3.4.6.2 - All valves will shut automatically if RCS pressure increases to 700 psig. [...]The pressure inputs for this interlock are from PT402 for 8701A and B and from PT403for 8702A and B to prevent isolation of both trains of RHR due to a single pressuretransmitter failing high.AOP-12:Entry Conditions1.4 Closure of loop suction valve Step 1. Check RHR loop suction valves 1. Stop any RHR PUMP with closed loop - OPEN suction valve(s)See Tech Spec 3.4.12Technical Specification 3.4.12, LTOP (Low Temperature Over Pressure ProtectionSystem requires two RHR suction relief valve with setpoints

< 450 psig when thetemperature of one or more RCS cold legs is

< 325°F. If one or more of the RHR Loopsuction valves closed, then this Technical Specification would not be met and LowTemperature Over Pressure Protection would not be satisfied.Distracter analysisA. Incorrect. See B. Plausible if applicant remembers that the RHR Loopsuctions are opened and de-energized at some point but cannotrecall that it is when RCS temperature is less than 180°F. If thevalves were de-energized they would not shut so no action wouldbe required. Also, the applicant could remember that PT-402 and403 provide interlocks to OPEN the RHR loop suctions but NOTremember they also will close the valves on high pressure. Thiswould make this a correct answer. B. Correct. PT-402 failing high will close MOV-8701A and MOV-8701B whichisolates the suction to the 1A RHR pump. AOP-12 requires the 1ARHR pump to be secured. C. Incorrect. See B. Plausible if the applicant improperly believes that PT-402affects the 1B RHR pump suction valves instead of 1A RHR pump suction valves. If PT-403 failed high, this would be the correctanswer.D. Incorrect. See B. Plausible since the RHR loop suction valves have interlocks to prevent opening them if certain parameters are notmet (See reference material FSD A181002). PT-402 must be less than 402.5 psig in order to open MOV-8701A and MOV-8702Awhich are on OPPOSITE trains. If the applicant thought that theclosing on high pressure works the same way as the openinginterlocks this would be a correct answer since they would believea suction valve in each train will close and both RHR pumps wouldbe required to be secured.

K/A: 005A2.02 Residual Heat Removal System (RHRS) - Ability to (a)predict the impacts of the following malfunctions oroperations on the RHRS, and (b) based on thosepredictions, use procedures to correct, control, or mitigatethe consequences of those malfunctions or operations:Pressure transient protection during cold shutdownImportance Rating: 3.5 3.7Technical

Reference:

FSD-A181002, Residual Heat Removal, Ver 44 FNP-1-AOP-12, Residual Heat Removal Malfunction, Ver 25 FNP Technical Specifications, Ver 190References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if entry intoAOP-12.0, RHR System Malfunction and/or STP-18.4, Containment Closure is required. (OPS-52520L02)Question History: MOD ANO 05 K/A match: The applicant has to predict how the failure of the looppressure transmitter affects the RHR system in that itremoves one of the required Low Temperature OverPressure Protection System reliefs from service. Theapplicant will apply that prediction to AOP-12 to get to theaction required to miti gate this malfunction.SRO justification: N/A

5. 006K5.06 005Unit 1 was operating at 100% power when the following conditions occurred:

• A LOCA is in progress.

• The operating crew is performing the actions of EEP-0.0, Reactor Trip or Safety Injection, and is at the step to "Check RCS intact".

• RCS pressure is 475 psig and lowering.Which one of the following describes the current status of the ECCSsystem?

SI Accumulator Level

RHR Injection Flow Stable and on-scale Zero Dropping or off-scale low Zero Dropping or off-scale low Rising Stable and on-scale Rising A.B.C.D.Desktop simulator IC-73 (100% power) - 10,000 gpm LOCA produced 477 psig in theRCS in ~17 min. It takes an estimated 15+ minutes to get to the "Check RCS intact"step of EEP-1 which makes the value of RCS pressure valid.EEP-1 9. Check if LSHI Pump s should be stopped.

9.1 Check

RCS pressure - GREATER THAN 275 psig {435 psig}EEB-1ERP Step Text:

Check if LHSI Pumps should be stopped.Purpose: To stop the low-head SI pumps if RCS pressure is above their shutoff headto prevent damage to the pumpsBasis: Upon safety injection initiation all safeguard pumps are started regardless of thepossibility of high RCS pressure with respect to the low-head safety injection pumpshutoff head. On low-head systems where the pump recirculates on a small volumecircuit there is concern for pump and motor overheating. Shutdown of the pump andplacement in the standby mode, when the RCS pressure meets the criteria outlined inthis step, allows for future pump operability. If SI has not been previously reset and thelow-head SI pumps should be stopped, SI should be reset prior to stopping the pumps.SI can be reset regardless of containment pressure.FSD - A181009 3.3.1.1 3 Safety injection accumulators shall function as passive safeguardscomponents to rapidly inject [...] whenever the RCS pressure decreases below the tankcover gas pressure of 601 - 649 psig due to a loss of coolant accident Distracter AnalysisA. Incorrect. First part is incorrect (See B.1). Plausible if the applicant isunfamiliar with the injection pressures of the accumulators. The applicant may believe that RCS pressure is high enough to preventaccumulator injection. Second part is correct (See B.2). Logical connection to the first partif the applicant recognizes that RHR injects at a lower pressurethan the accumulators.B. Correct. First part is correct. FSD-A181009, 3.3.13. Safety injectionaccumulators shall function as passive safeguards components torapidly inject [...] whenever the RCS pressure decreases below thetank cover gas pressure of 601- 649 psig due to a loss of coolantaccident. Second part is correct. At 500 psig in the RCS, the accumulatorswill have injected but the RCS pressure will be above the RHR shutoff head pressure. EEP-1 uses 435 psig (Adverse, becausecontainment pressure is > 4 psig) as the criteria for RHR pumpshut off head. (See EEP-1 and EEB-1 above). C. Incorrect. First part is correct (See B.1). Second part is incorrect (See B.2) Logical connection to the firstpart if the applicant is unfamiliar with the injection pressures of theRHR pumps. If the applicant knows that the accumulatorsare(have) injecting(ed), they may also assume that RCS pressureis low enough to allow RHR injection.D. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See B.2). Logical connection to the firstpart if the applicant is unfamiliar with the injection pressures of theRHR pumps. The applicant may believe that the RHR pumps injectbefore the accumulators.

The K/A: 006K5.06 Emergency Core Cooling System (ECCS) - Knowledge ofthe operational implications of the following concepts asthey apply to ECCS: Relationship between ECCS flow andRCS pressureImportance Rating: 3.5 3.9 Technical

Reference:

FSD-A181009, CVCS/HHSI/Accumulators/RMWS, Ver 39 FNP-1-EEP-1.0, Loss of Reactor or Secondary Coolant, Ver 31 FNP-0-EEB-1.0, Specific Background Document for FNP-1/2-EEP-1.0, Ver 4References provided: NoneLearning Objective:

RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the Emergency Core Cooling System, to include thecomponents found on Figure 2, Accumulators, Figure 3,Refueling Water Storage Tank, and Figure 4, EmergencyCore Cooling System (OPS-40302C02).Question History: SUMMER 11 K/A match: The applicant is required to know which ECCScomponents are injecting into the core (flow)based on RCS pressure.SRO justification: N/A

6. 006K6.18 006Unit 1 has experienced a Reactor Trip and Safety Injection due to a faulted SG. The following conditions exist:

• The operating crew is performing EEP-2.0, Faulted Steam Generator Isolation.

• SCMM is in the CETC mode.

• RCS pressure is 1900 psig and rising slowly.

• At the step for verifying SI termination criteria, the crew notes that PT-457, PRZR PRESS, has failed LOW.Which one of the following completes the statements below?Subcooling margin calculated by A Train ICCMS will (1) .Subcooling margin calculated by B Train ICCMS will (2) . (1)

(2) be affected NOT be affected NOT be affected be affected be affected be affected NOT be affected NOT be affected A.B.C.D.

SOP-68:3.2 The normal display mode for the SMM is the "CETC" mode. This displays themargin to saturation (°F) using the highest core exit thermocouple (excluding upperhead) and the lowest pressure. The "RTD" mode displays the margin tosaturation (°F) using the hottest reactor coolant system (RCS) RTD (Th or Tc) andthe lowest pressure.

The pressure inputs are from PT-402 and 403 and fromPT-455 for A-train and PT-457 for B-train.Distracter AnalysisA. Incorrect. First part is incorrect (See B.1). Plausible if the applicant thinks thatPT-457 inputs to 'A" train SMM. Second part is incorrect (See B.2). Logical connection to the firstpart if the applicant recognizes that PT-457 is train related but assumes it inputs to the wrong train.B. Correct. First part is correct. PT-457 inputs to 'B' Train so the 'A' Train isunaffected. Second part is correct. PT-457 inputs to 'B' Train and the SMMuses the lowest pressure therefore the subcooling value of the 'B'Train will be affected (lower).C. Incorrect. First part is incorrect (See B.1). Plausible if the applicant thinksthat PT-455 and 457 input to both trains of SMM which is incorrect. If they did input to both trains, this would be a correct answer. PT-402 and 403 input to both trains. This is a commonmisconception. Second part is correct (See B.2) A logical connection to the firstpart if the applicant thinks that PT-455 and 457 input to both trainsof SMM instead of PT-402 and 403 which would make this a correct answer.D. Incorrect. First part is correct (See B.1). Plausible if the applicant thinks thatPT-455 inputs to 'A' Train and PT-456 inputs to 'B' Train instead ofPT-457 which would make this a correct answer. Second part is incorrect (See D.1). Logical connection to the firstpart based on D.1 discussion.

K/A: 006K6.18 Emergency Core Cooling System (ECCS)

- Knowledge ofthe effect of a loss or malfunction on the following will haveon the ECCS: Subcooling margin indicatorsImportance Rating: 3.6 3.9 Technical

Reference:

FNP-1-SOP-68.0, Inadequate Core Cooling Monitoring System, Ver 8.1References provided: NoneLearning Objective:

RELATE AND IDENTIFY

the operational characteristicsincluding design features, capacities and protectiveinterlocks for the following components associated with theInadequate Core Cooling Monitor System (OPS-52202E02):

• Subcooled Margin Monitor (SMM)Question History: NEW K/A match: At step 8 of EEP 2, the Shift Supervisor is required toevaluate plant conditions to determine if ECCS flow can beterminated. Part of this determination is evaluatingsubcooling.

The applicant has to know the effect of theloss of PT-457 on the subc ooling margin monitors inorder to be able to provide the Shift Supervisor thecorrect subcooling value.SRO justification: N/A

7. 007EA2.06 007Unit 2 was operating at 2% power with a plant startup in progress per UOP-1.2, Startupof the Unit from Hot Standby to Minimum Load and the following conditions occurred:At 1000:

• DG-15-2, 2B S/U XFMR TO 2G 4160V Bus, trips open.

At 1005:

• DF-01-2, 2A S/U XFMR TO 2F 4160V Bus, trips open.Which one of the following completes the statements below at 1006 with no operatoractions taken?The Reactor Trip breakers will be (1) . DRPI rod bottom lights (2) be LIT. (1)

(2) OPEN WILL OPEN will NOT CLOSED WILL CLOSED will NOT A.B.C.D.

FSD-181007Figure 2 Sheet 2 shows all signals that open the reactor trip breakers and none arepresent in this scenario. Unit 2 Load list:2A CRDM MG Set powered from 600V LC 2D which is power from 4160V 2F.2B CRDM MG Set powered from 600V LC 2E which is power from 4160V 2G.Rod Position Indication System (DRPI) has two power sources:MCC 2D - Normal - is NOT powered from a DG.

MCC 2B - Alternate - IS powered from a DG and is the source which DRPI is NORMALLY aligned. Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible since the applicant maybelieve that the loss of power to both trains of safety related powercauses the reactor trip breakers to open. Second part is correct (See D.2). Logical connection to the firstpart since the rod bottom lights would be lit if the applicant thought the reactor trip breakers opened. B. Incorrect First part is incorrect (See A.1). Second part is incorrect (See C.2). Plausible if the applicantbelieves that Rod Position Indication System (DRPI) is aligned to its normal power supply which would make this a correct answer.DRPI comes off B Train power and will lose power for a time while the 2B DG starts and loads. Then the rod bottom lights will be LIT.C. Correct. First part is correct. When the loss of the 2F bus occurs, the 2ACRDM MG Set will de-energize causing the rods to fall into the core. No reactor trip setpoints are exceeded at 1006 so the Reactor trip Breakers will not open. Second part is correct. Rod Position Indication System (DRPI) isnormally aligned to its ALTERNATE power supply which is a vital bus. When the DG re-energizes the 2G bus, the rod bottom lights will be LIT.D. Incorrect. First part is correct (See C.1). Second part is incorrect (See B.2). Logical connection to the firstpart for two reasons. If the applicant thought there was no trip (rodsdid not fall into the core) this would be the correct conclusion. If theapplicant knew that the trip breakers would not open they could stillbelieve the plausibility of B.2.

K/A: 007EA2.06 Reactor Trip - Ability to determine or interpret the followingas they apply to a reactor trip:

Occurrence of a reactor trip Importance Rating: 4.3 4.5Technical

Reference:

FNP-2-SOP-41.0, Control Rod Drive and Position Indication System, Ver 35.1 A351199, Unit 2 Electrical Load List, Ver 61References provided: NoneLearning Objective:

RECALL AND DESCRIBE the operation and function of thefollowing reactor trip signals, permissives, co ntrol interlocks,and engineered safeguards actuation signals associatedwith the Reactor Protection System (RPS) and EngineeredSafeguards Features (ESF) to include setpoint, coincidence,rate functions (if any), reset features, and the potential consequences for improper conditions to include thoseitems in the following tables (OPS-52201I07):

• Table 1, Reactor Trip Signals Question History: MOD FNP EXAM BANK K/A match: The applicant is required to interpret plant conditions anddetermine if a reactor trip has occurred

.SRO justification: N/A

8. 007K5.02 008Unit 1 is in Mode 5 and forming a pressurizer steam space (drawing a bubble) perUOP-1.1, Startup of Unit from Cold Shutdown to Hot Standby.

The vacuum refillprocedure will NOT be performed.

The following conditions exist:

• RCS pressure is being maintained at 325-375 psig.

• 1B RCP is running.

• 'A' Train RHR is on service with low pressure letdown aligned.

• RCS is in solid plant pressure control.

• Pressurizer temperature is 178°F and slowly rising.

• All PRZR heaters have been energized.Which one of the following completes the statements below? Per UOP-1.1, the pressurizer is at saturation conditions when (1) increases.During this evolution, PRT level will (2) . (1)

(2) charging flow remain constantletdown flow remain constantletdown flow risecharging flow rise A.B.C.D.

UOP-1.1:

5.11 WHEN pressurizer temperature increases to the saturation temperature for 375psig (approximately 442°F) as indicated by increasing RCS pressure or letdownflow , THEN establish a steam space in the pressurizer as follows 5.11.5 WHEN VCT level increases to 81%, THEN verify VCT HI LVL DIVERT VLVQ1E21LCV115A in the fully diverted position.Distracter analysisA. Incorrect. First part is incorrect (See B.1). Plausible if the applicant thinksthat FCV-122 is in auto and will open to raise charging flow in response to the RCS pressure rise. FCV-122 operates in automaticbased on pressurizer level and median Tavg (See AOP-100 Section 1.2 Figure 1 in reference material) Second part is correct (See B.2).B. Correct. First part is correct. UOP-1.1: 5.11 WHEN pressurizer temperatureincreases to the saturation temperature for 375 psig (approximately442°F) as indicated by increasing RCS pressure or letdown flow

,THEN establish a steam space in the pressurizer as follows: Second part is correct.

LCV-115A diverts to the RHT.C. Incorrect. First part is correct (See B.1). Second part is incorrect (See above). Plausible if the applicantimproperly believes that letdown diverts to the PRT vice RHT.D. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See C.2).

K/A: 007K5.02 Pressurizer Relief Tank/Quench Tank System (PRTS) -Knowledge of the operational implications of the followingconcepts as the apply to PRTS:

Method of forming a steambubble in the PZRImportance Rating: 3.1 3.4 Technical

Reference:

FNP-1-UOP-1.1, Startup of Unit from Cold Shutdown to Hot Standby, Ver 94.3References provided: None Learning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associatedwith the safe operation of the Pressurizer Systemcomponents and equipment, to include the following(OPS-40301E07):

• Normal Control MethodsQuestion History: MOD FNP 12 K/A match: The applicant has to know the effect on the PRT level(operational implications) during the formation of a

steam bubble in the pressurizer

.SRO justification: N/A

9. 008AK1.01 009Unit 2 was operating at 100% power when a Reactor Trip occurs and the followingconditions exist:

• Q2B13PSV8010A, PZR SAFETY, has failed OPEN.* Pressurizer pressure is 1020 psig.

• PRT pressure rises to 55 psig.

• Core Exit Thermocouples read 560°F.Which one of the following completes the statements below?Temperature on TI-469, SAFETY VLVS, will indicate approximately (1) .Pressurizer level will be (2) .Reference provided (1)

(2) 546°F rising 546°F lowering 320°F lowering 320°F rising A.B.C.D.

WOG Executive Guideline - During situations where a steam vent path is establishedfrom the pressurizer vapor space and where RCS subcooling is not indicated,pressurizer level may not be a true indication of RCS inventory. This can result fromsteam generated in the reactor vessel, passing through the pressurizer surge line andpreventing the water inventory of the pressurizer from draining into the RCS loops. Thisholdup of water can result in a stable or even increasing indicated pressurizer levelwhile RCS water inventory is actually decreasing. Pressurizer level should be relied ononly with hot leg or core exit subcooling present. In SI termination steps in the ERGs,pressurizer level is only checked after adequate RCS subcooling is confirmed.Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible if applicant believes thatthe temperature of the steam in the Pressurizer is the sametemperature as the steam entering the PRT. 546°F is theapproximate saturation temperature for 1035 psia. This was theerror made at the TMI accident. Second part is correct (See D.2).B. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See D.2). This is initially true but in thescenario given, subcooling is lost in the core and a bubble is formed in the vessel upper head. This will result in the Pzr level rising instead of lowering as one would expect. Plausible sinceduring a LOCA event the normal response is that Pzr leveldecreases. C. Incorrect. First part is correct. (See D.1)

Second part is incorrect (See B.2). D. Correct. First part is correct. Using the steam tables and the following pressures: 1020 psig + 15 = 1035 psia (RCS) 55 psig + 15 = 70 psia (PRT) ~320°F Second part is correct. Since the break is at the top of thepressurizer, the pressurizer level will be rising. This scenario wasrun on the desktop simulator. Pressurizer level was rising at 1020psig.

K/A: 008AK1.01 Pressurizer (PZR) Vapor Space Accident (Relief ValveStuck Open) - Knowledge of the operational implications ofthe following concepts as they apply to a Pressurizer VaporSpace Accident:

Thermodynamics and flow characteristicsof open or leaking valvesImportance Rating: 3.2 3.7Technical

Reference:

Properties of saturated and superheated steam, 1967 Westinghouse Owners Group, ERG Executive Guideline.References provided: Steam tables Learning Objective: SELECT AND ASSESS the Pressurizer Systeminstrument/equipment response expected when performingPressurizer System evolutions, including the Normal Condition, the Failed Condition, Associated Alarms,Associated Trip Setpoints, to include the components foundon Figure 3, Pressurizer and Pressurizer Relief Tank (OPS-52101E07)Question History: MOD HARRIS 09K/A match: Applicant has to determine the safety valve tailpeicetemperature using steam tables (Thermodynamics andflow characteristics) and pressurizer level trend(operational implications) during a stuck open PORVcondition.SRO justification: N/A

10. 008K2.02 010Unit 2 is operating at 100% power when the following occurs:

• A simultaneous dual Unit LOSP occurs:Which one of the following completes the statement below?The (1) CCW pump is being powered by the (2) DG. (1)

(2) 2A 1-2A 2A 1C 2C 1-2A 2C 1C A.B.C.D.

2C CCW pump is power from the 2F 4160V bus which is powered from the 1C DG inthe above scenario.Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible since, with the exceptionof spent fuel pool cooling, every other train related pump with an 'A'designation is an 'A' train pump. The applicant may not recall thatthe CCW system is "backwards" Second part is incorrect (See C.2). Plausible because if there wasan LOSP on Unit 2 only, the 1-2A DG would be assigned to theUnit 2 'A' train busses. However, there is a DUAL UNIT LOSP andthe 1C DG gets assigned to the Unit 2 'A' Train busses. Applicantsoften get confused as to the assignment of 'A' train DGs duringvarious loss of power scenarios.

B. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2). C. Incorrect. First part is correct (See D.1).

Second part is incorrect (See A.2). D. Correct. First part is correct. FSD A-181000: 3.1.5.4 [...] Without offsitepower available and with or without the presence of SIAS signal,the on-service train CCW pump shall trip, then both train CCWpumps start by the diesel generator loading sequencers (ESS or LOSP). The 2C CCW pump is the 'A' train pump therefore it will bepowered from the 1C DG. Second part is correct. Under the conditions in the stem, the 1CDG will tie to Unit 2 and supply the 2F, 2K and 2H busses (See FSD A181005 in reference material).

K/A: 008K2.02 Component Cooling Water System (CCWS)

- Knowledge ofbus power supplies to the following: CCW pump, includingemergency backup Importance Rating: 3.0* 3.2*

Technical

Reference:

FSD-A181000, Component Cooling Water, Ver 24 FSD-A181005, Diesel Generators, Ver 44References provided: NoneLearning Objective:

NAME AND IDENTIFY the Bus power supplies, for thoseelectrical components associated with the CCW System, toinclude the following: (OPS-40204A04):Question History: NEW K/A match: Requires the applicant to know the normal bus powersupply to the CCW pumps in order to know the correctDG that is its emergency backup power supply.

SRO justification: N/A

11. 009EK2.03 011The following conditions exist on Unit 1:

• An RCS leak is in progress.

• RCS pressure is 1600 psig and stable.

• Containment pressure is 3.1 psig and slowly rising.

• The crew has just transitioned to ESP-1.2, Post LOCA Cooldown and Depressurization.Which one of the following completes the statement below? The minimum SG narrow range water level must be greater than (1) to (2) .1) 31% 2) ensure SG tubes are covered to promote reflux boiling1) 48%2) ensure SG tubes are covered to promote reflux boiling1) 31%2) ensure adequate SG inventory to provide a secondary heat sink1) 48%

2) ensure adequate SG inventory to provide a secondary heat sink A.B.C.D.

ESB-1.2ERP Step Text - Check intact SG levels Purpose: To ensure adequate feed flow or SG inventory for secondary heat sinkrequirementsESP-1.28. Check any intact SG narrow range level - GREATER THAN 31% {48%}Distracter analysisA. Incorrect. First part is correct (See C.1). Second part is incorrect (See C.2). Plausible if the applicant doesnot understand the mechanism of reflux boiling and improperly assumes this leak rate is sufficient to allow reflux boiling to be themethod of heat removal at some time during the event. Reflux boiling is mentioned in ECP-1.1, Loss of Emergency CoolingRecirculation as a method of cooling if the RCS is NOT full.B. Incorrect. First part is incorrect (See C.1). Plausible if the applicant appliesadverse containment numbers as this would be the correct level. Second part is incorrect (See A.2).C. Correct. First part is correct. With containment pressure < 4 psig, adversenumbers are not warranted. The required SGWL is >31% (See Step 8 of ESP-1.2 above). Second part is correct. This is the correct reason for maintainingSGWL above 31% (See ESB-1.2 above). Additionally, at this leakrate, the RCS will remain full as the HHSI pump flow exceeds break flow and reflux cooling will not occur. (Ran on desktop simulator - IC 073, 200 gpm leak rate, trip and SIat 2000 psig Pzr pressure -- SI flow rate ~230 gpm at 2200 psig)D. Incorrect. First part is incorrect (See B.1). Second part is correct (See C.2).

K/A: 009EK2.03 Small Break LOCA - Knowledge of the interrelationsbetween the small break LOCA and the following: S/GsImportance Rating: 3.0 3.3*Technical

Reference:

FNP-1-ESP-1.2, Post LOCA Cooldown and Depressurization., Ver 24. FNP-1-ESB-1.2, Specific Background Document for FNP-1/2-ESP-1.2, Ver 2.1References provided: None Learning Objective:

STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with ESP-1.2, Post LOCA Cooldown andDepressurization. (OPS-52531F03)Question History: VOGTLE 10 K/A match: The applicant must know how the SGs interrelate to theRCS during a Small Break LOCA in that they arerequired to provide a secondary heat sink for the RCS. SRO justification: N/A

12. 010A2.01 012Unit 1 was operating at 100% power when a Reactor Trip and LOSP occurred.The following conditions exist:

• The 1A PZR HTR GROUP BACKUP handswitch is in AUTO.

• RCS pressure is 2000 psig.Which one of the following correctly describes Pressurizer Heater operation perESP-0.1, Reactor Trip Response?The 1A PZR Heaters will have power available, NO actions are required to energizethem.The 1A PZR Heaters will have power available AND manual actions on the MCBare required to energize them.The 1A PZR Heaters will NOT have power available. Manual actions are required toalign power to them on the EPB but NO other actions are required to energizethem.The 1A PZR Heaters will NOT have power available. Manual actions are required toalign power to them on the EPB AND manual actions on the MCB are required toenergize them.

A.B.C.D.

ESP-0.1Attachment 3 1.10.4 WHEN pressurizer heatergroup 1A operation isdesired, THEN place HTR GRP 1A BLOCKING BYPASS SW toBYPASS.1.10.5 IF required,THEN manually energizepressurizer heater group 1A.Distracter analysisA. Incorrect. See B. Plausible if the applicant fails to recall that by procedure,the heater switch is taken to off, then the blocking bypass switch istaken to BYPASS and the heater control switch placed in ON toenergize the heaters.B. Correct. The BIF LOSP sequencer reenergizes the emergency section of600v LC A on an LOSP at step 6. By procedure, the heater switchis taken to off, then the blocking bypass switch is taken to BYPASSand the heater control switch placed in ON to energize the heaters.C. Incorrect. See B. Plausible if the if the applicant confuses the 1A with the 1BPZR heaters which require EPB alignment and fails to recall thatby procedure, the heater switch is taken to off, then the blockingbypass switch is taken to BYPASS and the heater control switch isplaced in ON to energize the heaters.D. Incorrect. See B. Plausible if the if the applicant confuses the 1A with the 1BPZR heaters which require EPB alignment but recognizes theheater switch is taken to off, then the blocking bypass switch istaken to BYPASS and the heater control switch placed in ON toenergize the heaters.

K/A: 010A2.01 Pressurizer Pressure Control System (PZR PCS)

- Ability to(a) predict the impacts of the following malfunctions oroperations on the PZR PCS; and (b) based on thosepredictions, use procedures to correct, control, or mitigatethe consequences of those malfunctions or operations:Heater failuresImportance Rating: 3.3 3.6Technical

Reference:

FNP-1-ESP-0.1, Reactor Trip Response, Ver 32 FNP-1-EEP-0.0, Reactor Trip or Safety Injection, Ver 44References provided: NoneLearning Objective: NAME AND IDENTIFY the Bus power supplies, for thoseelectrical components associated with the PressurizerPressure and Level Control System, to include those items in Table 4- Power Supplies (OPS-52201H04).Question History: FNP EXAM BANK K/A match:

The LOSP causes the pressurizer heaters to becomeunavailable for use (failed) until operator action is taken to mitigate their loss. The applicant must know how tore-energize the PRZR heaters when they are lost during an LOSP. SRO justification: N/A

13. 011EK2.02 013Unit 1 has experienced a Loss of Off-site Power and a Large Break LOCA. Thefollowing conditions exist:

• ESP-1.3, Transfer to Cold Leg Recirculation, has been completed.Subsequently, the Shift Supervisor directs the OATC to perform ESP-1.4, Transfer toSimultaneous Cold and Hot Leg Recirculation and the following occurs:

• Power is lost to the 1G 4160V Bus and will not be restored for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.Which one of the following completes the statement below?At the completion of ESP-1.4, the running LHSI pump will be aligned for (1) legrecirculation and the running HHSI pump will be aligned for (2) leg recirculation.

(1)

(2) HOT COLD COLD COLD HOT HOT COLD HOT A.B.C.D.

ESP-1.4 intends to align LHSI to HOT leg and leave HHSI aligned for Cold legrecirculation. However, during realignment, if any portion of the LHSI system cannot bereconfigured (Step 1), then the system is returned to its original lineup of Cold leg recirc and the available HHSI train is aligned for HOT leg recirculation. Step 4 has theoperator assess the re-alignment and if the final requirement is not met, they aredirected to Step 1 and contacting the Technical Support Center for guidanceWe have recently developed a JPM that causes the alignment to be in a cold/cold orhot/hot alignment since some failures can lead you there. That is precisely the reasonthe procedure will direct you to the TSC staff if a final alignment other than cold/hot orhot/cold is reached by the end of the procedure.The final alignment of LHSI and HHSI can be confusing when coupled with powerlosses and/or equipment failures. Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible since the applicant may believe the procedure allows only one train of LHSI to be aligned tothe Hot Leg. This would be correct if it were HHSI. Second part is incorrect (See B.2). Plausible if the applicantbelieves that if one train cannot be realigned then neither will be aligned. This would be correct if it were LHSI.B. Incorrect. First part is correct (See D.1) Second part is incorrect (See A.2). C. Incorrect. First part is incorrect (See A.1).

Second part is correct (See D.2). Plausible since when power islost or equipment malfunctions, there are allowances to come backto steps in the procedure and perform them when power isrestored and/or equipment repaired such as in ESP-1.3 when thecharging suction and discharge header MOVs are aligned. Anyalignment is possible in this procedure once malfunctions occur.D. Correct. First part is correct. Per ESP-1.4 Step 1, if both trains of LHSIcannot be aligned to Hot Leg recirc then both trains are left alignedto cold leg recirc. Second part is correct. ESP-1.4 Step 2 will align the A train HHSIto Hot Leg recirc and Step 3 will leave B train in its originalalignment.

K/A: 011EK2.02 Large Break LOCA -

Knowledge of the interrelationsbetween the Large Break LOCA and the following: Pumps Importance Rating: 2.6* 2.7*Technical

Reference:

FNP-1-ESP-1.4, Transfer To Simultaneous Cold and Hot Leg Recircul ation, Ver 16.References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing [...] (2)ESP-1.4, Transfer to Simultaneous Cold Leg and Hot Leg Recirculation. (OPS-52531G06)Question History: FNP EXAM BANKK/A match: The applicant is required to know the interrelation betweenthe RHR/Charging Pumps and the RCS during a LargeBreak LOCA. Based on the scenario given, the applicantmust determine the final pump alignment.SRO justification: N/A

14. 011K2.02 014Unit 1 is operating at 100% power when the 1E 4160V bus becomes de-energized dueto an electrical fault.Which one of the following completes the statement below?Pressurizer heater groups have lost their normal power supply.1C and 1D ONLY1A, 1C and 1D ONLY1B and 1E ONLY1B, 1D and 1E ONLY A.B.C.D.LoadPressurizer Heater Group A 600V LC A (Normal) 4160V D Pressurizer Heater Group B 600V LC C (Normal) 4160V E Pressurizer Heater Group C 600V LC M 4160V D Pressurizer Heater Group D 600V LC M 4160V DPressurizer Heater Group E 600V LC N (Unit 1 and 2) 4160V EDistracter analysisA. Incorrect. See C. Plausible since these heaters are powered from the sameLC. The applicant may believe that this is the only LC affected bythe power loss. Although 1A heaters are also powered from thesame 4160V Bus as these heaters, the applicant may believe the1A heaters are powered from the Emergency Bus (1F) since they are sequenced on after an LOSP.B. Incorrect. See C. Plausible since this would be the impact for the loss of 1D4160V bus and the applicant may think theses heaters aresupplied by 1E 4160V bus.C. Correct. Per Unit 1 Electrical Load List: 1E 4160V Bus supplies 1C Load Center (LC) and 1N LC. 1C LC - 1B pressurizer heaters. 1N LC - 1E pressurizer heatersD. Incorrect. See C. Plausible if the applicant knows that 2 sets of heaters arepowered from the same LC but cannot correctly recall which ones.The 1B heaters is a partially correct answer and would be includedif the applicant thinks these heaters are powered from the same4160V bus as the 1B heaters.

K/A: 011K2.02 Pressurizer Level Control System (PZR LCS) - Knowledgeof bus power supplies to the following: PZR heatersImportance Rating: 3.1 3.2Technical

Reference:

A506250, Unit 1 Electrical Load List, Ver 74.0References provided: NoneLearning Objective: NAME AND IDENTIFY the Bus power supplies, for thoseelectrical components associated with the PressurizerPressure and Level Control System, to include those items in Table 4- Power Supplies (OPS-52201H04).Question History: NEW K/A match: Applicant is required to know the power supplies to thepressurizer heaters in order to determine which ones havelost power. The power supply has to go back to the 4160V bus so the applicant also has to know the LC supplies aswell.SRO justification: N/A

15. 012A4.06 015The following conditions exist on Unit 1:

• A loss of 'A' Train Auxiliary Building 125V DC Bus has occurred.Which one of the following completes the statement below?Placing the MCB Reactor Trip handswitch in TRIP would

if they were closed.

open ALL reactor trip a nd bypass breakers ONLY open the 'B' reactor trip breaker and the 'B' reactor trip bypass breaker ONLY open the 'B' reactor trip breaker and the 'A' reactor trip bypass breaker open BOTH reactor trip breakers but NOT open either reactor trip bypass breaker A.B.C.D.

FSD-A181007:3.3.2 pg 3-10The first method of tripping the breaker (i.e., reactor trip or bypass breakers) is by aloss or drop of rated voltage to the Undervoltage Relay (UV

). The relay is normallyenergized from the 48 volt DC from the RPS.

When the voltage is removed by anautomatic reactor trip signal, the relay is de-energized and releases the UV trip lever,which actuates the trip shaft, causing the breaker to unlatch from the closed position. The second method of tripping the trip shaft is by the shunt trip lever when thenormally de-energized shunt trip (SHTR) coil is energized. When energized, the SHTRcoil is powered from the 125 volt DC system used to close the reactor trip andbypass breaker closing circuits.Distracter analysisA. Correct. Without 'A' train DC, the UV coils from the "A' Train ReactorProtection System (RPS) will still open 'A' Trip and 'B' Bypassbreakers. 'B' train RPS deenergizes the UV coils for 'B' Trip and 'A'Bypass breakers. 'B' Train Aux Building DC will open the 'B' Tripbreaker.B. Incorrect. See A. Plausible if the applicant did not recall that the UV coilsfrom RPS will trip ALL Trip and Bypass breakers. Since both listedbreakers are 'B' breakers, this adds to plausibility due to theapplicant thinking the 'B' train is unaffected and still would cause areactor trip if the system worked this way.C. Incorrect. See A. Plausible since this is how the RPS opens the Trip andBypass breakers. The applicant may recall that this is how the RPS works but not realize that the loss of 'A' Train DC has no effect onthe RPS. D. Incorrect. See A. Plausible if the applicant thinks that the Trip breakers aretripped by RPS and the Bypass breakers from Aux Building DC.Since the Shunt trip coils on the Bypass breakers can ONLY beoperated locally, the applicant may think that without DC theBypass breakers will not open.

K/A: 012A4.06 Reactor Protection System (RPS) - Ability to manuallyoperate and/or monitor in the control room: Reactor tripbreakersImportance Rating: 4.3 4.3 Technical

Reference:

FSD-A181007 Reactor Protection System, Ver 18 D-177198, Sheet 2, Ver 3References provided: NoneLearning Objective:

RECALL AND DESCRIBE the operation and function of thefollowing reactor trip signals, permissives, co ntrol interlocks,and engineered safeguards actuation signals associatedwith the Reactor Protection System (RPS) and EngineeredSafeguards Features (ESF) to include setpoint, coincidence,rate functions (if any), reset features, and the potential consequences for improper conditions to include thoseitems in the following tables (OPS-52201I07):

• Table 1, Reactor Trip Signals Question History: FNP 10K/A match: Requires the applicant to monitor the effect on theReactor Trip and Bypass Breaker Positions due to aloss of DC when they are manually tripped (operated).

SRO justification: N/A

16. 013G2.1.19 016The Integrated Plant Computer Display on the following page is provided forevaluation of this question.Unit 1 has experienced a reactor trip and the following conditions exist:

• The operating crew is verifying the immediate operator actions per EEP-0.0, Reactor Trip or Safety Injection.

• MLB-1, 1-1 and 11-1, SAFETY INJECTION, are NOT LIT.The STA reports the following indications on the Plant Computer:

• PT0455 PRESSURIZER PRESSURE CHAN 1 is 1841 psig.

• PT0456 PRESSURIZER PRESSURE CHAN 2 is 1855 psig.

• PT0457 PRESSURIZER PRESSURE CHAN 3 is 1845 psig.

• PT0444A PRESSURIZER PRESSURE CHAN 4 is 1857 psig.

• PT0445A PRESSURIZER PRESSURE CHAN 5 is 1855 psig.

• PT0464 STEAM HEADER PRESSURE is 6.4 psig.

Which one of the following completes the statements below?A Safety Injection (1) required.MSIV-3370A, B and C (2) OPEN.Reference Provided (1) (2) IS are NOT IS ARE is NOT are NOT is NOT ARE A.B.C.D.

EEP- 0.0 -

III. The following are symptoms that require safety injection, if one has not occurred:

SI Signal Instrumentation

Setpoint Coinc1. Pressurizer pressure low PT 455, 456, 457 1850psig 2/3FSD-A181007 - Pg 2-26The Main Steam Line Isolation is initiated by the following:b. Low steam pressure; 585 psig on 2/3 S.G.Distracter AnalysisA. Correct. First part is correct.

SI Signal Instrumentation

Setpoint Coinc 1. Pressurizer PT 455, 456, 457 1850psig 2/3 pressure low Second part is correct. Even though MSIV-3369A/B/C indicateOPEN, the downstream steam header pressure is 6.4 psig whilethe upstream pressure is >900 psig on all SGs, MSIV-3370A/B/Cmust be closed even though they are not modeled on the IPC.B. Incorrect. First part is correct (See A.1) Second part is incorrect (See A.2).

Plausible if the applicant fails toevaluate steam pressure and assumes that since MSIV-3369A/B/Care open then MSIV-3370A/B/C must also be open.C. Incorrect. First part is incorrect (See A.1). Plausible if the applicant has themisconception that the control channels, PT 444A and 445A areused to evaluate pressure instead of the protection channelsPT-455, 456 and 457. Second part is correct (See A.2).D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2).

K/A: 013G2.1.19 Engineered Safety Features Actuation System (ESFAS) -Ability to use plant computers to evaluate system orcomponent statusImportance Rating: 3.9 3.8 Technical

Reference:

FNP-1-EEP-0.0, Reactor Trip or Safety Injection, Ver 44 FSD - A181007, Reactor Protection System, Ver 18References provided: Screen Image of the Integrated Plant ComputerLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associatedwith the safe operation of the Emergency Core CoolingSystem components and equipment, to include the following(OPS-40302C07): [...]

• Automatic actuation including setpoint ( example SI, PhaseA, Phase B, MSLIAS, LOSP, SG level) [...]

• Actions needed to mitigate the consequence of theabnormality Question History: NEW K/A match: Applicant must evaluate a set of data from the plantcomputer and based on that determine if ESFAS systemactuation is necessary. SRO justification: N/A 013G2.1.19 P 016

17. 014K5.01 017Unit 1 is performing a reactor startup per UOP-1.2, Startup Of Unit From Hot StandbyTo Minimum Load, when the following conditions occurred:

• The OATC pulled Control Bank D to 100 Steps by Step Demand Counter.* Rod B8 was noted to be indicating 54 Steps by DRPI.Which one of the following completes the statements below?Rod B8's position is (1) .Per Tech Spec Bases 3.1.7, Rod Position Indication, (2) is(are) themost reliable indication.1) exactly 100 steps2) the group step counters 1) approximately 100 steps2) the group step counters 1) exactly 54 steps2) DRPI 1) approximately 54 steps

2) DRPI A.B.C.D.The Bank Demand Position Indication System counts the pulses from the Rod ControlSystem that move the rods. There is one step counter for each group of rods. Individualrods in a group all receive the same signal to move and should, therefore, all be at thesame position indicated by the group step counter for that group. The Bank DemandPosition Indication System is considered highly precise (+/- 1 step or +/- ? inch).

If a roddoes not move one step for each demand pulse, the step counter will still countthe pulse and incorrectly reflect the position of the rod. The DRPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based oninductive analog signals from a series of coils spaced along a hollow tube with a centerto center distance of 3.75 inches, which is 6 steps. To increase the reliability of thesystem, the inductive coils are connected alternately to data system A or B. Thus, ifone system fails, the DRPI will go on half accuracy with an effective coil spacing of 7.5inches, which is 12 steps. Therefore, the normal indication accuracy of the DRPI System is +/- 4 steps (all coils operable and 1 step added for manufacturing andtemperature tolerances), and the maximum uncertainty is +/- 10 steps (only one datasystem A or B coils operable). With an indicated deviation of 12 steps between thegroup step counter and DRPI, the maximum deviation between actual rod position andthe demand position could be 22 steps.

Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible if the applicant does notrecall that the Step Demand Counter only counts electricalimpulses and thinks it actually measures rod location. Second part incorrect (See D.2). Plausible if the applicant thinks that since the group step counters are the most precise indication,they are the most reliable. B. Incorrect. First part is incorrect (See D.1). Plausible if the applicant does notrecall that the Step Demand Counter only counts electricalimpulses. Each bank has 2 Step Demand Counters that step in 1step increments. Control Bank D Group 1 will move 1 step then Control Bank D Group 2 will move one step. If bank 1 moves 1 step and the rod control switch is released before bank 2 moves,bank 1 would be at 100 steps and bank 2 would be at 99 steps.This is commonly known as 991/2 steps. The applicant mayreason that this is a potential reason to call rods by step counter asapproximate. Second part is incorrect (See A.2).

C. Incorrect First part is incorrect (See D.1). Plausible since the DRPI lightschange only every 6 steps and 54 steps is a DRPI display lightlocation. The applicant may think that since a DRPI display light islit, the rod is exactly at that position. Second part is correct (See D.2).

D. Correct. First part is correct. Since DRPI measures actual rod positionbased on the location of the rod in reference to the measurementcoils and the step counter only counts electrical pulses, the rod is at ~54 steps. Also, the accuracy of DRPI is

+ 4 steps so the rodheight is approximate. Second part is correct. DRPI is the most reliable because itactually senses the location of the rod using coils.

K/A: 014K5.01 Rod Position Indication System (RPIS) - Knowledge of theoperational implications of the following concepts as theyapply to the RPIS: Reasons for differences between RPIS and step counterImportance Rating: 2.7 3.0 Technical

Reference:

FNP Technical Specifications Bases, Ver 58References provided: NoneLearning Objective:

RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the following components associated with theDigital Rod Position Indication System (OPS-52201F02):

• Rod Position Detectors Question History: FNP 01 K/A match: This question requires the applicant to determine thelocation of rod B8 (operational implication) based ontheir knowledge of the differences, based on design, ofhow rod heights are measured between rod control(step counters) and DRPI in that DRPI is the most

reliable indication.SRO justification: N/A

18. 015/17AA2.02 018Unit 1 is operating at 100% power.The following occurs:

• MOV-3052, CCW TO RCP CLRS, closes.

• DD3, CCW FLOW FROM RCP OIL CLRS LO, comes in to alarm.Which one of the following completes the statements below?The most limiting components for this event are the RCP (1) .The RCPs will be required to be stopped within approximately (2) . 1) Motor Bearings 2) 2 minutes1) Motor Bearings2) 60 minutes1) Pump Lower Radial Bearings2) 2 minutes1) Pump Lower Radial Bearings

2) 60 minutes A.B.C.D.

DD1 - CAUTION: RCP's with #1 Seal Leakoff less than 2.5 gpm may develop lowerbearing and seal temperatures that exceed 225°F within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a loss ofseal injection. DD3 - On a complete Loss of CCW Flow to RCP Motor Bearing Oil Coolers, thebearing temperature will exceed 195°F in approximately 2 minutes.

4. IF any RCP Motor Bearing Temperature exceeds 195°F, THEN: A. IF the Reactor is critical, THEN trip the reactor. B. Stop the RCP. C. Perform the actions required by FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION. D. Perform action of FNP-1-AOP-4.0, LOSS OF REACTOR COOLANT FLOW astime allows. MOV-3052 isolate CCW flow to the RCP oil coolers and the RCP thermal barrier heatexchanger which functions to cool the lower radial bearing on a loss of RCP sealinjection flow.Distracter analysisA. Correct. First part is correct. The RCP motor bearings are the most limitingcomponents for this scenario since the CCW flow is lost and RCP motor bearing temperatures will increase rapidly. Second part is correct. The RCP motor bearing temperatures willincrease to 195°F with 2 minutes. The ARP has the operator tripthe reactor and secure all RCPs for this failure. B. Incorrect. First part is correct (See A.1). Second part is incorrect (See A.2) Plausible since 60 minutes isthe time the lower radial bearing temperature will rise in 1-2 hours on a loss of RCP's with #1 Seal Leakoff less than 2.5 gpmfollowing a loss of seal injection. Plausible since this is a timerequirement for a RCP malfunction on the same Annunciator panelas DD3. C. Incorrect. First part is incorrect (See A.1) RCP lower radial bearings arecooled from two sources. A loss of the CCW will not cause the lower radial bearing temperatures to rise. Plausible since CCW isnormal cooling to components and this is one of a few componentswith 2 cooling sources. Second part is correct (See A.2).

D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2).

K/A: 015AA2.02 Reactor Coolant Pump (RCP) Malfunctions -

Ability todetermine and interpret Abnormalities in RCP air vent flow paths and/or oil cooling system as they apply to theReactor Coolant Pump Malfunctions (Loss of RC Flow): Importance Rating: 2.8 3.0Technical

Reference:

FNP-1-ARP-1.4, DD1 and DD3, Ver 53References provided: None Learning Objective:

LIST AND DESCRIBE the sequence of major actionsassociated with AOP-9.0, Loss of Component CoolingWater. (OPS-52520I04).

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing AOP-9.0,Loss of Component Cooling Water. (OPS-52520I06).Question History: NEW K/A match: The malfunction of the RCP is the closure of MOV-3052.The applicant must determine/interpret that a loss of CCWto the RCP oil coolers and lower radial bearings hasresulted, then must interpret how this malfunction affects theRCP components (oil coolers and seal) and the timerequired for action to be taken. SRO justification: N/A

19. 017K6.01 019Unit 1 has experienced a Reactor Trip and SI due to a LOCA and the followingconditions exist:

• The operators have transitioned to EEP-1.0, Loss of Reactor or Secondary Coolant.

• The Core Exit Thermocouples (CETCs) are reading as follows:

- TWO CETCs are indicating a SHORT circuit.

- THREE CETCs are 1204°F and rising.

- All other CETCs are reading between 950°F and 1150°F and rising.Which one of the following completes the statements below?The indication for the SHORT circuited CETCs fail (1) . The (2) CETC is used to evaluate entry into FRP-C.2, Response To DegradedCore Cooling.

(1)

(2) HIGH hottest HIGH 5 th hottest LOW hottest LOW 5 th hottest A.B.C.D.

CSF-0.25th hottest CETC <1200 °F?

NO Go to FRP-C.1YESRCS SUBCOOLING NO 5th hottest CETC <700 °F? NO Go Tofrom CETC >

FRP-C.216°F {45°F}?YESCSF - SATU263686 pg 3-5The signal conditioning panel contains the open thermocouple detection circuitry, noisefiltering capacitors, and the cold reference junction compensation circuitry. Cold junction compensation is accomplished by measuring the barrier temperature utilizing asemiconductor temperature sensor located on the signal conditioning panel. Thetemperature sensor circuit produces an output voltage, that is equivalent to the temperature of the barrier strip. This output voltage is read in through one of thechannels on the Analog Input Boards (DT1748 and DT1748-24EX boards'. Thethermocouple signals are then compensated in the softwa re by adding the value of thecold reference junction to the thermocouple signals.

If any of the thermocouples areopen or shorted the signa l conditioning panels open thermocouple detectioncircuitry will cause the input to be driven down to 0V.Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible if the applicant does notrecall if a thermocouple fails high or low when shorted. An RTDthat experiences an open circuit will cause a high temperature reading. The applicant could confuse RTD and thermocoupleoperating theory. Second part is incorrect (See D.2). Plausible if the applicantdoesn't recall that the 5th hottest is selected to allow for failed highthermocouples. This is a common misconception.B. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2). C. Incorrect. First part is correct (See D.1).

Second part is incorrect (See A.2).D. Correct. First part is correct. Thermocouples that are shorted fail low. Second part is correct. See CSF-0.2 above.

K/A: 017K6.01 In-Core Temperature Monitor System (ITM) - Knowledge ofthe effect of a loss or malfunction of the following ITMsystem components: Sensors and detectorsImportance Rating: 2.7 3.0 Technical

Reference:

FNP-1-CSF-0.2 Core Cooling, Ver 17 FNP-0-CSB-0.0, Specific Background Document For FNP-1/2-CSF-0, Critical Safety Function Status Trees, Ver 1U-263686, ICCMS Tech Manual Vol II, Ver 2 OPS-31701G, Sensors and detectors, Ver 4References provided: NoneLearning Objective:

RELATE AND IDENTIFY

the operational characteristicsincluding design features, capacities and protectiveinterlocks for the following components associated with theInadequate Core Cooling Monitor System (OPS-52202E02):

• CETC MonitorQuestion History: NEW K/A match: The applicant is required to have a knowledge of theeffect of a shorted CETC on the incore temperaturemonitoring system. SRO justification: N/A

20. 022A1.03 020Unit 1 is operating at 100% power with the following conditions:At 1000:

• Containment Pressure is 0 psig.

• Containment temperature is 100°F.

• MI-3319A, B, C and D, CTMT CLR SUPP AIR MOISTURE, indicate 72 Dewpoint °F.

• Containment Coolers are running in slow speed.At 1015:A LOCA occurs and the following conditions exist:

• Containment Pressure is 5 psig.

• Containment temperature is 150°F.

• MI-3319A, B, C and D, CTMT CLR SUPP AIR MOISTURE, indicate 130 Dewpoint °F.Which one of the following completes the statements below at 1015?The Containment Cooler discharge will be through the (1) .The Containment Cooler fans will be drawing (2) amps than at 1000.

(1)

(2) ductwork MORE dropout plate MORE ductwork LESS dropout plate LESS A.B.C.D.Added that the Containment Coolers are running in Slow Speed prior to theLOCA. The FSD states that the design of the Containment Cooler motor is basedon 80Hp in Fast and 115 Hp during a LOCA when the coolers are running in slow. However, there is no data to determine at what point the mass of Containmentatmosphere increases to the point where the Slow speed fan would draw morecurrent than Fast speed. With the Containment Cooler in Slow at the beginningof the event, the applicant will only have to evaluate the current drawn by thecooler based on atmospheric conditions (humidity) which meets the K/A -Containment Cooling System (CCS) - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating theCCS controls including: Containment humidity.

Tech Spec 3.6.6 Bases:In post accident operation following an actuation signal, unless an LOSP signal ispresent, the Containment Cooling System fans are designed to start automatically inslow speed if not already running. If an LOSP signal is present, only the two fansselected (one per train) will receive an auto-start signal and will start in slow speed. Ifrunning in high (normal) speed, the fans automatically shift to slow speed.

The fans are operated at the lower speed during accident conditions to prevent motoroverload from the higher mass atmosphere. In addition, if temperature at thecooler discharge reaches 135°F, fusible links holding dropout plates will open and the fan discharge will no longer be directed through the common dischargeheader. This function helps to protect the fans in a post-accident environment byreducing the back pressure on the fans.FSD- A181013

3.1.2.5 A 125 hp motor is provided for fan operation to meet the design brakehorsepower requirement of 105 hp during low-speed operation following a LOCA

.During normal operation, the design brake horsepower of the fan in high-speedoperation is 80 hp. Distracter analysisA. Incorrect. First part is incorrect (See B.1). When containment temperaturereaches ~135°F, the dropout plates fall open. Plausible if theapplicant believes that the links melt at 160°F to 175°F which iswhen the fire damper fusible links melt. Second part is correct (See B.2). B. Correct. First part is correct. The dropout plates open at ~ 135°F. Second part is correct. The higher mass atmosphere inContainment due to the LOCA will cause the fan motors to drawmore current.C. Incorrect. First part is incorrect (See A.1).

Second part is incorrect (See B.2). Plausible if the applicant onlythought that the increase in temperature caused air density tolower and did not consider other factors affecting Containment atmosphere.D. Incorrect. First part is correct (See B.1). Second part is incorrect (See C.2).Medium break LOCA run on desktop simulator produced the approximate values.

K/A: 022A1.03 Containment Cooling System (CCS) - Ability to predictand/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment humidity Importance Rating: 3.1 3.4 Technical

Reference:

FSD-A181013, Containment Ventilation System, Ver 14.References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associated with the safe operation of the Containment Spray andCooling System components and equipment, to include thefollowing (OPS-40302D07):

• Normal Control Methods

• Abnormal and Emergency Control Methods

• Automatic actuation including setpoint (example SI,Phase-B, LOSP) and the effect of selecting the containment cooler control to local.Question History: NEW K/A match: Requires the applicant to monitor changes in containmenthumidity and based on this predict the changes in theparameters of the containment cooler fans in an environment with more humidity (water vapor) and willdraw more current. The dropout plate will actuate at135°F to reduce the back pressure on the fans.SRO justification: N/A

21. 022AK1.04 021Unit 1 is operating at 100% power and the following conditions exist:

• AOP-16.0, CVCS Malfunction, has just been exited after a charging flow controller failure.

• FK-122, CHG FLOW, is in MANUAL and has been repaired. Subsequently, FK-122 is placed in AUTOMATIC and the following conditions exist:

• One 60 gpm orifice is on service.

• Charging flow is stable at 62 gpm.Which one of the following completes the statement below?If FK-122 were to go to minimum demand, charging flow woulddecrease to a minimum flow rate of (1) , which is designed to prevent (2) . 1) 18 gpm2) flashing downstream of the letdown orifices 1) 18 gpm2) overheating of the charging pumps1) 40 gpm2) flashing downstream of the letdown orifices 1) 40 gpm

2) overheating of the charging pumps A.B.C.D.

3.3 In auto, CHG FLOW FK 122 minimum demand corresponds to 18 gpm chargingflow. This ensures adequate cooling to the regenerative heat exchanger to preventflashing downstream of the letdown orifices with one 60 gpm orifice on service. Withtwo orifices on service, approximately 40 gpm charging flow is required forregenerative heat exchanger cooling.

Distracter analysisA. Correct. First part is correct. Minimum charging flow in AUTOMATIC is 18gpm. Second part is correct. Per P&L 3.3, 18 gpm ensures adequatecooling to the regenerative heat exchanger to prevent flashingdownstream of the letdown orifices with one 60 gpm orifice on service. B. Incorrect. First part is correct. (See A.1) Second part is incorrect (See A.2). Plausible if the applicant thinksthat reduced charging flow would equate to reduced mini-flow flow which is incorrect. Also, the charging miniflow goes through theseal water return HX and temperature would be unaffected bychanging charging flows. C. Incorrect. First part is incorrect (See A.1). Plausible since Figure 1 ofSOP-2.1, re-establishing LTDN after isolation with no equipment malfunction, has the operator establish 40 gpm flow rate in step 1 when placing one orifice on service. Second part is correct (See A.2).D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2).

K/A: 022AK1.04 Loss of Reactor Coolant Makeup -

Knowledge of theoperational implications of the following concepts as theyapply to Loss of Reactor Coolant Makeup: Reason forchanging from manual to automatic control of charging flowvalve controller.Importance Rating: 2.9 3.0Technical

Reference:

FNP-1-SOP-2.1, Chemical and Volume Control SystemPlant Startup and Operation, Ver 131References provided: None Learning Objective:

RECALL AND DISCUSS the Precautions and Limitations(P&L), Notes and Cautions (applicable to the "ReactorOperator") found in the following Procedures (OPS-52101F08).

• SOP-2.1, CVCS Plant Startup and Operation. [...] Question History: NEW K/A match:

There has been a loss of CVCS flow due to a controllerfailure. FK-122 has been placed in manual and is now beingplaced in AUTO.

A reason for placing FK-122 in AUTOand not leaving it in MANUAL is to ensure adequate cooling to the regenerative heat exchanger to preventflashing downstream of the letdown orifices with one 60gpm orifice on service should the controller fail to minimum demand based on current plant conditions.The operational implication would be that flashingwould occur if Chg flow were to fall to <18 gpm and

cause damage to the orifices and piping due to waterhammer and the flashing of water to steam. A loss ofletdown would be the result. This question meets the KAin that it asks the minimum flow rate for being in auto if acontroller were to fail and the reason.SRO justification: N/A

22. 022K4.03 022Unit 1 is operating at 100% power when a Steam Break occurs on 1B SG and thefollowing conditions exist:

• EE5, CTMT ISO PH B, is in alarm.

• All Phase B automatic actions have occurred.Which one of the following completes the statements below?CCW to the RCP Thermal Barrier Heat Exchanger (1) isolated.Seal Injection (2) isolated.

(1)

(2) IS is NOT is NOT is NOT IS IS is NOT IS A.B.C.D.

EE5Automatic Action:1. The following valves closec) CCW FROM RCP THRM BARR Q1P17HV3045f) CCW FROM RCP THRM BARR Q1P17HV3184FSD A181003 3.3.1.2 The seal water injection lines to the RCP are considered as open flow pathspost-LOCA. The high pressure inflow through these lines during the injection andrecirculation phases precludes any containment to atmosphere leakage. In the event ofa loss of seal water flow through these lines, a water seal in the charging pump suctionand discharge piping precludes containment to atmosphere leakage. Distracter analysisA. Correct. First part is correct. Phase B isolates CCW cooling to the RCPThermal Barrier Heat Exchanger. Second part is correct. Seal injection is NOT isolated by SI, PhaseA or Phase B.B. Incorrect. First part is incorrect (See A.1). Plausible if the applicant does notrecall all the components isolat ed on a Phase B. Since CCW iswater solid and cools the thermal barrier hx, they may believe it is not isolated on a phase B. Second part is correct (See A.2).C. Incorrect. First part is correct (See A.1).

Second part is incorrect (See B.2). Plausible if the applicant believes that Phase B isolates RCP seal injection lines. Seal returnis isolated on an SI and the applicant could confuse the two. This is a plausible combination if the applicant reasons that theshutdown seal will actuate and seal injection and CCW to the thermal barrier hx are no longer neededD. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See C.2).

K/A: 022K4.03 Containment Cooling System (CCS) - Knowledge of CCSdesign feature(s) and/or interlock(s) which provide for thefollowing: Automatic containment isolation.Importance Rating: 3.6* 4.0Technical

Reference:

FNP-1-ARP-1.5, EE5, CTMT ISO PH B, Ver 58.0 FSD-A181009, CVCS/HHSI/ACCUM/RMWS, Ver 38 FSD-181003, Containment Isolation System, Ver 26.References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associatedwith the safe operation of the Containment Structure andIsolation System components and equipment, to include thefollowing (OPS-40302B07): * [-]

• [-]

• Automatic actuation including setpoint (example SI, PhaseA, Phase B, MSLIAS, LOSP, SG level)Question History: MOD SUMMER 11 K/A match: Requires the applicant to know the design feature of thePhase B Containment Isolation that isolates CCWcooling to the RCP Thermal Barrier Heat Exchanger.

SRO justification: N/A

23. 025AA1.01 023Unit 1 is in Mode 5 with the following conditions:

• 1B RHR pump is tagged out.

• All SG Wide Range levels are 84%.

• Pzr level is being maintained at 21% on LI-462, PRZR LVL.

• RCS temperature is 155°F.

• RCS pressure is 325 psig.

• All RCP's are secured.

• 1A RHR pump is running in the cooldown lineup.Subsequently, the following occurs:

• 1A RHR pump trips on overcurrent and cannot be restarted.

• RCS temperature is 175°F and slowly rising.Which one of the following completes the statements below?Per AOP-12.0, Residual Heat Removal System Malfunction, the preferredmethod to re-establish core cooling is to establish (1) .Core cooling is monitored using (2) . (1)

(2) feed and bleed RCS cold leg temperaturesa secondary heat sink RCS cold leg temperaturesfeed and bleed CETCsa secondary heat sink CETCs A.B.C.D.ARG-1If the RCS is intact and the loops are not isolated with SG nozzle dams or loop isolationvalves, a secondary heat sink using half or more SGs will be an effective alternatemode of decay heat removal that will last for several hours or longer. Since there wouldbe no significant fluid inventory losses for this case, makeup requirements can easilybe met with a minimum amount of charging flow or possibly RWST (or VCT) gravityfeed if initiated early enough. For this situation, it should also be possible to refill andpressurize the RCS and then operate the RCPs to sweep the noncondensibles fromthe loops and thereby improve the primary-to-secondary heat transfer.AOP-12:24. Check SGs available.

• Check SG primary nozzle dams- REMOVED.

• Check SG primary manways -INSTALLED.

• Check SG secondary handholecovers - INSTALLED.NOTE: Establishing a secondary heat sink will reduce RCS heat up andpressurization rate to provide more time for recovery actions.25. Verify secondary heat sink established.25.1 Maintain wide range level inall available SGs greater than 75% using FNP-1-SOP-22.0, AUXILIARY FEEDWATER SYSTEM.25.2 IF SG steam space intact,THEN open atmospheric reliefvalves to prevent SG pressurization.1A(1B,1C) MS ATMOSREL VLVPC 3371A adjustedPC 3371B adjusted PC 3371C adjusted25.3 IF SGBD system available, ANDAFW system available,THEN establish blowdown fromavailable SGs using FNP-1-SOP-16.3, STEAM GENERATOR FILLING AND DRAINING.Feed and Bleed or Feed and Spill would be established if both of these conditions were met.29.1 Check RCS level LESS than121 ft 11 in AND core exitT/Cs GREATER than 200°F

.Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible if the applicant believesthat establishing a secondary heat sink is not correct because RCStemperature is less than 200°F and so steaming the SG would notbe an option. Feed and Bleed is only used when RCS level is < 121 ft 11 in and RCS temp is > 200°F. Second part is incorrect (See D.2). Plausible since Tcold is used inother procedures (ESP-0.2) to evaluate cooldown and the applicant could believe that it is used here. Also, Tcold would not give an accurate indication of core temperature.B. Incorrect. First part is correct (See D.1). Second part is incorrect (See A.2).

C. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2).D. Correct. First part is correct. Since the RCS is filled and intact, establishinga secondary heat sink is the correct action per AOP-12. Second part is correct. AOP-12 directs the use of CETCs K/A: 025AA1.01 Loss of Residual Heat Removal System (RHRS) - Ability tooperate and / or monitor the following as they apply to theLoss of Residual Heat Removal System: RCS/RHRScooldown rateImportance Rating: 3.6 3.7 Technical

Reference:

Background Information for WOG Abnormal Response Guideline ARG-1 Loss of RHR While Operating at Mid-LoopConditions, Ver 2 FNP-1-AOP-12.0, RHR System Malfunction, Ver 25References provided: None Learning Objective:

LIST AND DESCRIBE the sequence of major actionsassociated with AOP-12.0, RHR System Malfunction and/orSTP-18.4, Containment Closure. (OPS-52520L04)

Question History: MOD FNP EXAM BANK K/A match: Requires the applicant to know how the RCS is operatedto establish a cooldown rate on a loss of RHR and themethod which temperature is monitored.SRO justification: N/A

24. 026AG2.4.50 024Unit 1 is operating at 100% power when the following occurs:

• A leak develops in the CCW system.

• CCW Surge Tank level is slowly lowering.

• AA4 and AB4, CCW SRG TK LVL A(B) TRN HI-LO, are in alarm.

• AA5, CCW SRG TK LVL A TRN LO-LO, has come into alarm.Which one of the following completes the statements below?CCW system automatic isolations are designed to occur at (1) in theCCW Surge Tank.Using the NORMAL source of makeup water, the operator will open

(2) to make up to the CCW Surge Tank.

(1)

(2) 35 inches MOV-3031A, MKUP TO CCW FROM RMW 20 inches MOV-3031A, MKUP TO CCW FROM RMW 35 inches MOV-3030A, MKUP TO CCW FROM DW STOR TK 20 inches MOV-3030A, MKUP TO CCW FROM DW STOR TK A.B.C.D.

ARP AA5: 20 inchesAutomatic Action1. Closes CCW Valves (Q1P17HV3096A&B) to isolate CCW to/fromEvaporator Packages and H2 Recombiners. (Q1P17LSLL3027CD-A)2. Trips closed Q1P17HV2229, CCW to Sample Cooler(Q1P17LSLL3027CD-A).Operator Action4.1 Attempt to fill CCW surge tank using Normal Make-up to maintain level above the lolevel alarm point as follows;4.2. IF unable to fill the CCW Surge Tank per the Normal Make-up method, THENattempt to fill CCW surge tank using Emergency Make-up to maintain level above the lolevel alarm point as follows;5. IF a loss of CCW cooling has occurred, THEN refer FNP-1-AOP-9.0, LOSS OFCOMPONENT COOLING WATER.Distracter analysis A. Incorrect. First part is incorrect (See D.1). Plausible since this is the Surgetank LO Level alarm setpoint. Second part is incorrect (See D.2). Plausible if candidate cannotrecall which of the two makeup sources is the NORMAL source.B. Incorrect. First part is correct (See D.1). Second part is incorrect (See A.2). C. Incorrect. First part is incorrect (See A.1).

Second part is correct (See D.2).D. Correct. First part is correct. AA5 Setpoint is 20 inches which causes theautomatic closure of HV3096A&B and HV2229. Second part is correct. Demin water storage tank is the normalsource for makeup to the CCW surge tank.

K/A: 026AG2.4.50

Loss of Component Cooling Water (CCW) - Ability to verifysystem alarm setpoints and operate controls identified in thealarm response manual.Importance Rating: 4.2 4.0 Technical

Reference:

FNP-1-ARP-1.1 - AA5, CCW SRG TK LVL A TRN LO-LO, Ver 53.1References provided: NoneLearning Objective:

SELECT AND ASSESS the following instrument/equipmentresponse expected when performing CCW System evolutions including the fail condition, alarms, and tripsetpoints (OPS-52102G07).

• Surge Tank LevelQuestion History: NEW K/A match: Requires the applicant to determine at which level theautomatic isolations of the CCW system occur (verifysystem alarm setpoints which is when these valves close) and know what source of water is used to fill thesurge tank (operate controls identified in the ARP toraise the CCW Surge Tank level).SRO justification: N/A

25. 026K2.01 025Unit 2 is operating at 50% power when a simultaneous Dual Unit LOSP occurs.

• 4160V Bus 2G remains de-energized due to the DG not starting for that emergency bus.Three (3) minutes after the LOSP, a Large Break LOCA occurs on Unit 2.

• Containment pressure peaked at 29 psig and is trending down.Which one of the following completes the statement below?The (1) Containment Spray pump is currently running and is powered from the (2) DG. (1) (2) 2A 1C 2B 1C 2A 1-2A 2B 1-2A A.B.C.D.

In the LOSP the 1-2A DG will tie onto the unit 1 A Train busses. Then when the SI andsubsequent phase B signal comes in the 1-2A DG and 1C DG will load shed, and then1-2A DG will align to Unit 2 and the 2A CS pump will start at step 2.FSD-A181008: 3.1.5.2Without offsite power available, the CSS pumps shall start by the diesel generator ESSloading sequencer. Starting will occur at step two of the sequence if the "P" signal ispresent at that time. If the "P" signal occurs between the completion of step two andstep six of the ESS sequence, then starting will occur at the completion of step six ofthe loading sequence. If the "P" signal occurs after the completion of step six, startingwill take place immediately.Pg 2-1 CSS initiation is automatic upon a containment pressure hi-3 signal ("P" signal)FSD-A181005LOSP on both units and LOCA on Unit 2:

For LOSP on both units and LOCA on Unit 2, the alignment of the diesel generators willbe as follows:1-2A Unit 2 Buses 2F and 2K1C Unit 1 Buses 1F, 1K and 1H1B Unit 1 Buses 1G, 1L and 1J2B Unit 2 Buses 2G, 2L and 2JDistracter analysisA. Incorrect. First part is correct (See C.1). Second part is incorrect (See C.2). Plausible since the 1C and1-2A DGs align to either Unit's 'A' Train depending on the scenarioand the applicant may not recall the proper DG alignment for this scenario.B. Incorrect. First part is incorrect (See C.1). Plausible if the applicant does notrecall the CS pump power supplies. Second part is incorrect (See A.2)C. Correct. First part is Correct. The normal power supply to 2A CS pump is2F 4160V AC bus. Second part is correct. For LOSP on both units and LOCA on Unit2, the alignment of the diesel generators will be as follows: 2A Unit 2 Buses 2F and 2K D. Incorrect. First part is incorrect (See B.1). Second part is correct (See C.1).

K/A: 026K2.01 Containment Spray System (CSS) -

Knowledge of buspower supplies to the following: Containment spray pumps.Importance Rating: 3.4* 3.6Technical

Reference:

FSD-A181008, Containment Spray System, Ver 24. A-351199, Unit 2 Load List, Ver 61 FSD-A181005, Diesel Generators, Ver 44. FSD-A181007, Reactor Protection System, Ver 18References provided: None Learning Objective:

NAME AND IDENTIFY the Bus power supplies, for thoseelectrical components associated with the ContainmentSpray and Cooling System, to include those items in Table3- Power Supplies (OPS-40302D04).Question History: NEW K/A match: Requires the applicant to know the normal power supplyto the 2B CS pump and the 1-2A DG alignment and powersupply to the 2A CS pump upon an LOSP with asubsequent SI.SRO justification: N/A

26. 029A1.02 026Unit 2 plant conditions are as follows:* Containment Main Purge system is running.* Containment radiation levels are rising.Subsequently, R-24A, CTMT PURGE, loses control power.

Which one of the following completes the statements below?Radiation levels (1) stop rising in the Main Exhaust Plenum. CTMT Main Purge supply and exhaust fans (2) trip. (1)

(2) WILL will NOT will NOT will NOT WILL WILL will NOT WILL A.B.C.D.

SOP-45:3.5 The radiation monitors fail to a "High Radiation" condition on loss of instrumentand/or control power that will result in actuation of associated automatic functions.FH4: CP RE 24A or B HI RADPROBABLE CAUSE1. High Radiation Level in the Containment Purge Exhaust Line.2. The radiation monitors fail to a "High Radiation" condition on loss of instrumentand/or control power that will result in actuation of associated automatic functions.AUTOMATIC ACTION1. Isolates Containment by closing Purge Supply and Exhaust Valves 2-CP-HV-3196,2-CP-HV-3197, 2-CP-HV-3198A, B, C, & D, 2-CP-HV-2867C & D and 2-CP-HV-2866C& D.Distracter analysisA. Correct. First part is correct. Per SOP-45, a radiation monitor that has lostcontrol power will initiate its automatic actions. Second part is correct. R-24A will NOT automatically secure themain purge supply and exhaust fans.

B. Incorrect. First part is incorrect (See B.1). Plausible if candidate does notrecall that a loss of control power will cause the actuation of associated automatic functions. Second part correct (See B.2). C. Incorrect. First Part is incorrect (See A.1). Second part is incorrect. (See B.1). Plausible since it could seemlogical to the applicant that when the main purge supply andexhaust dampers shut, the fan would also automatically secure.D. Incorrect. First part is correct (See B.1)

Second part is incorrect (See C.2). Plausible if the applicant doesnot recall what auto functions are actuated by R-24A and believesthat R-24A will trip the fans. Additionally, with the Aux Building main exhaust fan running, the applicant could reason that there isstill a negative pressure on the CTMT purge outlet causing CTMT radiation release to the plant vent stack to continue.

K/A: 029A1.02 Containment Purge System (CPS) - Ability to predict and/ormonitor changes in parameters to prevent exceeding designlimits) associated with operating the Containment PurgeSystem controls including:

Radiation levelsImportance Rating: 3.4 3.4 Technical

Reference:

FNP-2-ARP-1.6, FH4, CP RE 24A or B HI RAD, Ver 59 FNP-2-SOP-45.0, Radiation Monitoring System, Ver 38.1References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with theContainment Ventilation and Purge System, to include thoseitems in Table 6-Component Locations (OPS-40304A02).Question History: NEW K/A match: Requires the applicant to predict, based on a loss of control power to R-24A, when the earliest time the radiation release is terminated thereby preventing theoff site radiation exposure limit from potentially beingexceeded.SRO justification: N/A

27. 033K4.05 027Unit 1 is operating at 100% power and the following conditions exist:

• A blended make-up to the Spent Fuel Pool (SFP) is occurring.

• A calibration error results in FT-168, PRI WATER MKUP FLOW, providing a flow input to the Reactor Makeup System that is less than the actual flowrate.Which one of the following completes the statements below?The blended flow makeup resulted in a (1) of the SFP.Per Tech Spec 3.7.14, Fuel Storage Pool Boron Concentration, theMINIMUM required SFP boron concentration is (2) ppm. (1)

(2) boration 2000 boration 2200 dilution 2000 dilution 2200 A.B.C.D.

Tech Specs 3.7.14The fuel storage pool boron concentration shall be

> 2000 ppm.APPLICABILITY: When fuel assemblies are stored in the fuel storage pool.FSD-A181009 5.30.5.1 Flow measurement (FT-168) shall be provided downstream of the blenderto indicate total makeup flow to the charging header an d as input to the ReactorMakeup Control System.5.30.5.2 Upstream boric acid flow measurement (FT-113) shall be provided toserve as input to the Reactor Makeup Control System.5.78.1.1 This differential pressure transmitter (and associated orifice flow element) shallprovide measurement of the total makeup flow from the boric acid blender.

It shall alsoprovide input to the Reactor Makeup Control System for regulation of RMW flowand shall alert the operator of a deviation from the selected flow setpoint.Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible if the candidate thinksthat the lower signal results in less RMW added and thereforemore acid for a given volume which would result in a boration of the SFP. Second part is correct (See C.2).

B. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See C.2). Plausible because the applicantcould confuse the SFP minimum boron concentration with theaccumulators minimum boron concentration which is 2200 ppm.C. Correct. First part is correct. FT-168 will regulate total makeup flow to 120gpm. This is a combination of acid flow and Reactor MakeupWater (RMW) flow. The amount of acid flow will be determined by the properly functioning FT-113. This means that the malfunctioning LOWER signal sent to FK-168 by FT-168 will cause the system to raise the flow of RMW to achieve a "sensed" totalflow of 120 gpm resulting in more RMW than expected therefore alower boron concentration in the makeup water supplied to theSFP. This will result in a dilution of the SFP.

Second part is correct. Tech Spec 3.7.14 requires the SFP boronconcentration to be

> 2000 ppm.D. Incorrect. First part is correct (See C.1). Second part is incorrect (See B.2).

K/A: 033K4.05 Spent Fuel Pool Cooling System (SFPCS) -

Knowledge ofdesign feature(s) and/or interlock(s) which provide for thefollowing: Adequate SDM (boron concentration)Importance Rating: 3.1 3.3 Technical

Reference:

Unit 1 Technical Specifications, Ver 190 D-175043, SH1, Spent Fuel Pool Cooling, ver 27 D-175036, SH 1, Reactor Makeup Water, Ver 22References provided: None Learning Objective:

RECALL AND APPLY the LCO and APPLICABILITY forTechnical Specifications (TS) or TRM requirements, and theREQUIRED ACTIONS for 1 HR or less TS or TRMrequirements, and the relevant portions of BASES thatDEFINE the OPERABILITY and APPLICABILITY of theLCO associated with the Spent Fuel Pool Cooling and Purification and Refueling Water Storage Tank Purification Systems components and attendant equipment alignment,to include the following (OPS-52108L01):

[-]

• 3.7.14, Fuel Storage Pool Boron Concentration RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the ReactorMakeup Control and Chemical Addition System, to include the following (OPS-40301G02): [-]

• Inter connections with other systems

• Primary Water Makeup Flow Controller, FK-168 Question History: MOD NORTH ANNA 08 K/A match: Requires th e applicant to evaluate knowledge of TSrequirement for minimum boron concentration, whichprovides for adequate SDM.

Also evaluates candidatesability to predict effect of an equipment malfunctionwhich could adversely affect ability to maintain desiredboron concentration.SRO justification: N/A

28. 035A2.01 028Unit 1 is operating at 100% power with the following conditions:

• Reactor power is now 100.5% and slowly rising.

• Tavg is 570.5°F and slowly lowering.

• Pressurizer pressure is 2210 psig and slowly lowering.

• Turbine load is 890 MWe and lowering.

• SG pressures are 720 psig and slowly lowering.

• Containment pressure is 2.1 psig and slowly rising.Which one of the following completes the statements below?The event in progress is a (1) line break. Per AOP-14.0, Secondary System Leakage, the operators arerequired to (2) . (1) (2) steam reduce turbine load steam trip the reactor feed reduce turbine load feed trip the reactor A.B.C.D.AOP-14:1. [CA] Evaluate plant status 1. Perform the followingfor safe operation.Pressurizer level 1.1 Verify reactor trippedGREATER THAN 15%

ANDPressurizer pressure 1.2 IF reactor tripped, GREATER THAN 2000 psig THEN CLOSE SG AND main steam isolation andSteam generator pressure bypass valvesGREATER THAN 650 psigANDContainment pressure LESS THAN 2 psigAND IF main generator on line, THEN (check reactor power) - (turbine power + any steam dump power) mismatch LESS THAN 10%.

ANDIF main generator off line, THEN check reactor power less than ~ 15%

AOP-14, step 9 and note above step 9 says:The intent of step 9 is to reduce reactor power to within the capacity of the AFWsystem if possible and step 9 has the crew reduce power per UOP-3.1 and UOP-2.1 ifthe above trip criteria is not met.Distracter analysisA. Incorrect. First part is correct (See B.1). Second part is incorrect (See B.2). Plausible since UOP-3.1, PowerOperation, requires a reduction in turbine load if 100% power isexceeded. This would be the correct thing to do if containmentpressure did not meet the reactor trip criteria. Also Step 7 and 9and note above step 9 addresses ramping the unit down to mode 2if the trip criteria is not exceeded in the previous steps.B. Correct. First part is correct. All the conditions in the stem - Tavg lowering,RCS Pressure lowering and MWe lowering are indicative of asteam break where the steam is exiting the piping before reachingthe turbine. The containment parameters show that the break is incontainment. Second part is correct. With containment pressure greater than 2psig, reactor trip criteria is met.C. Incorrect. First part is incorrect on a feedline break, Rx power would bestable, turbine MWe would be stable, and RCS pressure would bestable. SG pressure would not lower and Tavg would be rising. Plausible if the applicant misdiagnoses the event. AOP-14 addresses a steam or feed break and they have similarcharacteristics. Second part is incorrect (See A.1). D. Incorrect. First part is incorrect (See C.1). Second part is correct (See B.2).

K/A: 035A2.01 Steam Generator System (S/GS) - Ability to (a) predict theimpacts of Faulted or ruptured S/Gs on the S/GS; and (b)based on those predictions, use procedures to correct,control, or mitigate the consequences of those malfunctionsor operationsImportance Rating: 4.5 4.6Technical

Reference:

FNP-1-AOP-14.0, Secondary System Leakage. Ver 10.0References provided: None Learning Objective:

STATE AND EXPLAIN the operational implications for allCautions, Notes, and Actions associated with AOP-14,Secondary System Leakage. (OPS-52521O03)

Question History: MOD VOTGLE 12 K/A match: Applicant must predict that a steam break has occurredbased on the impact to plant parameters (which is abackward logic way to meet the first part of the KA). The parameters include but are not limited to SGs since thetrip criteria in AOP-14 is due to ctmt pressure.

Applicantmust determine the proper procedural response tomitigate a faulted SG inside ctmt.SRO justification: N/A

29. 036AG2.1.7 029Unit 1 is operating at 100%. A fuel shuffle is being performed in the Unit 1 SFP.At 1000:

• EH2, SFP LVL HI-LO, is in alarm.

• SFP level is 153' 3" and stable. At 1015:The SRO in charge of refueling reports that a fuel assembly has beendropped.

• FH5, SFP AREA RE-25 A OR B HI RAD, is in alarm.

• R-25A & B, SPENT FUEL BLDG EXH, reads off scale high. Which one of the following completes the statements below?The operating crew is required to enter (1) .The crew is required to dispatch personnel to (2) per the applicable AOP.Procedure titles are as follows: AOP-30.0, Refueling Accident AOP-49.3, Spent Fuel Pool Emergency (1) (2) AOP-49.3 make up to the SFP using the RWST AOP-30.0 make up to the SFP using the RWST AOP-49.3 ensure all SFP hatches and doors are closed AOP-30.0 ensure all SFP hatches and doors are closed A.B.C.D.AOP-30 Symptoms or entry conditions1. This procedure is entered when a fuel handling accident causes damage to a fuelassembly in conjunction with a high radiation indication on any of the following:[] R-2 CTMT 155 ft[] R-5 SFP ROOM

[] R-24A(B) CTMT PURGE[]R-25A(B) SPENT FUEL BLDG EXHStep 1.6. Dispatch personnel to close all spent fuel area fuel handling hatches.Step 5. Verify all access doors to accident area - CLOSEDStep 20 has the operator makeup to the refueling cavity from the RHR system if thecavity is low. AOP-49.3 B. Symptoms or entry conditions1. A report of damage to and/or leakage from the SPENT FUEL POOL caused by an external threat is received.2. Any condition outside the design basis of the plant that will resultin a long term loss of Spent Fuel Pool cooling.Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible based on the name ofthe procedures. The applicant may believe that since there is norefueling occurring (Unit 1 at 100%), AOP-30 does not apply andAOP-49.3 applies since there is "an emergency" in the SFP. Second part is incorrect (See D.2). Plausible since this is an actionof AOP-49.3 to keep all assemblies covered with water and wouldseem logical since there was a SFP HI-LO level alarm and a damaged fuel assembly lying on the racks. The applicant may thinkthat keeping the damaged assembly covered with water is arequired action. B. Incorrect. First part is correct (See D.1)

Second part is incorrect (See D.2). Plausible since AOP-30 directsfilling the refueling cavity. The applicant could easily confuse thisaction with filling the SFP. It could seem logical since there was aSFP HI-LO level alarm and a damaged fuel assembly lying on the racks. The applicant may think that keeping the damagedassembly covered with water is a required action.C. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2). This is a logical connection toAOP-49.3 since during a SFP Emergency, the applicant couldassume the affected area would be isolated as radiation levels are high.D. Correct. First part is correct. This scenario meets the entry requirements ofAOP-30.0. Second part is correct. This action is taken per step 1.6 and 5.

K/A: 036AG2.1.7 Fuel Handling Incidents - Ability to evaluate plantperformance and make operational judgments based onoperating characteristics, reactor behavior, and instrumentinterpretation.Importance Rating: 4.4 4.7 Technical

Reference:

FNP-1-AOP-30.0, Refueling Accident, Ver 19References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if entry intoAOP-30.0, Refueling Accident is required. (OPS-52521H02) Question History: MOD FNP 05 K/A match: Requires the applicant to interpret plant instrumentation(rad monitors and SFP alarm) and reports from the field and determine the ap plicable procedure to enter and theappropriate action to take (operational judgment).SRO justification: N/A

30. 037AK1.01 030Unit 1 is performing the actions of AOP-2.0, Steam Generator Tube Leakage, due to atube leak on the 1A SG. The following conditions exist:

• RCS pressure is currently being reduced to minimize break flow.The following parameters are observed:

• SG pressures are:

1A SG 1B SG 1C SG 948 psig 905 psig 900 psig - RCS pressure is 916 psig. - The highest reading non-upperhead CETC is 518°F. - PRZR level is 43%.

• BOTH Subcooled Margin Monitors are malfunctioning.Which one of the following completes the statements below?The current value of subcooling is approximately (1) .The RCS pressure reduction (2) required to be stopped.Reference Provided (1)

(2) 22°F IS 22°F is NOT 18°F IS 18°F is NOT A.B.C.D.

AOP-2:Step 34.2[CA] WHEN one of the following conditions occur, THEN stop the RCS pressure reduction.

[] RCS pressure is less than affected SG pressure, AND pressurizer level greater than 15%.

OR

[] Pressurizer level greater than 63%.

OR[]SUBCOOLED MARGIN MONITOR indication less than 16°F subcooled inCETC mode.Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible since the applicant maydetermine subcooling based on ruptured SG pressure vs. RCSpressure. This would be a common misconception since in aSGTR procedure the ruptured SG is the focus for determining SGpressure less than RCS pressure and is the pressure referred towhen determining the required CETC temperature to cooldown to. 948 psig + 15 = 963 psia which is 540°F 540°F - 518°F = 22°F subcooling.

Second part is correct (See C.2). B. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See C.2). Subcooling is greater than16°F, however, RCS pressure is less than the affected SGpressure and pzr level is >15%.Therefore the pressure reduction isrequired to be stopped. Plausible since the subcooling is met andthe Pzr level is <63%, so one more evaluation as to be done. Thisevaluation has two components and one of the two components ismet. Also if the candidate were to compare RCS pressure vs oneof the other two SG pressure, then the pressure reduction would be continued.C. Correct. First part is correct. 916 psig + 15 = 931 psia which is ~536°F 536°F - 518°F = 18°F subcooling.Second part is correct. AOP-2.0 has the operator evaluate 3components, two of which do not require the RCS pressurereduction to be stopped. However, RCS pressure is less thanaffected SG pressure, (RCS pressure is 916 psig and SG pressureis 948 psig) AND pressurizer level greater than 15% (at 43%).D. Incorrect. First part is correct (See C.1) Second part is incorrect (See B.2).

K/A: 037AK1.01

Steam Generator (S/G) Tube Leak - Knowledge of theoperational implications of the following concepts as theyapply to Steam Generator Tube Leak: Use of steam tablesImportance Rating: 2.9* 3.3 Technical

Reference:

FNP-1-AOP-2.0, Steam Generator Tube Leakage, Ver 35 Properties of saturated and superheated steam, 1967References provided: Steam tables and AOP-2.0 step 34.2 Ver 35.0Learning Objective:

ANALYZE plant conditions and DETERMINE the successfulcompletion of any step in AOP-2.0, SG Tube Leakage.

(OPS-52520B07)Question History: MOD CATAWBA 09 K/A match: Applicant is required to use the steam tables to determinecurrent value of subcooling and the whether or not theRCS depressurization is re quired to be stopped duringa SG tube leak scenario.SRO justification: N/A

31. 038EK3.02 031Unit 1 has experienced a tube rupture on the 1C SG.The operating crew is at the step in EEP-3.0, Steam Generator TubeRupture, to "Check SI termination criteria."

• The following plant conditions are observed: - RCS Subcooling is 22°F and slowly rising.

- RCS pressure is 950 psig and slowly rising.

- Pressurizer level is 45% and slowly rising.

- AFW flow is 450 gpm.

- 1A SG NR level is 29% and slowly rising. - 1B SG NR level is 26% and slowly rising.

- 1C SG NR level is 65% and rising rapidly.Which one of the following completes the statements below?SI termination criteria (1) been met.Per EEP-3.0, SI termination is necessary to prevent overfilling the (2) . (1)

(2) has NOT Steam Generator has NOT Pressurizer HAS Steam Generator HAS Pressurizer A.B.C.D.EEP-3 20 [CA] Check SI termination criteria.

20.1 Check SUBCOOLED MARGIN MONITOR indication - GREATER THAN16°F{45°F} SUBCOOLED IN CETC MODE.

20.2 Check secondary heat sink available. Total feed flow to SGs - GREATER THAN 395 gpm AVAILABLE.

Narrow range level in at least one intact SG - GREATER THAN 31%{48%}.20.3 Check RCS pressure - STABLE OR RISING.20.4 Check pressurizer level - GREATER THAN 13%{43%}.

EEB-3Step 20 Basis: [...]

If SI flow is not terminated, leakage into the secondary willeventually fill the steam generator with water and lift the atmospheric reliefvalves. This could damage the relief valve and main steamline which would complicatesubsequent recovery and aggravate the radiological consequences. Hence, SI must beterminated when the criteria in subsequent steps are satisfied to prevent steamgenerator overfillDistracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible if the applicant believesthat due to the SI, adverse numbers are applicable. If they were applicable, then this would be a correct answer due to subcooling. Second part is correct (See C.1) B. Incorrect. First part is incorrect. (See A.1). Second part is incorrect (See A.2). Plausible since this is thereason to terminate SI in EEP-0 and ESP-1.1 for a spurious SI. The applicant could confuse the basis for these procedures withthe basis for the step in EEP-3.C. Correct. First part in correct. SI termination criteria has been met. Second part is correct. EEP-3 background document -

If SI flow isnot terminated, leakage into the secondary will eventually fillthe steam generator with water and lift the atmospheric reliefvalves. This could damage the relief valve and main steamlinewhich would complicate subsequent recovery and aggravate theradiological consequences. Hence, SI must be terminated whenthe criteria in subsequent steps are satisfied to prevent steamgenerator overfillD. Incorrect. First part in correct (See C.2). Second part is incorrect (See B.2) This question was written with these values for the following reasons:RCS subcooling is low but above the 16°F{45°F}. If adverse numbers were used itmakes plausibility greater for this parameter.Przr level is about where you would expect it after cooldown and depress and still above both parameters. 13%{43%} and to meet plausibility for KA.SG NR is below value of Narrow range level in at least one intact SG - GREATERTHAN 31%{48%} but AFW flow is > 395 gpm. One does not meet SI terminationand one does, and SGWL for 1C SG is so high to meet plausibility for KA.

K/A: 038EK3.02 Steam Generator Tube Rupture (SGTR) - Knowledge of thereasons for the following responses as the apply to theSGTR: Prevention of secondary PORV cyclingImportance Rating: 4.4 4.5Technical

Reference:

FNP-1-EEP-3.0, Steam Generator Tube Rupture, Ver 27 FNP-0-EEB-3.0, Specific Background Document For FNP-1/2-EEP-3.0, Ver 2References provided: NoneLearning Objective: STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with EEP-3, Steam Generator Tube Rupture. (OPS-52530D03)Question History: NEW K/A match: Requires the applicant to know that preventing the SGPORVs from lifting due to filling the SGs solid willprevent a radiological release from the atmospherics.SRO justification: N/A

32. 039K1.09 032Concerning R-70A/B/C, 1A/1B/1C SG TUBE LEAK DET, on Unit 1:Which one of the following completes the statements below?The R-70s are located (1) of the MSIVs.A minimum reactor power level that the R-70s can accurately estimate a SG leak rate is (2) . (1)

(2) upstream 25% downstream 25% upstream 10% downstream 10%

A.B.C.D.

FSD-A1810153.2.8 These detectors are located to monitor the main steam lines upstream of thesafety relief valves for the presence of Nitrogen-16 activity in the steam lines and alertthe operator when setpoints are exceeded.SOP-69Step 3.1 -

The system receives a reactor power input from power range channelN-43. IF N-43 fails OR is in Test OR is less than 20% power, THEN the systemcannot accurately estimate a leak rate in the AV mode, and the indicators willdisplay "PN <20%".

If desired, the Counting Room can configure the N-16 system inthe ME counts per second (C/S) mode using FNP-0-CCP-31, LEAK RATE DETERMINATION. While not able to provide a leak rate determination, this mode canbe used to indicate if leakage is increasing based on the indication trending up. The AVmode is the preferred mode of operation above 20% reactor power. The ME modeshould only be utilized below 20% reactor power. Distracter analysisA. Correct. First part is correct. R-70s are located to monitor the main steamlines upstream of the safety relief valves. Second part is correct. R-70s are accurate at reactor power >20%.B. Incorrect. First part is incorrect (See A.1). Plausible if the applicant does notrecall the location of these monitors. Second part is correct (See A.2).C. Incorrect. First part is correct (See A.1).

Second part is incorrect (See A.2). Plausible if the applicant failsto recall the power at which the R-70's are accurate. D. Incorrect. First part is incorrect (See B.1).

Second part is incorrect (See C.2).

K/A: 039K1.09 Main and Reheat Steam System (MRSS) - Knowledge ofthe physical connections and/or cause-effect relationshipsbetween the MRSS and the following systems:

RMSImportance rating: 2.7 2.7Technical

Reference:

FSD-A181015, Radiation Monitoring System, Ver 14 FNP-1-SOP-69, N-16 Primary to Secondary Leak DetectionSystem, Ver 5 D-175033, SH1, Main and Aux Steam, Ver 38References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the RadiationMonitoring System to include those items in Table 4-Remote and Local Indications and Controls (OPS-40305A02). RECALL AND DESCRIBE the physical in-plant location ofthose components associated with the Radiation MonitoringSystem, to include those items in Table 4- Remote andLocal Indications and Controls (OPS-40305A03).Question History: MOD ROBINSON 04 K/A match: The applicant is required to know the physical location/connection of the R-70s in relation to the main steamsystem and the cause-effect (power level) of when theRad monitors go into alarm.SRO justification: N/A

33. 045K1.19 033Which one of the following coincidences will cause an anticipated transientwithout trip (ATWT) mitigation system actuation circuitry (AMSAC) Main Turbine Trip? (1) Turbine impulse pressure channels > 40%

AND (2) SG NR levels < 10% for > 25 seconds.

(1)

(2) 1 of 2 2 of 3 2 of 2 2 of 3 1 of 2 1 of 3 2 of 2 1 of 3 A.B.C.D.

FSD- A181007 pg 2-37C-20 Interlock. Control interlock C-20 is used to enable the Anticipated TransientWithout Trip (ATWT) Mitigation System Actuation Circuitry (AMSAC) . When turbineload is > 40% on 2/2 turbine impulse channel detectors and steam generator narrowrange water level decreases to <10% on 2/3 steam generators for 25 seconds, theAMSAC system will trip the main turbine and provide an auto start signal to all AFWpumps. There is a time delay drop out associated with the impulse pressure portion ofthe signal such that for 260 sec after impulse pressure decreases below 40%,AMSAC is still enabled.A. Incorrect. First part is incorrect (See B.1). Plausible since various control andpermissive interlocks use a 1 of 2 logic to enable or disablefunctions. The applicant could confuse AMSAC (C-20) with any of these. Second part is correct (See B.2).B. Correct. First part is correct. 2 of 2 turbine impulse channels > 40% enablesAMSAC. Second part is correct. 2 of 3 SG NR levels < 10% for > 25%actuates AMSAC.C. Incorrect. First part is incorrect (See A.1).

Second part is incorrect (See B.2). Plausible because the Low LowSGWL is 1 of 3 SGWL less that 28% NR. The applicant couldimproperly believe that AMSAC is 1 of 3 as is the Low Low SGWLlogic.D. Incorrect. First part is correct (See B.1). Second part is incorrect (See C.2).

K/A: 045K1.19 Main Turbine Generator (MT/G) System - Knowledge of thephysical connections and/or cause-effect relationshipsbetween the MT/G system and the following systems:

ESFASImportance Rating: 3.4* 3.6 Technical

Reference:

FSD-A181007, Reactor Protection System, Ver 18References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associated with the safe operation of the Main Turbine and Auxiliaries System components and equipment, to include the following(OPS-40202A07):

[-]

• Turbine Trips

Actions needed to mitigate the consequence of theabnormality.Question History: MOD CALLOWAY AUG 05 K/A match: AMSAC is listed as a back up to the reactor trip system andESFAS in the FSAR. This question requires the applicant toknow the cause and effect of relationship between AMSACand the Main Turbine. Conditions which cause AMSAC tobe enabled and produce a turbine trip.SRO justification: N/A

34. 051AK3.01 034Unit 1 is operating at 40% power when PR-4029, CONDENSER PRESSUREindicates as follows:* PT0501 and PT0502 are 6 psia and rising rapidly.Subsequently, Condenser pressure stabilizes at 12 psia.Which one of the following completes the statements below?The Steam Dump (1) controller is enabled.The Steam Dumps are (2) . (1)

(2) Plant Trip CLOSED Plant Trip OPEN Loss of Load CLOSEDLoss of Load OPEN A.B.C.D.

FSD-A181007 Pg 2-36/37C-9 Interlock. C-9 is the condenser-available interlock. This interlock allows the steamdump valves to be armed if the condenser is available. It also prevents an overpressurecondition which could damage the condenser. To activate C-9, both condenser pressures shall be < 8 inches Hg vacuum, and 1/2 circulation water pump motorbreakers must be shut.8 inches of Hg vacuum is 10.8 psia.See references Figure 2, Sheet 10 of FSD-A181007.Distracter analysisA. Correct. First part is correct. A turbine trip results which causes a reactortrip, thus enabling the plant trip controller. Second part is correct. C-9 is NOT enabled at 12 psia therefore thesteam dumps do not operate and are closed.B. Incorrect. First part is correct (See A.1). Second part is incorrect (See A.2) Plausible if the applicant cannotrecall that the vacuum setpoint for the C-9 interlock is <10.8 psia and believes that adequate condenser vacuum exists for steamdump operation.C. Incorrect. First part is incorrect (See A.1). Plausible if the applicant fails torecognize that the turbine trip causes a reactor trip at this power. Ifrx power were less than 35% then a rx trip would not occur and theturbine trip would cause the LOL controller to be the controlling controller. Second part is correct (See A.2).D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2)

K/A: 051AK3.01 Loss of Condenser Vacuum - Knowledge of the reasons forthe following responses as they apply to the Loss ofCondenser Vacuum:

Loss of steam dump capability uponloss of condenser vacuumImportance Rating: 2.8* 3.1 Technical

Reference:

FSD-A181007, Reactor Protection System, Ver 18References provided: NoneLearning Objective:

RELATE AND IDENTIFY

the operational characteristicsincluding design features, capacities and protectiveinterlocks for the following components associated with theSteam Dump System to include the components found onFigure 5, Steam-Dump Control (OPS-52201G02).Question History: FNP EXAM BANK K/A match: Requires the applicant to

know on a loss of vacuum whichcontroller the steam dumps will operate on and the reasonthe steam dumps will not operate (loss of capability). On aloss of vacuum the reason is because the C-9 interlock(vacuum) is not met. This is not stated in the stem but isinherent to the question.SRO justification: N/A

35. 054AA2.05 035Unit 1 is operating at 4% power. The following conditions exist:

• 1A SGFP is running.

• All SG NR levels are in the programmed band.

• FCV-479/489/499,1A/1B/1C SG FW BYP FLOW, controllersare in MANUAL and 35% open.Subsequently, the 1A SGFP trips.

Which one of the following completes the statements below?MOV-3232A/B/C, MAIN FW TO 1A/1B/C SG, will (1) .FCV-479/489/499,1A/1B/1C SG FEED FLOW BYPASS FCVs, will (2) . (1)

(2) remain OPEN remain OPEN remain OPEN CLOSE CLOSE remain OPEN CLOSE CLOSE A.B.C.D.

For this event the applicant has to analyze the situation. For a SGFP trip, AOP-13 isrequired to be entered and a Rx trip is initiated >5% power. If the applicant thought theRx was tripped, then the dumps would be controlling at 547°F and a FWI signal wouldbe generated. This would directly affect the bypass valves. Since the bypass valvesare rarely used, an applicant may not realize the link and open/close signals.Since we are <5% power, the RTBs are not opened and AFW will auto start to raiseSGWL due to both SGFPs tripped. This will keep level high. MOV-3232A/B/C close when both SGFPs are tripped. This has to be analyzed and known for these twoparticular valves.Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible if the applicant thinksthat this valve only automatically shuts on a feedwater isolation(FWI). A FWI has NOT occurred at this time. Second part is correct (See C.1). B. Incorrect. First part is incorrect (See A.1). Second part is incorrect. (See C.1). Plausible if the applicant thinksthat a FWI has occurred.C. Correct. First part is correct. D175073, Sheet 1 shows that these valvesclose on a SGFP trip.

Second part is correct. The bypass valves are in manual andtherefore remain open since there is NO feedwater isolation (FWI).

A FWI occurs with a P-4 signal (Rx Trip) coincident with a lowTavg, Safety Injection and a Hi-Hi SGWL (P-14).D. Incorrect. First part is correct (See C.1). Second part is incorrect (See B.2).

K/A: 054AA2.05 Loss of Main Feedwater (MFW) - Ability to determine andinterpret the following as they apply to the Loss of MainFeedwater (MFW): Status of MFW pumps, regulating andstop valvesImportance Rating: 3.5 3.7 Technical

Reference:

D-175073, SH 1, Main Feedwater System, Ver 18 FSD-181007, Reactor Protection System, Ver 18References provided: None Learning Objective:

EVALUATE plant conditions and DETERMINE if entry intoAOP-13, Loss of Main Feedwater is required.(OPS-52520M02)

Question History: NEWK/A match: Requires the applicant to determine the status of feedsystem STOP valves and by pass FCVs upon a loss ofMain Feedwater.SRO justification: N/A

36. 055A3.03 036Unit 1 is at 70% power with the following conditions:

• R-15A, SJAE EXH, is in alarm.

• R-15B, TURB BLDG VNTL, is in alarm.

• AOP-2.0, Steam Generator Tube Leakage, is in progress.

• The Turbine Building SO has placed the SJAE Filtration System in service.Which one of the following completes the statement below?After the SJAE Filtration system is placed in service, the reading on R-15B will (1) and the SJAE Filtration system will (2) . (1)

(2) decrease be aligned in a recirc alignment remain the same be aligned in a recirc alignment decrease discharge to the Turbine Bldg roofremain the same discharge to the Turbine Bldg roof A.B.C.D.

D170064/D-175027:These drawings show that R-15A is upstream of the normally off service SJAE filtrationsystem. R-15B is downstream of the SJAE filtration system. When the SJAE filtrationsytem is placed on service, R-15B reading will decrease. The SJAE filtration systemdischarges directly to the turbine building roof and cannot be diverted elsewhere.Distracter analysisA. Incorrect. First part is correct (See C.1). Second part is incorrect (See C.2). Plausible since the turbinebuilding ventilation system is capable of bypassing the SJAE filter system which is similar to a recirc alignment. Recirc would seem reasonable to minimize radioactive release to the outsideatmosphere. The SJAE filtration system discharges directly to theturbine building roof and cannot be diverted elsewhere. ThePenetration Room Filtration system on the rad side does haverecirc alignment MOVs and a student could confuse the twosystems or apply the concepts from one system to the other.B. Incorrect. First part is incorrect (See C.1). Plausible if the applicant cannotrecall the location of R-15B and believes it is upstream of the SJAEfiltration system. R-15A is located before the SJAE filtration systemand remain the same would be the correct answer. Second part is incorrect (See A.2).C. Correct. First part is correct. R-15B is downstream of the SJAE filtrationsystem. When the SJAE filtration system is placed on service,R-15B reading will decrease. Second part is correct. The SJAE filtration system dischargesdirectly to the turbine building roof and cannot be divertedelsewhere.D. Incorrect. First part is incorrect (See B.1).

Second part is correct (See C.2).

K/A: 055A3.03 Condenser Air Removal System (CARS) - Ability to monitorautomatic operation of the CARS, including: Automaticdiversion of CARS exhaustImportance Rating: 2.5* 2.7*

Technical

Reference:

D-170064, SH1, Condenser Vacuum System, Ver 19 D-175027, SH 1, HVAC: TUBINE BLDG, Ver 21References provided: NoneLearning Objective:

LABEL, DRAW AND ILLUSTRATE the Condensate andFeedwater System flow paths, to include the components onthe following figures (OPS-40201B05, Part A):

• Figure 3, Condenser Air Removal System Question History: MOD FNP 11 K/A match: Requires the applicant to monitor the R-15B reading anddetermine its response when the SJAE filtration systemis placed on service.

FNP has no automatic diversion ofthe SJAE exhaust system. (10/24/12) Chief Examiner saidusing manual diversion based on our plant design isacceptable.SRO justification: N/A

37. 056AK3.02 037Unit 1 is operating at 100% when a LOSP occurred. The following conditions exist:

• The Emergency Diesel Generators failed to energize the ESF busses.

• The operating crew is conducting a secondary depressurization per ECP-0.0, Loss Of All AC Power.

• SG pressures are as follows:

- 1A SG: 245 psig and lowering

- 1B SG: 247 psig and lowering

- 1C SG: 244 psig and loweringWhich one of the following completes the statements below?Per ECP-0.0, this secondary pressure reduction is required to (1) .The reason the secondary pressure reduction is required to be stopped atthe SG pressure specified in ECP-0.0 is to prevent (2) . (1)

(2) be STOPPED injection of accumulator nitrogen into the RCS CONTINUE injection of accumulator nitrogen into the RCS be STOPPED a challenge to the Integrity Critical Safety Function CONTINUE a challenge to the Integrity Critical Safety Function A.B.C.D.

ECP-0.0: 17. Reduce intact SGs pressure to 260 psig.ECB-0.0:The target SG pressure for Step 16 should ensure that RCS pr essure is above theminimum pressure to preclude injection of accumulator nitrogen into the RCS. Thetarget SG pressure should be based on the nominal SG pressure to preclude nitrogenaddition, plus margin for controllability (e.g., 100 psi).Distracter analysisA. Correct. First part is correct. Per ECP- 0.0, Reduce intact SGs pressure to260 psig. Second part is correct: Per ECB-0.0, [...] Should ensure that RCSpressure is above the minimum pressure to preclude injection ofaccumulator nitrogen into the RCS.B. Incorrect. First part is incorrect (See A.1). Plausible since the limit in thebackground document is 160 psig. The limit in the procedure addsa 100 psig for margin of controllability and the applicant couldconfuse these two numbers and believe that the depressurizationmust continue.

Second part is correct (See A.2).C. Incorrect. First part is correct (See A.1) Second part is incorrect (See A.2). Plausible since this is thereason for the Tcold temperature limit of 280°F during the pressurereduction but NOT the reason for stopping at 260 psig.D. Incorrect. First part is incorrect (See B.1) Second part is incorrect (See C.2)

K/A: 056AK3.02 Loss of Offsite Power - Knowledge of the reasons for thefollowing responses as they apply to the Loss of OffsitePower: Actions contained in EOP for loss of offsite power Importance Rating: 4.4 4.7 Technical

Reference:

FNP-1-ECP-0.0, Loss Of All AC Power, Ver 26 FNP-0-ECB-0.0, Specific Background Document for FNP-1/2-ECP-0.0, Ver 3.1References provided: None Learning Objective:

STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with (1) ECP-0.0, Loss of All AC Power;[...] (OPS-52532A03)Question History: NEW K/A match: This question presents a scenario where a Loss of OffsitePower occurs and the Emergency DGs fail to energize theESF busses. The Applicant is required to know the reasonthat the secondary depressurization is stopped at 260psig (reasons for the actions contained in the EOP).SRO justification: N/A

38. 059A4.01 038Unit 1 is operating at 100% power when the 1B SGFP trips.Which one of the following completes the statements below for the 1B SGFP?The HIGH PRESS. GOV. VALVE CLOSED light is (1) . The LOW PRESS. GOV. VALVE CLOSED light is (2) . (1)

(2) LIT LIT LIT NOT lit NOT lit LIT NOT lit NOT lit A.B.C.D.U-161792 - Tab 3, Section 5, Page 1When a trip condition occurs, signals from the electronic controller close the steamvalves.Ran on desktop simulator. Inserted SGFP trip from 100% power and both governorvalves went closed.From OPS-52104C Ver 2 pg 15 -

Initially, as the feed pump turbine accelerates from operation on the turning gear tooperating speed, both the LP and HP stop valves are open. The first governor valve toopen on an increase speed signal from the control system is the LP governor. Sincereheat steam is not available, the turbine speed does not in crease. Once the LPgovernor valves begin to reach their fully open position, the HP governor valve begins to open. The turbine now accelerates to the demanded speed using the main steamsupply via the HP governor valve. As main turbine load is increased, reheat steam pressure in the shell side of the MSRsalso increases. At approximately 25 percent main turbine power, the reheat steampressure is high enough to cause the feed pump turbine speed to increase. In an effortto maintain the desired feed pump turbine speed, the control system begins to shut theHP governor valve. Once the HP governor valve approaches the fully shut position, thecontrol system starts closing the LP governor valves. During 100 percent poweroperation, the governor valve alignment is as follows: 1. The HP governor valve is fully shut. 2. The LP governor valve is throttled partially shut and consequently controls feedpump turbine speed.* Some validators selected the correct answer but stated that they struggled withdetermining the response of governor valves when the SGFP tripped.Distracter analysis A. Correct. First part is correct. The Servo controller shuts the HP governorvalve when the SGFP turbine trips. Second part is correct. The Servo controller shuts the LP governorvalve when the SGFP turbine trips.B. Incorrect. First part is correct (See A.1). Logical connection to the secondpart because this is the normal position of the HP governor valve at 100% power. Second part is incorrect (See A.2). Plausible if the applicant thinksthat the HP and LP Governor Valves remain in their pre-trip positions on a SGFP trip because the HP and LP STOP valves goshut. C. Incorrect. First part in incorrect (See A.1). Plausible if the applicant thinksthat the HP and LP Governor Valves remain in their pre-trippositions on a SGFP trip because the HP and LP STOP valves goshut. If the applicant thought the HP governor valve controlledspeed at high power then it would make this a plausible correct answer coupled with the second part. Second part is correct (See A.2). Logical connection to the first partif the applicant thinks that the HP governor valve controls speed athigh power. D. Incorrect. First part is incorrect (See A.1) Plausible if the applicant thinks thatthe HP and LP Governor Valves remain in their pre-trip positionson a SGFP trip because the HP and LP STOP valves go shut. Logical connection to the second part if the applicant thinks thatboth the HP and LP governor valves are open at 100% power. Second part is incorrect (See A.2) Plausible if the applicant thinksthat the HP and LP Governor Valves remain in their pre-trippositions on a SGFP trip because the HP and LP STOP valves goshut. Logical connection to the first part if the applicant thinks thatboth the HP and LP governor valves are open at 100% power.

K/A: 059A4.01 Main Feedwater (MFW) System - Ability to manually operateand monitor in the control room: MFW turbine trip indicationImportance Rating: 3.1* 3.1*Technical

Reference:

U-161792, SGFP Drive Turbine and Accessories, Ver 12References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if entry intoAOP-13, Loss of Main Feedwater is required.

(OPS-52520M02)Question History: DIABLO CANYON 12-07 K/A match: Requires the applicant to monitor MCB indications anddetermine the proper SGFP GOV valve positions on aSGFP trip.SRO justification: N/A

39. 059AK2.01 039Unit 2 is operating at 100% power when a SG tube leak occurs.Which one of the following completes the statement below?When R-23A, SGBD HX OUTLET, alarms, it will cause

, to automaticallyclose.HV-7614A/B/C, 2A/B/C SGBD ISOFCV-1152, SGB INLET STOP VALVERCV-023B, SGBD DISCH TO ENVIRONMENTHV-7697A/B, 7698A/B and 7699A/B, 2A/B/C SGBD ISO A.B.C.D.

FH1Automatic Actions:R23A :( Steam Generator Blowdown Processing) closes 2-BD-FCV-1152 S/GBlowdown Heat Exchanger Discharge Valve.R-23A is in the SGBD line after FCV-1152 but before the SGBD Surge tank. R-23B isafter the surge tank and is the last rad monitor and isolation signal before an accidentalrelease would make it to the environment. R-23 A and B are often confused on thefunctions and locations. Two other sets of valves isolate SGBD due to otherconditions, such as High Penetration room pressure and AFW autostart. All of thesevalves complete the same function but for different reasons.Distracter analysisA. Incorrect. See B. Plausible since these valves will isolate SG Blowdown(SGBD) and automatically close on an AFW autostart. Theapplicant could believe they also close on a high radiation signal.B. Correct. R-23A automatically closes FCV-1152. FCV-1152 also closes onSGBD high ST level, High pressure in the SGBD system and High flow.C. Incorrect. See B. Plausible since R-23B automatically closes RCV-23B andthe applicant could confuse which radiation monitor closes whichvalve. RCV-023B will isolate SGBD to the environment and isdownstream of FCV-1152.D. Incorrect. See B. Plausible since these valves are two series isolation valveslocated inside the containment on each line from the steam generator. The air-operated isolation valves (7697A/B, 7698A/B,7699A/B) automatically close when high pressure (0.28-0.33 psig)is sensed in any room outside the containment where the blowdown piping, upstream of the heat exchanger, is located.

Since these valves isolate on High pressure in the PPRs, they could be confused with closing signals for FCV-1152.

K/A: 059AK2.01 Accidental Liquid Radwaste Release - Knowledge of theinterrelations between the Accidental Liquid Radwaste Release and the following: Radioactive-liquid monitorsImportance Rating: 2.7 2.8Technical

Reference:

FNP-2-ARP-1.6, FH1 - RMS HI-RAD, Ver 70 References provided: NoneLearning Objective:

RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the RadiationMonitoring System to include those items in Table 4-Remote and Local Indications and Controls(OPS-40305A02).

Question History: FNP 06 K/A match: In this scenario, a SG tube leak results in an accidentalliquid radwaste release.

The applicant is required toknow the interrelations between R-23A and the SGBD system that will terminate the accidental liquid radwaste release.SRO justification: N/A

40. 059G2.2.44 040Unit 1 is operating at 33% power and the following conditions exist:

• 1A and 1B Condensate pumps are running.

• 1C Condensate pump is in OFF with a CAUTION TAG that says, "EMERGENCY USE ONLY."

• 1A SGFP is running.Subsequently, the 1B Condensate pump trips and the following conditions areobserved:KB4, SGFP SUCTION PRESS LOW, comes into alarm and the operating crewobserves the following on PR4039, SGFP SUCT PRESS:Time 0 sec 10 sec 20 sec 30 sec 40 sec300 psig 275 psig 265 psig 270 psig 285 psigAt time 20 seconds, the 1C condensate pump was started.Which one of the following completes the statements below?At time 30 seconds, the 1A SGFP (1) be tripped.The operating crew is required to (2) .1) will NOT2) rapidly reduce Turbine load using AOP-17.1, Rapid Turbine Power Reduction1) will NOT2) check SGFP suction pressure stabilizes1) WILL2) trip the Reactor and enter EEP-0.0, Reactor Trip or Safety Injection.1) WILL

2) trip the Main Turbine and enter AOP-3.0, Turbine Trip Below P-9 Setpoint.

A.B.C.D.

Not a true 2+2 question to improve the plausibility of the distracters.KB4 comes into alarm at 300 psig.At 275 psig decreasing on 2/3 pressure switches (PS625, PS626, PS627),1. The standby condensate pump will start after 10 sec delay. (63IP relay)2. The SGFP(s) will trip after 30 sec. delay (63IPX relay).AOP-136.1 Check SGFP suction pressure stabilizes above 275 psig.6.1.2 RNO:IF suction pressure still falling, THEN reduce turbine load rapidly usingFNP-1-AOP-17.1, RAPID TURBINE POWER REDUCTION.Distracter analysisA. Incorrect. First part is correct (See B.1). Second part is incorrect (See B.2). Plausible since this is the actionto take if the SGFP suction does NOT stabilize (6.1.2 RNO).B. Correct. First part is correct. The SGFP's will trip 30 seconds after suctionpressure falls below 275 psig which would be at 40 seconds in this scenario. Second part is correct. This is the correct action per AOP-13 step

6.1 since

suction pressure is rising and within the band to keep theSGFP from tripping at time 40 secC. Incorrect. First part is incorrect (See B.1) Plausible if the applicant confusesthe condensate pump autostart setpoint with the SGFP trip. Thestandby condensate pump, if in AUTO, would start 10 secondsafter SGFP suction pressure falls below 275 psig. OR plausible if the applicant thought that when the low pressurealarm comes in the SGFP would trip 30 sec later. Second part is incorrect (See B.2) Plausible since this is thecorrect response if the SGFP tripped.D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2) Plausible since power is less than 35% (P-9) and tripping the turbine would stop most of the steam flow from the SG. This wa s the correct action s to take until 2 years ago when the station decided the most conservative action would be to trip the reactor if power is >5% power K/A: 059G2.2.44 Main Feedwater System - Ability to interpret control roomindications to verify the status and operation of a system,and understand how operator actions and directives affectplant and system conditions.Importance Rating: 4.2 4.4 Technical

Reference:

FNP-1-AOP-13, Condensate and Feedwater Malfunction,Ver 33 References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing AOP-13,Loss of Main Feedwater. (OPS-52520M06).

Question History: NEW K/A match: This question requires the applicant to interpret the SGFPsuction pressure to verify the status of the MFW system(SGFP is not tripped). Also, understand how operator actions, starting the 1C cond pump, and directives(AOP-13.0) affect the plant status which is to check thatthe suction pressure is rising and not reduce power ortrip the reactor. SRO justification: N/A

41. 061A3.01 041Unit 1 is in Mode 3 with the following conditions:

• 1A MDAFW pump was started per UOP-1.2, Startup of Unit From Cold Shutdown to Hot Standby.

• There are no other AFW pumps running.

• All SG NR levels are 65%.Subsequently, power is lost to the 1A Startup Transformer.

Which one of the following completes the statements below?The TDAFW Pump (1) be running.Total design AFW flow rate will be approximately (2) gpm. (1)

(2) will NOT 350 will NOT 700 WILL 700 WILL 1050 A.B.C.D.Not a true 2+2 question to improve distracter plausibility.The applicant has to evaluate how the loss of the 1A Startup transformer (SUT) affectsthe TDAFW and MDAFW pumps auto starts. Since the 1A SUT powers the 1A bus andthe 1B SUT powers the 1B and 1C busses, only the 1A bus is lost. The opposite is trueon Unit 2 so the applicant has to recall how each unit is configured.Secondly, the applicant has to recall how the loss of power affects the MDAFW pumps.The 1B MDAFWP is unaffected since the 1G bus did not lose power as it is poweredfrom 1B SUT. The 1A MDAFW pump did lose power and will be sequenced on the buswhen the DG starts and the LOSP sequencer runs.Thirdly, once the applicant determines which AFW pumps are running, then they willhave to recall design flow rates for each (350 gpm for the MDAFW and 700 gpm for theTDAFW pump) to determine total approximate flow. Distracter analysisA. Correct. First part is correct. FSD-A181010 - 3.9.2.3 - The TDAFW pumpshall start by opening the steam supply valves to the turbine driveon a loss of power signal, low-low water level signals from twoout of three level transmitters of any two out of three steamgenerators, or an AMSAC signal. The loss of power signal comesfrom the loss of power to 2 of 3 RCP busses (1A, 1B and 1C).Since ONLY the 1A bus loses power, the TDAFW pump does not start.) Second part is correct. The 1A MDAFWP pump will autostart andthe FCV's will open fully providing ~350 gpm design flow.

.B. Incorrect. First part is correct (See A.1). Second part is incorrect (See A.2). Plausible if the applicant incorrectly thinks an LOSP has occurred which would start 2 MDAFWP's and provide approx 700 gpm flow.C. Incorrect. First part is incorrect (See A.1). Plausible if the applicant confusesthe Startup transformer alignment with Unit 2. The TDAFW pump on Unit 2 would start under these conditions. Second part is incorrect (See A.2). Plausible if the applicantdoesn't recall that the MDAFWP receives an auto start signalduring an LOSP. This would make this a logical connection to thefirst part and a correct answer if the applicant thought that only theTDAFW pump started.D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See A.2).Plausible since this is the designflow for one MDAFW pump and the TDAFW pump and a logicalconnection to the first part if the applicant thought that the TDAFWpump started.

K/A: 061A3.01 Auxiliary / Emergency Feedwater (AFW) System - Ability tomonitor automatic operation of the AFW, including: AFWstartup and flowsImportance Rating: 4.2 4.2 Technical

Reference:

FSD-A181010, Auxiliary Feedwater System, Ver 25 FSD-A181007, Reactor Protection, Ver 18 U166235, Primary Coolant Trip Signals, Ver 2 A506250, U1 Load List, Ver 74References provided: None Learning Objective:

RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the AFWSystem to include the components found on Figure 2,Auxiliary Feedwater System, Figure 3, TDAFWP Steam Supply, and Figure 4, Air Supply to TDAFWP SteamAdmission Valves (OPS-40201D02).

NAME AND IDENTIFY the Bus power supplies (Off-sitesources and Emergency source-to- Load), for those electrical components associated with the AFW System toinclude those items in Table 3- Power Supplies(OPS-40201D04).

SELECT AND ASSESS the AFW Systeminstrument/equipment response expected when performingauxiliary feedwater evolutions including (OPS-52102H05): [...]

The Failed Condition [...] Associated Trip Setpoint(s)

[...]Question History: NEWK/A match:

The applicant is required to evaluate the loss of powerand determine which AFW pump auto starts (monitorstartup) and the resultant flow (monitor flow).SRO justification: N/A

42. 061K5.01 042Unit 1 was operating at 100% power when a Reactor Trip occurred and the followingconditions exist:

• ESP-0.1, Reactor Trip Response, has just been entered.

• Pressurizer level is 12% and slowly lowering.

• SG NR levels are 40% and slowly rising.

• Tavg is 534°F and slowly lowering.

• RCS pressure is 2050 psig and slowly lowering.Which one of the following actions will be performed FIRST as required by ESP-0.1 toaddress the cooldown?Minimize total AFW flow.Emergency borate the RCS.Close all MSIVs and MSIV Bypass Valves.Manually initiate SI and return to EEP-0.0, Reactor Trip or Safety Injection.

A.B.C.D.ESP-0.1 -

Step 1.1 RNO:IF RCS temperature less than 547°F and falling, THEN perform the following. IF NOT,THEN proceed to RNO Step 1.2.Step 1.1.4 RNO:IF cooldown continues, THEN minimize total AFW flow.Distracter analysisA. Correct Step 1.1.4 RNO of ESP-0.1 has the operator minimize AFW tostop the cooldown.B. Incorrect. See A. Plausible since this is an action in ESP-0.1 if Tavg fallsbelow 525°F. This action is at step 4 and would not be requiredsince Tavg is >525°F. The applicant could confuse thistemperature limit to emergency borate with P-12, 543°F Lo-Lo Tavg.C. Incorrect. See A. This is done AFTER AFW flow is reduced at step 1.1.5 .Plausible since this would address the cooldown. Also there are a number of steps completed before the AFW flow is addressed thatequates to steam in the TB reduced, and stm dumps checked.D. Incorrect. See A. Plausible because the Pzr level meets the SI reinitiationcriteria (13%) for a number of other Emergency procedures (suchas ESP-1.1) and the applicant could confuse it with the correct Pzrlevel SI initiation criteria of ESP-0.1 foldout page of 4%.

K/A: 061K5.01 Auxiliary / Emergency Feedwater (AFW) System -Knowledge of the operational implications of the followingconcepts as the apply to the AFW: Relationship betweenAFW flow and RCS heat transferImportance Rating: 3.6 3.9 Technical

Reference:

ESP-0.1, Reactor trip Response, Ver 32.References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing ESP-0.1,Reactor Trip Response. (OPS-52531B06)Question History: INDIAN POINT 07 K/A match: The applicant is required to know that in order to stop theexcessive RCS cooldown (operational implication), theymust know that reducing AFW flow will reduce the heattransfer rate of the RCS.SRO justification: N/A

43. 062AA1.07 043Unit 2 is operating at 100% power. The following conditions exist:

• SGBD is on service. * #1 WMT release is in progress.

• The service water pond level has dropped to 179 feet, 10 inches.Which one of the following combinations predicts the plant response to the change inpond level?1) SW Dilution Flow on FR-4107, SW DILUTION FLOW, will lower;2) RCV-023B, SGBD DISCH TO ENVIRONMENT, will automatically close.1) SW Dilution Flow on FR-4107, SW DILUTION FLOW, will lower; 2) RCV-018, WMT DISCH TO ENVIRONMENT, will automatically close.1) SW Pressure on PI-3001A & B, SW TO CCW HX HDR PRESS, will lower;2) PCV-562 and 563, TRN B (A) DILUTION BYPASS PCV, will fully open.1) SW Pressure on PI-3001A & B, SW TO CCW HX HDR PRESS, will lower;2) MOV-538 and 539, SW B (A) HDR EMERG RECIRC TO POND, will fully open.

A.B.C.D.Not a true 2+2 to improve distracter plausibility.SOP-16.14.5 Defeating the Low SW Dilution Flow Trip of N2G24RCV023B NOTES* At low dilution flow below 14,500 GPM SGBD will isolate [...]AOP-31 2. At a pond level of 180 ft 0 in the following sequence of events occurs

• SW A(B) HDR EMERG RECIRC TO POND valves on both units will open.

• SW HDR NORMAL DISCH ISO A(B) TRN valves on both units will close.

• SW TO WET PIT EAST(WEST) HDR ISO valves will open.

• SW TO POND EAST(WEST) HDR ISO will partially close to divert approximately 50% of the SW recirculation flow to the wet pit.Ran on desk top simulator and Discharge pressure ROSE ~1.5 psigDistracter analysisA. Correct First part is correct. When the pond level drops to 180 ft 0 in, theSW the SW system changes valve alignments such that theemergency recircs to the pond open and the discharges from eachtrain closes which lowers the dilution flow as seen on FR-4107.

Second part is correct. The dilution line flow drops to less than14,500 gpm (goes to 0 gpm), which in turn causes the auto-closureof RCV-023B, terminating this Release path. See D200013 for line up.B. Incorrect. First part is correct (See A.1) Second part is incorrect (See A.2). RCV-018 does not have a lowdilution line auto closure. Plausible since RCV-023B is also aradioactive release point isolation and will close on both HighRadiation and Low Flow so the app licant could think it also closedon low flow.C. Incorrect. First part is incorrect (See A.1). SW discharge pressure is virtuallyunchanged due to the lineup. When run on desk top simulator,pressure ROSE ~1.5 psig. Plausible since the SW header willoperate on RECIRC back to the POND, the applicant may believe this would cause a lowered backpressure on SW header whichwould translate into a higher flow but at a lower pressure(Centrifugal pump curves). Second part is incorrect (See A.2). Since SW discharge pressure isvirtually unaffected then these valves will NOT fully open becausedischarge pressure is less than 110 psig. Plausible if the applicantbelieves that the system "DILUTION BYPASS" valves open toensure a minimum dilution flow is maintained for Radioactivereleases. .

D. Incorrect. First part is incorrect (See C.1). Second part is correct. These valves open on a low level in theSW pond but will not lower pressure.

K/A: 062AA1.07 Loss of Nuclear Service Water - Ability to operate and / ormonitor the following as they apply to the Loss of NuclearService Water (SWS): Flow rates to the components andsystems that are serviced by the SWS; interactions among

the componentsImportance Rating: 2.9 3.0Technical

Reference:

FNP-2-SOP-16.1, SG Blowdown Processing System, Ver 43.2 FNP-0-AOP-31, Loss of Service water Pond, Ver 12 FNP-2-SOP-24, Service Water System, Ver 73 FSD-A181001, Service Water System, Ver 61 D-200013, Sh 8, Service Water System, Ver 36References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associated with the safe operation of the Service Water Systemcomponents and equipment, to include the following(OPS-40101B07): [...]

• Automatic actuation including setpoint (example SI, PhaseA, LOSP)

• Protective isolations such as high flow, low pressure, lowlevel including setpoint

• Protective interlocks [...]Question History: FNP 12. The bank was checked and this question is theonly one that meets this K/A. We have spent hoursdeveloping this question and have encountered difficulty inthe final product due to our system design. K/A match: Requires the applicant to know what they are expecting to see (monitor) on the MCB (PI-3001 and FR4107)

andwhat will occur due to the flow to other system components (interactions among the components).

Thecandidate will have to know what happens to the SW system on low pond level (loss of SW) and then the effectsof the new valve line up on system pressure and flow to other system components (ie. RCV-18 and 23B and PCV-562 and MOV-538).SRO justification: N/A

44. 062K3.01 044The following conditions exist on Unit 2:

• DG02-2, 2G 4160 V bus tie to 2L 4160 V bus, has tripped opened. Which one of the following completes the statement below?The has lost Service Water cooling.2C Instrument Air Compressor2C Reactor Coolant Pump Motor Air Cooler2C Component Cooling Water Heat ExchangerSteam Generator Blowdown Heat Exchanger A.B.C.D.

AOP-10:Step 15. Minimize SW loads in affected train.15.2 For 'A' train affected minimize 'A' TRAIN SW LOADS as required.

15.2.1 Secure SGBD using FNP-2-SOP-16.1, STEAM GENERATOR BLOWDOWNPROCESSING SYSTEM.15.2.2 Close SW to blowdown and BTRS heat exchangers valve. SW TO BLDN HX & BTRS CHLRS[] Q2P16MOV3149 - closed15.3 For 'B' train affected minimize B TRAIN SW LOADS, as required.

15.3.1 Close SW to RCP motor air coolers.SW TO RCP MTR AIR CLRS

[] Q2P16MOV3135 - closedU2 Load List

2L 4160V bus is the power supply to the B Train SW pumps. When that power supplyis lost, All B Train SW pumps will be lost and cooling to B Train components areaffected. Distracter analysisA. Incorrect. See B. Plausible since the 2C designation could make theapplicant believe this is a 'B' train component. All instrument air compressors are normally supplied from a common SW header,which is fed from both trains of SW.B. Correct. ALL RCP motor air coolers are supplied from "B" Train SW.C. Incorrect. See B. 2C CCW Heat Exchanger is supplied from "A" Train SW. Plausible since 2C is an A Train component and 2A is B Traincomponent and this is a common mistake made for these components.D. Incorrect. See B. The SGBD Heat Exchanger is supplied only from "A" TrainSW. Plausible since this and the RCPs each are supplied fromdifferent trains and a common mistake made by students as towhich train supplies which components.

K/A: 062K3.01 A.C. Electrical Distribution - Knowledge of the effect that aloss or malfunction of the ac distribution system will have onthe following: Major system loadsImportance Rating: 3.5 3.9 Technical

Reference:

FNP-2-AOP-10, Loss of Service Water, Ver 18 A-351199, Unit 2 Load List, Ver 61.References provided: NoneLearning Objective:

RELATE AND DESCRIBE the effect(s) on the ServiceWater System for a loss of an AC or DC bus, or a malfunction of the Instrument Air System (OPS-40101B06).Question History: FNP EXAM BANK K/A match:

The 2L 4160V bus has been lost due to a malfunction and the effect is the loss of cooling to various majorsystems loads. The applicant will have to know whichSW pumps have lost power and then equate that towhich major system load has lost cooling. SRO justification: N/A

45. 063A1.01 045Unit 1 has experienced a Reactor trip with the following conditions:* A Loss of All AC has occurred.* ECP-0.0, Loss of All AC Power, is in progress.Which one of the following completes the statements below?The 1B Aux Building DC bus voltage will (1) . Per ECP-0.0, there may not be enough DC capacity to start a DG and sequenceneeded loads if power is not restored to the 125V DC battery chargers on each trainwithin a MINIMUM of (2) .1) drop slowly at first; then later drop rapidly as the battery nears exhaustion2) 30 min1) drop slowly at first; then later drop rapidly as the battery nears exhaustion2) 90 min1) drop at a constant, linear rate the entire time the battery discharges2) 30 min1) drop at a constant, linear rate the entire time the battery discharges2) 90 min A.B.C.D.

DOE Fundamentals Handbook Vol 2 of 4, Jun 1992 (This is a reference for lesson planOPS-30501D, Batteries.) -

During Battery discharge, voltage will slowly drop until thebattery approaches exhaustion. As the battery approaches exhaustion, voltage willdecrease exponentially until exhaustion. ECP-0.0 Caution prior to Step 5:IF power is not restored to the 125 V DC battery chargers on each train within 30minutes, THEN there may not be enough DC capacity to start a DG and sequenceneeded loads.Distracter analysis:A. Correct. First part is correct. The battery voltage will drop slowly then at anexponential rate towards the end of discharge per the graph in thereferences. Second part is correct. Per the Note, 30 minutes is the minimumtime in which the battery charger must be restored to ensure the DG can start and sequences loads.B. Incorrect. First part is correct (See A.1). Second part is incorrect (See A.1). Plausible since the designcapacity of the Aux building battery is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. 90 minutes wouldgive a 30 minute buffer so the applicant could confuse the 30 minutes in the note with "30 minutes left" of the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> designbattery capacity.C. Incorrect. First part is incorrect (See A.1). Plausible if the applicant is notfamiliar with battery discharge characteristics. Second part is correct (See A.2).D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2).

K/A: 063A1.01 D.C. Electrical Distribution - Ability to predict and/or monitorchanges in parameters associated with operating the DCelectrical system controls including: Battery capacity as it isaffected by discharge rateImportance Rating: 2.5 3.3 Technical

Reference:

FNP-1-ECP-0.0, Loss of All AC Power, Ver 26. DOE Fundamentals Handbook Vol 2 of 4, Jun 1992References provided: None Learning Objective:

STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with (1) ECP-0.0, Loss of All AC Power;[...] P-0.2, Loss of All AC Power Recovery, With SIRequired. (OPS-52532A03)Question History: MOD FNP11/12 NRC examK/A match: Requires the applicant to predict the change in batteryvoltage (parameters) as the battery is discharged(capacity versus discharge rate) during a Loss of All ACevent. In addition, knowledge of the time expected to restorethe battery charger (30 mins) to prevent the LOSS of DCPOWER, which equates to the time limit (how long can weoperate this way) that the battery capacity is affected.SRO justification: N/A

46. 064K3.02 046A loss of all AC power has occurred on Unit 1 and the following conditions exist:

• VA2, 1B DG GEN FAULT TRIP, has come into alarm.

• The crew has completed the step in ECP-0.0, Loss Of All ACPower, to verify breakers for major loads OPEN.

• A Safety Injection occurs on Unit 1 at this time. Which one of the following completes the statements below?The 2C DG will be started from the EPB in (1) using the START pushbutton. All ESF loads will (2) .1) Mode 22) automatically start1) Mode 22) have to be manually aligned1) Mode 12) automatically start1) Mode 1

2) have to be manually aligned A.B.C.D.

ECP-0.05.2.1 RNO Perform 2C DG SBO start as follows.5.2.1.1 RNO Verify 2C DG MODE SELECTOR switch in MODE 1.Note before Step 5.2.1.5 - NOTE: The LOSP sequencer should run when output breaker closes, if no SI signal ispresent. If an SI signal is present, neither sequencer will run and SI loads must be started manually.Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible since all other DGswould be started in Mode 2 in ECP-0.0. Second part is incorrect (See D.1). Plausible since the ESFsequencer would run if it were the 1-2A or 1B DG that was started.

The operation of the 2C DG in this scenario is complicated andeasily confused.B. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2)

C. Incorrect. First part is correct (See D.1) Second part incorrect (See A.2)D. Correct. First part is correct. Step 5.2.1.1 RNO of ECP-0.0 starts the 2C DGin Mode 1. Second part is correct. The note before step 5.2.1.5 RNO ofECP-0.0 states that under the conditions in the stem, the SIsequencer will NOT run and ESF loads must be manually aligned.

K/A: 064K3.02 Emergency Diesel Generators (ED/G) - Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following ESFAS controlled or actuated systems.Importance Rating: 4.2 4.4 Technical

Reference:

FNP-1-ECP-0.0, Loss of All AC Power, Ver 26References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing (1)ECP-0.0, Loss of All AC Power; [...] (OPS-52532A06)

ANALYZE plant conditions and DETERMINE if actuation orreset of any Engineered Safety Features Actuation Signal(ESFAS) is necessary. (OPS-52532A05)Question History: FNP 07 K/A match: This requires the applicant to know what effect a 1B DGmalfunction has on the ESFAS system in that ESF loads must be manually aligned.SRO justification: N/A

47. 064K6.07 047Unit 1 is operating at 100% power with the following conditions:

• A problem with 1B DG starting air system has occurred.

• The B Air receiver has been tagged out.Which one of the following completes the statement below?A MINIMUM of (1) psig must be available in the remaining air receiver to ensure five (5) start attempts are available.

1B DG's required minimum time to reach rated speed and voltage is (2) secondsafter receiving an emergency start signal. (1) (2) 200 7 200 12 350 7 350 12 A.B.C.D.

FSD - A181005:

2.1.2 - The DGS shall be capable of achieving > 3952 V and > 57 Hz within 12seconds after receipt of an engine start signalTech Specs Bases: 3.8.3

- With both starting air receiver pressures on a DG < 350psig for the 4075 kW DGs or < 200 psig for DG 1C, sufficient capacity for fivesuccessive DG start attempts does not exist.ANSWER / DISTRACTOR ANALYSISA. Incorrect. First part is incorrect (See D.1). Plausible if candidate thinks thatthe 1B DG is a "little DG" (Fairbanks Morse) instead of a "Big DG" (Colt Peilstick) which would make this a correct answer. Second part is incorrect (See D.2). Plausible if the candidateconfuses the required time to reach 115 RPM for the Fail to StartDG trip with the time to achieve rated voltage and speed. 7seconds in not a subset of 12 seconds. 12 seconds is therequirement which implies a maximum and any time > 7 seconds

but < 12 seconds would be acceptable also.B. Incorrect. First part is incorrect (A.1). Second part is correct (See D.2).

C. Incorrect. First part is correct (See D.1). Second part is incorrect (See A.2).D. Correct. First part is correct. The 1B DG is required to have one air receiver>350 psig to have 5 start attempts available. Second part is correct. The DGS shall be capable of achieving >3952 V and > 57 Hz within 12 seconds after receipt of an enginestart signal K/A: 064K6.07 Emergency Diesel Generators (ED/G) - Knowledge of theeffect of a loss or malfunction of the following will have onthe ED/G system: Air receivers Importance Rating: 2.7 2.9 Technical

Reference:

FSD-A181005, Diesel Generator, Ver 44 FNP Tech Specs Bases, Amendment No. 58References provided: NoneLearning Objective:

RECALL AND APPLY the LCO and APPLICABILITY forTechnical Specifications (TS) or TRM requirements, and the REQUIRED ACTIONS for 1 HR or less TS or TRMrequirements, and the relevant portions of BASES thatDEFINE the OPERABILITY and APPLICABILITY of the LCO associated with the Diesel Generator and Auxiliaries System components and attendant equipment alignment, toinclude the following (OPS-52102I01): [...] 3.8.3, Diesel Fuel Oil, Lube Oil, Starting AirQuestion History: NEW K/A match: Requires the applicant to know how a malfunction of theDG air start receivers affect the operation of the ED/Gsystem.SRO justification: N/A

48. 065AG2.4.11 048Unit 1 was operating at 100% power when the following occurred:

• A complete loss of instrument air caused an automatic Reactor Trip.The following conditions exist:

• All AFW pumps are running.

• All SG NR Levels are 25% and rising.

• The Shift Supervisor has directed AFW flow to be reduced.

Per AOP-6.0, Loss of Instrument Air, which one of the following methods belowwill be successful in reducing AFW flow? Valve nomenclature:

• HV-3228A / B / C, TDAFWP TO 1A/1B/1C SG

• MOV-3764A / D / F, MDAFWP TO 1A/1B/1C SG ISO

• MOV-3350A / B / C, AFW TO 1A/1B/1C SG STOP VLVPlace BOTH MDAFW pump MCB hand switches in the STOP position andrelease them.Throttle HV-3228A / B / C on the MCB.Close MOV-3764A / D / F on the BOP.Close MOV-3350A / B / C on the MCB.

A.B.C.D.

AOP-6Step 8. Maintain SG narrow range levels between 35-69%.8.1 RNO WHEN required to limit SG level rise,THEN perform the following:.a) Alternately cycle closed and open oneMDAFWP isolation valve to each SG.[ ] MDAFWP TO 1A SG ISO,Q1N23MOV3764A(E)

[ ] MDAFWP TO 1B SG ISO, Q1N23MOV3764B(D)

[ ] MDAFWP TO 1C SG ISO, Q1N23MOV3764C(F)b) STOP/START MDAFWPs as required.[ ] 1A MDAFWP

[ ] 1B MDAFWPDistracter analysisA. Incorrect. See C. Plausible since this is a method per AOP-6 step 8.1 RNObut since SG NR Level is less than 28%, the MDAFW pumpscannot be stopped due to the auto-start signal. The applicant maynot recall the MDAFW pump start logic and believe the pumps can be stopped.B. Incorrect. See C. Plausible if the applicant believes that the air receiver thatkeeps the TDAFW pump steam admission valves open is alsoused to control the TDAFW pump FCVs.C. Correct. Of the available choices, this is the only method to control AFWflow per AOP-6. Step 8 of AOP-6 also directs the use of theMDAFW and TDAWF FCV's locally but these are not an availablechoice due to the loss of airD. Incorrect. See C. Plausible since these valves are not addressed in AOP-6.0and do not have power supplied during full power ops. Plausiblesince these valves are used in AOP-4.0 on loss of RCP flow to stop AFW flow and could be used to stop flow to all SGs if power was supplied to the MOV.

K/A: 065AG2.4.11 Loss of Instrument Air - Knowledge of abnormal conditionprocedures.Importance Rating: 4.0 4.2 Technical

Reference:

FNP-1-AOP-6.0, Loss Of Instrument Air, Ver 40 References provided: NoneLearning Objective:

ANALYZE plant conditions and DETERMINE the successfulcompletion of any step in AOP-6.0, Loss of Instrument Air.(OPS-52520F07) Question History: FNP EXAM BANK K/A match: Applicant must know what equipment is directed to be usedby AOP-6 to control the cooldown rate.SRO justification: N/A

49. 068AK2.07 049There is a fire in the Control Room and the following conditions exist for Unit 1:

• FNP-1-AOP-28.2, Fire In The Control Room, has been entered.Which one of the following completes the statements below?During the conduct of AOP-28.2, the Diesel Generators are required tobe placed in (1) and the output breakers (2) automatically closewhen the DGs are started after a Loss of Offsite Power.

(1)

(2) MODE 3 WILL MODE 3 will NOT MODE 4 WILL MODE 4 will NOT A.B.C.D.

AOP-28.2Step 6.3.1 -

Dispatch personnel to the diesel building to perform ATTACHMENT 18,PLACING DIESEL GENERATORS IN LOCAL CONTROL.Attachment, 18 Step 1 - PLACE 1B DIESEL IN MODE 4.Attachment 21, Note prior to Step 1.10 - Diesel generator and diesel generator outputbreaker must be controlled locally (155' DG BLDG) while diesel generator is in MODE 4.Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible if the applicant assumesthat the most reliable condition of the DGs in this scenario wouldbe in Mode 3 as in AOP-49.2, Complete Loss of Service Water. Second part is incorrect (See C.2). Plausible because this wouldbe the correct answer if the DG was required to be in Mode 1 perAOP-28.2. Once the DG's are no longer in MODE 1 (MODE 2, 3 or 4), applicants can have difficulty recalling how the output breakerresponds on a DG start for an LOSP) B. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2). This is a logical connection to thefirst part if the applicant improperly believes that control power tothe DG output breakers is removed as part of shifting local controlof the Main Control Room operated equipment to the HSDP. TheRCP breakers are tripped locally and have control power removedper AOP-28.2.C. Incorrect. First part is correct (See D.1).

Second part is incorrect (See C.2). Plausible if the applicant failedto recall that the DG output breaker will NOT automatically closewhen started in Mode 4 after a loss of power.

D. Correct. First part is correct. Step 6.3.1 requires the DGs to be placed inMODE 4. Second part is correct. While in MODE 4, the DG output breakerwill NOT automatically close.

K/A: 068AK2.07 Control Room Evacuation - Knowledge of the interrelationsbetween the Control Room Evacuation and the following:

ED/G.Importance Rating: 3.3 3.4 Technical

Reference:

FNP-1-AOP-28.2, Fire In The Control Room, Ver 28References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performingAOP-28.1, Fire or Inadvertent Fire Protection System Actuation in the Cable Spreading Room, and AOP-28.2, Firein the Control Room. (OPS-52521C06)Question History: NEW K/A match: Requires the applicant to know the interrelation of the DGmode of operation to a control room evacuation.SRO justification: N/A

50. 071K3.05 050Unit 1 is operating at 100% power with the following conditions:

• The 1A Waste Gas Compressor is running and aligned to #7 WGDT.

• R-13, WGC SUCT, alarms.Subsequently, the #7 WGDT relief valve lifts and fails to reseat.Which one of the following completes the statements below?R-22, VENT STACK GAS, (1) trend up.#7 WGDT relief valve (2) be manually isolated.

(1)

(2) WILL CANNOT WILL CAN will NOT CANNOT will NOT CAN A.B.C.D.

D-175045 SH 1: Shows R-22 located in the vent stack.D175042 SH 6, Shows that the #7 WGDT relief discharges to the vent stack and hasno manual isolations.

Distracter analysisA. Correct. First part is correct. #7 WGDT relief valve discharges to the ventstack and would cause R-22 to trend up since the 1A Waste GasCompressor is aligned to it and the compressor suction has a highrad alarm. Second part is correct. There is no manual isolations for the #7WGDT relief valve. B. Incorrect. First part is correct (See A.1). Second part is incorrect (See A.2). Plausible since some systemshave isolation valves upstream of their relief valves such as LPFeedwater heaters (See 170116 SH 1 in reference material).C. Incorrect. First part is incorrect (See B.1). Plausible because WGDT 1through 6 relieve to #8 WGDT and the applicant could think that #7also relieved to #8 WGDT. Second part is correct (See A.2).

D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2).

K/A: 071K3.05 Waste Gas Disposal System (WGDS) - Knowledge of theeffect that a loss or malfunction of the Waste Gas DisposalSystem will have on the following: ARM and PRM systemsImportance Rating: 3.2 3.2 Technical

Reference:

D175045, Unit 1 HVAC - P&ID SFP Vent Sys, Sheet 1, Ver 22.0 D175042, Unit1 Waste Processing System, Sheet 6, Ver 33.0 References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associatedwith the safe operation of the Radiation Monitoring System components and equipment, to include the following(OPS-40305A07): [...]

• Automatic actuation

• Protective isolations

• Protective interlocks

• Actions needed to mitigate the consequence of theabnormality Question History: NEW K/A match: Requires the applicant to know the effect on R-22 (PRMS)when WGDT #7 relieves to the vent stack (malfunction of the WG system resulting in relief lifting).SRO justification: N/A

51. 073A2.02 051Unit 1 is operating at 100% when the following condition occurs:

• R-19, SGBD SAMPLE, fails HIGH.Which one of the following completes the statements below?

(1) valves will automatically close.Per SOP-45.0, Radiation Monitoring System, the actions required to allow the Shift Chemist to obtain a sample of the SGs is to (2) .1)

HV-3328, HV-3329 AND HV-3330, STEAM GEN 1A/1B/1C SAMPLE ISO,2) pull the INSTRUMENT power fuses for R-19 1) HV-3328, HV-3329 AND HV-3330, STEAM GEN 1A/1B/1C SAMPLE ISO,2) place R-19 Operations Selector Switch to the RESET position1) HV-3179A, 3180A, AND 3181A, STEAM GEN 1A/1B/1C LOWERBLOWDOWN, 2) pull the INSTRUMENT power fuses for R-19 1) HV-3179A, 3180A, AND 3181A, STEAM GEN 1A/1B/1C LOWERBLOWDOWN,2) place R-19 Operations Selector Switch to the RESET position A.B.C.D.

ARP-1.6, FH1 - R-19 isolates HV-3328, 3329 and 3330.SOP-454.4 Obtaining a Steam Generator Sample with R-19 in Alarm or Inoperable: 4.4.1 Notify Health Physics and Chemistry that R-19 will be inoperable during the timerequired to obtain a sample 4.4.2 IF in alarm, THEN place the switch for R-19 to the Reset position. 4.4.3 Open the Steam Generator Blowdown sample valves listed below as necessaryto obtain a Steam Generator sample:

Q1P15HV3328 1A Steam Generator Blowdown sample valve Q1P15HV3329 1B Steam Generator Blowdown sample valve Q1P15HV3330 1C Steam Generator Blowdown sample valve Distracter analysisA. Incorrect. First part is correct (See B.1). Second part is incorrect (See B.2). Plausible since this is theprocedure directed action for a monitor in saturation, but not toallow the chemist to sample the SG.

B. Correct. First part is correct. R-19 failing in the "High Radiation" conditionshuts HV-3328, 3329, and 3330. Second part is correct. Per SOP-45, the Rad monitor switch mustbe taken to reset to allow SGBD sample valves to be opened.C. Incorrect. First part is incorrect (See B.1). Plausible if the applicant doesn'trecall that R-19 will isolate HV-3328, 3329, and 3330. They maybelieve that R-19 closes HV-3179A, 3180A, and 3181A which areimmediately upstream of the correct valves and closed by the AFWpump start signal and on High Penetration Room DIFFERENTIAL

Pressure.

Second part is incorrect (See A.2). D. Incorrect. First part is incorrect (See C.1). Second part is correct (See B.2).

NOUN NAME for HV-3179A, 3180A, AND 3181A, 1A/1B/1C SG LOWERBLOWDOWN SAMPLE ISO came from ARP BK1.

K/A: 073A2.02

Process Radiation Monitoring (PRM) System - Ability to (a)predict the impacts of the following malfunctions oroperations on the PRM system; and (b) based on thosepredictions, use procedures to correct, control, or mitigatethe consequences of those malfunctions or operations:Detector failureImportance Rating: 2.7 3.2 Technical

Reference:

FNP-1-ARP-1.6, FH1, RMS HI RAD, Ver 70 FNP-1-SOP-45, Radiation Monitoring System, Ver 46.2References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associated with the safe operation of the Radiation Monitoring System components and equipment, to include the following(OPS-40305A07): [...]

• Automatic actuation

• Protective isolations

• Protective interlocks

• Actions needed to mitigate the consequence of theabnormality Question History: MOD FNP 07 K/A match: The applicant is required to know the impact on the SGsample system due to R-19 failing high (Desktopsimulator shows a level amp failure - high will alarmR-19) and that SOP-45 provides procedural guidance tosample the SGs under this condition.SRO justification: N/A

52. 073G2.2.42 052Unit 1 is operating at 100% power.Which one of the following meets the MINIMUM reactor coolant leakage detectionsystem(s) that must be in operation and OPERABLE to prevent entering a REQUIREDACTION STATEMENT of Tech Spec 3.4.15, RCS Leakage Detection Instrumentation?

• R CTMT PARTICULATE

• R CTMT GAS

• Containment Air Cooler Condensate Level Monitoring System (CACCLMS)R-11 ONLY ()R-11 AND R-12The CACCLMS ONLYR-12 AND the CACCLMS A.B.C.D.

Technical Specifications:3.4 REACTOR COOLANT SYSTEM (RCS)3.4.15 RCS Leakage Detection InstrumentationLCO 3.4.15 The following RCS leakage detection instrumentation shall beOPERABLE:a. One containment atmosphere particulate radioactivity monitor; andb. One containment air cooler condensate level monitor or one containmentatmosphere gaseous radioactivity monitor.APPLICABILITY: MODES 1, 2, 3, and 4.SOP-45R-11 is the containment atmosphere particulate radioactivity monitor and R-12 is the containment atmosphere gaseous radioactivity monitor.Distracter analysisA. Incorrect. See B. Plausible if the applicant recalls that RE-67 grab samplesare required by this tech spec when R-12 and the CACCLMS areOOS but does NOT to prevent entry into a Required Action Statement (RAS). The applicant may believe that RE-67 is anacceptable substitute for R-12 to prevent LCO entry. B. Correct. Per TS - 3.4.15 this is the correct combination. C. Incorrect. See B. Plausible if the applicant thought that grab samples fromRE-67 could provide the R-11 function. Grab samples are part ofthis Tech Spec.D. Incorrect. See B. Plausible if the applicant confused R-12 with R-11. This isa common misconception.

K/A: 073G2.2.42 Process Radiation Monitoring (PRM) System - Ability torecognize system parameters that are entry-level conditionsfor Technical Specifications.Importance Rating: 3.9 4.6 Technical

Reference:

Technical Specifications, Ver 190. FNP-1-SOP-45, Radiation Monitoring System, Ver 46.2References provided: NoneLearning Objective:

Given a set of Plant Conditions ASSESS those conditionsand DETERMINE the ability of plant equipment and structures to meet their intended, designated function(OPS-52302A06)

Question History: FNP 05 K/A match: Requires the applicant to know which RCS leakagedetection systems (Process Radiation Monitors) that arerequired to meet Technical Specifications.SRO justification: N/A

53. 076AA1.04 053Unit 1 has been operating at 100% power and the Gross Failed Fuel Detector (GFFD)has been steady at 2000 cpm during the entire fuel cycle.At 1000:

• FG5, GFFD SYS TRBL, has just come into alarm.At 1015:

• A Reactor Trip and Safety Injection occurs.Which one of the following completes the statements below?The minimum GFFD reading that would cause FG5 to come intoalarm is (1) above background.At 1020, flow through the GFFD (2) be isolated. (1) (2) 1 X 10 4 cpm will NOT 1 X 10 4 cpm WILL 1 X 10 5 cpm will NOT 1 X 10 5 cpm WILL A.B.C.D.

FG5 setpoint 1 x 10 4 cpm ABOVE background.D175009- Sheet 1 -

SV-3333 and SV-3765 close on a T signal (Phase A) which willactuate on a Safety Injection.Distracter analysisA. Incorrect. First part is correct (See B.1) Second part is incorrect (See B.2). Plausible since the RHR toGFFD detector valves do not close on a T signal. In Mode 1, theGFFD will be lined up the RCS not RHR.B. Correct. First part is correct. 1 X 10 4 cpm > background will cause thealarm. Second part is correct. On a safety injection, a Phase A isgenerated isolating the GFFD from the RCS. In Mode 1, the GFFDis aligned to the RCS.C. Incorrect. First part is incorrect (See B.1). Plausible since this is the setpointin AOP-32 to reduce power by 25%. The applicant could confusethe two numbers. Second part is incorrect (See A.2).

D. Incorrect. First part is incorrect (See C.1). Second part is correct (See B.2).

K/A: 076AA1.04 High Reactor Coolant Activity -

Ability to operate and / ormonitor the following as they apply to the High ReactorCoolant Activity: Failed fuel-monitoring equipment.Importance Rating: 3.2 3.4 Technical

Reference:

FNP-1-ARP-1.6, FG5 GFFD SYS TRBL, Ver 70 D175009, SH 1, Sampling System, Ver 32 References provided: NoneLearning Objective: RELATE AND IDENTIFY

the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the GrossFailed Fuel Detector, to include the components found onFigure 2, GFFD Failed Fuel Detector System, and Figure 3, Sampling Assembly Flow Diagram (OPS-52106E02).Question History: FNP 08 K/A match: Requires the applicant to be able to monitor the failed fuelmonitoring equipment

and determine the minimum levelat which the GFFD system trouble alarm actuates whichdirects the operators to AOP-32, Reactor Coolant HighActivity.SRO justification: N/A

54. 076K2.08 054Unit 1 is stable in Mode 3 following a Reactor Trip when the following conditions occur:

• Power has been lost to 4160V AC buses G, J, and L. Which one of the following lists the valves that the OATC can close from the MCB toisolate a Service Water rupture in the Turbine Building? Q1P16V514, SW TO TURB BLDG ISO B TRN Q1P16V515, SW TO TURB BLDG ISO A TRN Q1P16V516, SW TO TURB BLDG ISO A TRN Q1P16V517, SW TO TURB BLDG ISO B TRNMOVs 514 and 517MOVs 514 and 516 MOVs 515 and 517MOVs 515 and 516 A.B.C.D.Electrically:1N MCC (A Train) - V515 and V5171T MCC (B Train) - V514 and V516Mechanically:A Train - V515 and V516B Train - V514 and V517Distracter analysisA. Incorrect. See B. Plausible since the 4 SW to Turbine Building Isolationvalves are powered from and mechanically aligned to differenttrains and the applicant could easily confuse which valve is powered by which train and which valve is in which mechanical train.B. Incorrect. See A.C. Correct. Both of these valves are powered from A train power and inopposite trains mechanically (See Above). D. Incorrect. See A.

K/A: 076K2.08 Service Water System (SWS) -

Knowledge of bus powersupplies to the following: ESF-actuated MOVsImportance Rating: 3.1* 3.1*Technical

Reference:

A506250, Unit 1 Electrical Load List, Ver 74.0 D-170119, SH 2, Service Water, Ver 47References provided: NoneLearning Objective:

NAME AND IDENTIFY the Bus power supplies, for thoseelectrical components associated with the Service WaterSystem, to include those items in Table 7- Power Supplies(OPS-40101B04). Question History: FNP 08 K/A match: Applicant is required to know the bus power supplies toService Water ESF actuated MOVs.SRO justification: N/A

55. 076K4.03 055Unit 1 is operating at 100% power with the following conditions:

• 1A Containment Cooler is isolated per SOP-12.1, Containment Air CoolingSystem.

• The following valves are closed with power available: - MOV-3019A, SW TO 1A CTMT CLR AND CTMT FPS - MOV-3441A, SW FROM 1A CTMT CLR

- MOV-3024A, EMERG SW FROM 1A CTMT CLR

• MOV-3023A, 1A CTMT CLR SW DISCH, is OPEN.Subsequently, a steam break occurs and containment pressure rises to 5 psig.Which one of the following completes the statement below?1A Containment Cooler service water flow will be

.0 gpmapproximately 600 gpmapproximately 800 gpmapproximately 2000 gpm A.B.C.D.

FSD-A-181013:Post-accident, the containment coolers provide for long-termcontainment heat removal.

Following a safety injection signal and depending uponthe availability of offsite power, the containment coolers are restarted on low speed (Aloss of off site power (LOSP) would result in one fan from each train being started).Each cooler is nominally operated at a low speed generating 40,000 cfm with a servicewater flow rate of approximately 2000 gpm. During post-accident operation, eachcooler provides approximately 80 x 106 Btu/hr of cooling capacityDistracter analysisA. Incorrect. See D. Plausible since the applicant may know that an MOV in theservice water supply/return to each cooler will not open on an SI(MOV-3023A) and improperly think that it is MOV-3441A and thereforethere would be no flow. B. Incorrect. See D. Plausible since this is the minimum design flow per techspec bases. The applicant may not be able to recall the proper SWflow. C. Incorrect. See D. Plausible since this is the normal flow through the 1Acontainment cooler. The applicant may not be able to recall theproper SW flowD. Correct. This is the post accident flow through the 1A containment cooler.

K/A: 076K4.03 Service Water System (SWS) - Knowledge of SWS designfeature(s) and/or interlock(s) which provide for the following:Automatic opening features associated with SWSisolation valves to CCW heat exchangersImportance Rating: 2.9* 3.4*Technical

Reference:

FSD-A181013, Containment Ventilation System, Ver 14 References provided: NoneLearning Objective: DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associatedwith the safe operation of the Service Water System components and equipment, to include the following(OPS-40101B07): [...]

Automatic actuation including setpoint (example SI, PhaseA, LOSP)

Question History: FNP 07 K/A match: Requires the applicant to know the Containment CoolerIsolation valves open by design on a safety injectionand what the resultant SW flow to the coolers for

accident conditions will be.

10/24/12 - Per discussion withChief Examiner, using SWS Turbine Building Isolationclosure was acceptable due to FNP plant design.

Since theSW to TB MOVs were being addressed in a different KAand since this was an auto opening, Service Water to theContainment Coolers have automatic opening features andmore closely fit the K/A based on FNP design.SRO justification: N/A

56. 077AK3.02 056Unit 1 is at 100% power with the following conditions:

• 1B DG is running for STP-80.1, Diesel Generator 1B Operability Test.

• The 1B DG is currently loaded to 1 MW.

• DG01, 1B S/U XFMR TO 1G 4160 V BUS, breaker is CLOSED.Subsequently, the following occurs:

• WE2, 1F, 4KV BUS OV-OR-UV OR LOSS OF DC, and VE2, 1G, 4KV BUS OV-OR-UV OR LOSS OF DC, come into alarm.

• 1F and 1G 4160V bus voltages are reading 3825 volts.

• The crew has entered AOP-5.2, Degraded Grid.Per AOP-5.2, which one of the following completes the statements below?The 1B DG (1) .The reason for the above action is (2) .1) is required to be secured and aligned for AUTO START2) because this places the 1B DG is in the most reliable condition1) is required to be secured and aligned for AUTO START2) because the LOSP Sequencer will not run LOSP loads if DG01 opens with 1B DG output breaker aligned to 1G 4160V bus1) load is required to be raised to FULL LOAD and DG01 opened2) to ensure adequate voltage for safety related equipment1) load is required to be raised to FULL LOAD and DG01 opened2) to prevent extended low load operation which would result in the buildup of combustion products in the engine exhausts A.B.C.D.

Not a true 2+2 question for improved plausibility.AOP-5.2:CAUTION: Diesel generators are in the most reliable condition when secured andaligned for auto start. The intent of step 4 is to secure any diesel generators which arerunning and not required.4 Verify All Emergency Diesel Generators -ALIGNED FOR AUTO START using:FNP-0-SOP-38.0, DIESEL GENERATORSDistracter analysisA. Correct. First part is correct. Step 4 of AOP-5.2 requires any DG notrequired to be running to be secured and aligned for auto start. Second part is correct. Diesel generators are in the most reliablecondition when secured and aligned for auto start.B. Incorrect. First part is correct (See A.1). Second part is incorrect (See A.2). Plausible because this used tobe correct until recently a design change was implemented toinstall a Test Trip Override Switch. This modification causes theDG output breaker to trip open during testing if the normal supplybreaker to 1G 4160V bus opens. Once the DG output breaker tripsopen, the LOSP Sequencer will function properly to re-close the DG output breaker and sequence on LOSP loads. This modification makes this reason incorrect. C. Incorrect. First part is incorrect (See A.1). Plausible if the applicant thoughtthat since the grid is degraded, operation of the DG is required tocontinuously maintain power to the ESF busses.

Second part is incorrect (See A.2). Plausible since it would be ahigh priority to maintain adequate voltage to safety relatedequipment in order to prevent damage to the equipment due tohigh current.D. Incorrect. First part is incorrect (See C.1).

Second part is incorrect (See A.2). Plausible because the normaloperation procedure (FNP-0-SOP-38.0, 2.2.8 ) has a precautionthat the DGs should be loaded to full load for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> eachtime they are started to reduce the possibility of an exhaust fire.

However, under these conditions, the DGs are required to besecured and aligned for auto start.

K/A: 077AK3.02 Generator Voltage and Electric Grid Disturbances -Knowledge of the reasons for the following responses asthey apply to Generator Voltage and Electric GridDisturbances:

Actions contained in abnormal operatingprocedure for voltage and grid disturbances.Importance Rating: 3.6 3.9 Technical

Reference:

FNP-1-AOP-5.2, Degraded Grid, Ver 15 References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing [...] andAOP-5.2, Degraded Grid. (OPS-52521N06)Question History: VNP 10 K/A match: Applicant is required to know the actions contained inAOP-5.2 during a grid disturbance for DGs that arerunning and the reason for that action

.SRO justification: N/A

57. 078K1.01 057The following conditions exist on Unit 1:* A rupture in the Instrument Air system has occurred.* Instrument Air header pressure is 65 psig and lowering slowly.Which one of the following completes the statements below?V-902, AIR DRYER AUTO BYP, will be (1) . V-904, NON-ESSENTIAL IA HDR AUTO ISO, will be (2) . (1)

(2) OPEN OPEN OPEN CLOSED CLOSED OPEN CLOSED CLOSED A.B.C.D.KD2AUTOMATIC ACTION3. Pressure downstream of inst air dryers, bypasses dryers (V902) at 70 psig.4. Pressure downstream of inst air dryers, isolates inst air to service bldg(V904) at 55 psig.Distracter analysisA. Correct. First part is correct. V-902 opens at 70 psig. Second part is correct. V-904 closes at 55 psig.B. Incorrect First part is correct (See A.1). Second part is incorrect (See A.2). Plausible since there arenumerous setpoints for alarms and automatic valve repositioningsin the air system and they are easily confused.C. Incorrect. First part is incorrect (See A.1). Plausible since there arenumerous setpoints for alarms and automatic valve repositioningsin the air system and they are easily confused. Second part is correct (See A.2)

D. Incorrect. First part is incorrect (See C.1). Second part is incorrect (See B.2).

K/A: 078K1.01 Instrument Air System - Knowledge of the physicalconnections and/or cause-effect relationships between theIAS and the following systems:

Sensor airImportance Rating: 2.8* 2.7*

Technical

Reference:

FNP-1-ARP-1.10, KD2, IA PRESS LO Ver 70.2References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associatedwith the safe operation of the Compressed Air Systemcomponents and equipment, to include the following(OPS-40204D07): [...]Automatic actuation including setpoints for selectiveisolation on decreasing header pressure. [...]

Question History: FNP 04 K/A match: Requires the applicant to know the cause and effectrelationship between the sensed air header pressureand the automatic operation of isolation valve V-904 and bypass valve 902. SRO justification: N/A

58. 079G2.4.34 058The following plant conditions exist on Unit 1:

• The Reactor has been tripped due to loss of Instrument Air.

• The operating crew is performing the actions of ESP-0.1, Reactor Trip Response.

• SG Atmospheric Relief Valves (ARVs) are aligned per SOP-62.0,Emergency Air System.Subsequently, the operator applies 18 psig to the valve actuator for PCV-3371A,1A MS ATMOS REL VLV.Which one of the following completes the statements below?PCV-3371A (1) open.If PCV-3371A were fully open, (2) .1) IS2) a High Steam Flow - Lo Lo Tavg Main Steam Isolation may occur1) IS2) Technical Specification cooldown limits may be exceeded1) is NOT2) a High Steam Flow - Lo Lo Tavg Main Steam Isolation may occur1) is NOT

2) Technical Specification cooldown limits may be exceeded A.B.C.D.

This question is not a true 2 + 2 to improve plausibility of distracters.SOP-62.0Caution after step 4:Atmospheric relief valves will start to open at 24 +/- 2 psig and will be full open at 45 psig. IF the atmospheric relief is full open, THEN Tech Speccooldown limits may be exceeded.Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible if the applicant is notfamiliar with the actuation pressure of the ARV's. Second part is incorrect (See D.2). Plausible since a caution existsin numerous procedures that excessive opening of the STEAMDUMPS will cause this isolation of the MSIV's. Hi Steam Flow LoLo Tavg is 1 of 2 flow instruments on 2 of 3 steam lines.

This isolation closes the MSIV's and not the ARV's. The applicant could think that this isolation in fact does close the ARV's toprevent exceeding a technical specification cooldown.B. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2).

C. Incorrect. First part is correct (See D.1). Second part is incorrect (See A.2).D. Correct. First part is correct. The ARV will not open until at least 22 psig ofair is applied. Second part is correct. Per the caution of SOP-62, a fully openARV may cause tech spec limits to be exceeded.

K/A: 079G2.4.34

Station Air System - Knowledge of RO tasks performedoutside the main control room during an emergency and theresultant operational effects.Importance Rating: 4.2 4.1Technical

Reference:

FNP-1-SOP-62.0, Emergency Air System, Ver 23 References provided: NoneLearning Objective:

DEFINE AND EVALUATE the operational implications ofnormal / abnormal plant or equipment conditions associatedwith the safe operation of the Main and Reheat SteamSystem components and equipment, to include the following(OPS-40201A07): [-]

• Abnormal and Emergency Control Methods

• Automatic actuation including setpoint (example SI, PhaseA, Phase B, MSLIAS, LOSP, SG level)

• Protective isolations such as high flow, low pressure, lowlevel including setpoint [-]

Question History: MOD FNP 05 K/A match: Requires the applicant to know the operational effects oflocal operator actions to control the SG ARV's. Theoperator must use the Emergency Air system locally tocontrol ARV position due to the loss of the Station Airsystem.SRO justification: N/A

59. 103K4.06 059Unit 1 has experienced a large break LOCA and the following conditions exist:

• PT-953, CTMT PRESS has reached the PHASE B setpoint.Which one of the following completes the statement below?A MINIMUM of (1) OR (2) PHASE B CTMT ISO CS ACTUATIONhandswitch(es) is(are) required to actuate a Phase B isolation.1) 1 additional Containment pressure channel reaching 16.2 psig2) TWO1) 2 additional Containment pressure channels reaching 16.2 psig2) ONE1) 1 additional Containment pressure channel reaching 27 psig2) TWO1) 2 additional Containment pressure channels reaching 27 psig2) ONE A.B.C.D.

Not a true 2 + 2 for plausibility.FSD-A181007 2.7.1Phase B isolation is initiated by containment pressure High-3 (27 psig) on 2 of 4 b/s orby manual actuation ( using 2/4 Containment Phase B Isolation/Containment Spray Actuation handswitches).Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible since this is the correctnumber of channels but the wrong setpoint. 16.2 psig is HI-2 mainsteam line isolation and NOT HI-3 Phase B isolation. Second part is correct (SeeC.2). B. Incorrect. First part is incorrect (See C.1). Plausible since there are 4bistables and the applicant could reason that 3 of 4 are required toactuate Phase B. Additionally, the setpoint is incorrect asdiscussed in A.1. Second part is incorrect (See C.2). Plausible since Phase A and SIonly require 1 handswitch to actuate. The applicant may confusethem.C. Correct. First part is correct. Per the FSD, High-3 Containment IsolationPhase B coincidence is 2 of 4 bistables. Second part is correct. Per the FSD, 2 handswitches are requiredto actuate Phase B Containment Isolation. D. Incorrect. First part is incorrect (See C.1). Plausible since there are 4bistables and the applicant could reason that 3 of 4 are required toactuate Phase B. Permissives such as P-8 and P-9 require 2 of 4to enable and 3 of 4 to disable. There are many coincidences inthe reactor protection system and they are easily confused. Second part incorrect (See B.2)

K/A: 103K4.06 Containment System - Knowledge of containment systemdesign feature(s) and/or interlock(s) which provide for thefollowing: Containment isolation system Importance Rating: 3.1 3.7 Technical

Reference:

FSD-A181007, Reactor Protection System, Ver 18References provided: NONELearning Objective:

SELECT AND ASSESS the following instrument/equipmentresponse expected when performing Containment Structureand Isolation System evolutions including the fail condition,alarms, and trip setpoints (OPS-52102A05):

• PT-950, PT-953Question History: VOGTLE 02 - Changed to containment phase B isolation tomeet K/A vs ctmt spray actuation. K/A match: Requires the applicant to have knowledge of the

designfeature of the Phase B Containment Isolation System inthat 2 of 4 bistables or 2 of 2 handswitches are required foractuation.SRO justification: N/A

60. G2.1.18 060The electronic log is malfunctioning. The control room has shifted to manual logs andthe following entries have been made: 1000 Q1E21V061A, HHSI to 1C RCS loop CL iso, as left position; 1.5 turns OPEN. 1012 Started 1B CCW Pump.

At 1030:

• The OATC recognizes that an error was made on the 1000 log entry.

• Q1E21V061A should have been logged as throttled to 1.25

turns OPEN.Per SOP-0.11, Watch Station Tours and Operator Logs, the OATC is required tocorrect the 1000 log entry by which one of the following methods?* Circle the incorrect entry in red.

• Enter the correct information next to the incorrect information andrecord the date and initial.* Circle the incorrect entry in red.* At 1030 make a log entry with the correct information anddesignate it as a LATE ENTRY.* Draw a single line through the incorrect entry.

• Enter the correct information next to the incorrect information andrecord the date and initial.* Draw a single line through the incorrect entry.

• At 1030 make a log entry with the correct information and designateit as a LATE ENTRY.

A.B.C.D.

SOP- 0.11 Pg 13IF an error is made when recording hand written entries, THEN a single line willbe drawn through the incorrect entries AND the correct entries recorded. Theperson making the correction must initial AND date the change.Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible because NMP-OS)007-001, Conduct of Operations Standards andExpectations, Step 6.13.2.3 requires out of specification reading inmanual logs to be circled. The applicant could confuse theserequirements. Second part is correct (See C.2).B. Incorrect. First part is incorrect (See A.1).

Second part is incorrect (See C.2). Plausible since the correct datawould be placed in the log. The SOP-0.11 uses a LATE ENTRY toadd additional information to the log and NOT to correct errors.C. Correct. First Part is correct. IF an error is made when recording handwritten entries, THEN a single line will be drawn through the incorrect entries AND the correct entries recorded. The person making the correction must initial AND date the change. Second part is correct. The person making the correction mustinitial AND date the change.D. Incorrect. First Part is correct (See C.1). Second part is incorrect (See B.2).

K/A: G2.1.18 Ability to make accurate, clear, and concise logs, records,status boards, and reports.Importance Rating: 3.6 3.8 Technical

Reference:

FNP-0-SOP-0.11, Watch Station Tours and Operator Logs,Ver 26.4 NMP-OS-007-001, Conduct of Operations Standards andExpectations, Ver 13References provided: None Learning Objective:

Identify the required entries into the Plant Operator'sLogbook and the position with overall responsibility formaintenance of the Reactor Operator's Logbook. (OPS52303O01)Question History: NEW K/A match: Applicant is required to have the ability to correct logerrors to ensure the operator logs are accurate, clear,and concise.SRO justification: N/A

61. G2.1.5 061Both Units are operating at 100% power with the following conditions:

• A non-licensed Fire Protection Administrator who is qualified as a Shift Communicator is on shift.Which one of the following completes the statements below?Per EIP-0.0, Emergency Organization, a minimum of (1) licensed PlantOperators is required to staff the shift.The maximum number of hours that a Plant Operator may work inany 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is (2) per NMP-AD-016-003, Scheduling andCalculating Work Hours.

(1)

(2) 3 12 3 16 4 12 4 16 A.B.C.D.

EIP-0.0 Table 1 requires:1 OATC per Unit - Total of 21 UO Shared - Total of 1Shift Communicator (Least affected UO) - 1NMP-AD016-0036.1.1 The following work hour ceiling limits apply to covered individuals regardless ofunit status:

• No more than 16 work hours in any 24-hour period* No more than 26 work hours in any 48-hour period* No more than 72 work hours in any 7-day/168-hour periodDistracter analysisA. Incorrect. First part is correct (See B.2). Second part is incorrect (See B.2). Plausible since this is thenormal number of hours work and the applicant could not be ableto recall the correct limit.B. Correct. First part is correct. Per EIP-0.0, 3 Licensed operators are requiredto man the shift since a shift communicator is also on shift. Second part is correct. The following work hour ceiling limits applyto covered individuals regardless of unit status:

• No more than 16 work hours in any 24-hour periodC. Incorrect. First part is incorrect (See B.2). Plausible since without anon-licensed shift communicator, this would be a correct answer. Second part is incorrect (See A.2).D. Incorrect. First part is incorrect (See C.2).

Second part is correct (See B.2).

K/A: G2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.Importance Rating: 2.9* 3.9 Technical

Reference:

FNP-0-EIP-0.0, Emergency Organization, Ver 29 NMP-AD-016-003, Scheduling and Calculating Work Hours,Ver 5 References provided: NoneLearning Objective:

Given the plant mode for each unit, STATE AND EXPLAINthe minimum manning requirements for manning one orboth units (OPS40502H04).Question History: NEW K/A match: Requires the applicant to have the ability to determineminimum crew manning as well as maximum hours thatthe operator may work.

Since this question asks for thereactor operator position it is deemed to be an RO questionand since an RO objective exists for this knowledgerequirement.SRO justification: N/A

62. G2.2.20 062Unit 1 is operating at 100% power when the following occurs:

• STP-4.1, 1A Charging Pump Quarterly Inservice Test, is in progress.

• 1A Charging pump failed to start when the handswitch was taken to START. Which one of the following are the required actions per SOP-0.0, General Instructionsto Operations Personnel?Obtain Shift Manager's permission, THEN take the handswitch to START a second time.Take the handswitch to START a second time, THEN write a Condition Report todocument the action.Write a condition report documenting the event and contact Maintenance. The System Operator will rackout and perform a visual inspection of the circuitbreaker and write a condition report.

A.B.C.D.SOP-0.0 15.1.3. For handswitches on the MCB, EPB, BOP, and HSDP, if the associatedcomponent fails to actuate (pump-start, valve-move in open or closed direction, etcetera) when operating a handswitch, a second actuation may NOT be attempted untilthe cause can be thoroughly investigated

.15.2.4 IF a breaker has malfunctioned (i.e., failed to close, open, trip, or charge whenexpected) contact appropriate Maintenance personnel for involvement introubleshooting prior to attempting restoration efforts. [...]Distracter analysisA. Incorrect. See C. Plausible since this is correct when backing up ESFequipment actuation on the third, fourth attempt etc per step15.1.5of SOP-0.0.B. Incorrect. See C. Plausible since this is correct when backing up ESFequipment actuation per step 15.1.4 and 15.1.5 of SOP-0.0.C. Correct. A second attempt is not allowed under normal operatingconditions.D. Incorrect. See C. Per SOP-0.0, Step 15.2.4, racking out a circuit breaker thathas malfunctioned is NOT allowed. Plausible since all breaker malfunctions are investigated by the Systems Operators (without racking the breaker out) and the applicant could think that a visualinspection of a racked out breaker is appropriate before callingmaintenance.

K/A: G2.2.20 Knowledge of the process for managing troubleshootingactivities.Importance Rating: 2.6 3.8Technical

Reference:

FNP-0-SOP-0.0, General Instructions to OperationsPersonnel, Ver 152.2References provided: NoneLearning Objective:

Using plant procedures, describe the work control processand associated program interfaces, including ToolpouchWork (for example, tagging, radiation protection, foreign material exclusion, fire protection, and industrial safety).(OPS-40502N09).Question History: NEWK/A match: The applicant is required to know what actions arerequired to support troubleshooting activities for acircuit breaker that failed to close. The actions are thesame regardless of if the applicant assumes the breakerfailed to shut or the handswitch failed to actuate.SRO justification: N/A

63. G2.2.25 063Per Tech Specs Bases 3.3.1, Reactor Trip System (RTS) Instrumentation, what is thebasis of the Pressurizer Water Level - High Reactor trip?Protects the pressurizer safety valves against water relief.Provides the primary protection for preventing RCS over pressurization.Protects against loss of pressure control due to spray nozzle being submerged.Provides protection against exceeding containment design pressure in the event ofa LOCA.A.B.C.D.3.3.1 Bases

9. Pressurizer Water Level - High:The Pressurizer Water Level-High trip Function provides a backup signal for thePressurizer Pressure-High trip and also provides protection against water reliefthrough the pressurizer safety and power-operated relief valves (PORV). Thesevalves are designed to pass steam in order to achieve their design energy removalrate, but are also qualified for limited water relief following specific transients. A reactortrip (Pressurizer Pressure - High) is actuated prior to the pressurizer becoming watersolid.Distracter analysisA. Correct. Per Tech Specs bases 3.3.1. the Pressurizer Pressure-High tripand also provides protection against water relief through the pressurizer safety and power-operated relief valves (PORV).B. Incorrect. See B. Plausible because this is the bases for the High PressureReactor Trip. Przr High Level trip is the BACKUP for the HighPressure trip.C. Incorrect. See B. Plausible if the applicant believes that the spray nozzle willbe covered at 92% Pressurizer level. This is not correct. D. Incorrect. See B. Plausible if the applicant assumes that the additional waterin the RCS will cause a pressure rise that exceeds the containmentdesign pressure.

.

K/A: G2.2.25 Knowledge of the bases in Technical Specifications forlimiting conditions for operations and safety limits.Importance Rating: 3.2 4.2Technical

Reference:

Technical Specifications Bases, Ver 58 Core Operating Limit Report, Unit 1 Cycle 25, Ver 1References provided: NoneLearning Objective:

RECALL AND APPLY the LCO and APPLICABILITY forTechnical Specifications (TS) or TRM requirements, and theREQUIRED ACTIONS for 1 HR or less TS or TRMrequirements, and the relevant portions of BASES thatDEFINE the OPERABILITY and APPLICABILITY of theLCO associated with the Reactor Protection System (RPS) ,to include the following (OPS-52201I10): 3.3.1 Reactor Trip System (RTS) Instrumentation Question History: FNP EXAM BANK K/A match: Requires the applicant to know the bases for Tech Spec3.3.1, Reactor Trip System (RTS) Instrumentation -Pressurizer Water Level - High.SRO justification: N/A

64. G2.3.12 064Unit 1 is in Mode 6 for a refueling outage.

• Two Plant Operators are required to enter a room that is posted as a Locked High Radiation Area (LHRA) to perform work.Which one of the following completes the statements below? The radiation level at which this posting is required is (1) . The LHRA key is obtained from (2) . 1) > 100 mrem/hr2) Health Physics Supervision1) > 100 mrem/hr2) the Shift Support Supervisor (SSS)1) > 1000 mrem/hr2) Health Physics Supervision1) > 1000 mrem/hr

2) the Shift Support Supervisor (SSS)

A.B.C.D.

RCP-0 5.2.5.1 A LHRA means an area, accessible to individuals, in which radiation levels fromradiation sources external to the body could result in an individual receiving a doseequivalent in excess of 1 Rem/hr at 30 centimeters (11.81 inches or ~ 12 inches) fromthe radiation source or 30 centimeters from any surface that the radiation penetrates.RCP-0.1 APP A:2.1 Maintain Individual Locked High Radiation Area keys under the control of HPSupervision.Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible because this is the limitfor a high radiation area and the applicant could confuse the twolimits. Second part is correct (See C.2).B. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See C.2). Plausible since a LHRA MasterKey is locked within a key storage cabinet located in the ControlRoom. The key is available for issue by the OPS Shift Supervisorto support mitigation activities associated with an NMP-EP-110Emergency. Additionally, the SSS issues numerous keys to personnel duringplant operation and the applicant could assume this is one of them.C. Correct. First Part is correct. 1000 mrem/hr is a Locked HRA. Second part is correct. Individual Locked High Radiation Area keysare maintained under the control of HP Supervision.D. Incorrect. First part is correct (See C.1). Second part is incorrect (See B.2). This would be a correct answerif a declared emergency were in progress and emergency actionswere required. The Shift Supervisor could issue a key from the SSS office.

K/A: G2.3.12 Knowledge of radiological safety principles pertaining tolicensed operator duties, such as containment entry requirements, fuel handling responsibilities, access tolocked high-radiation areas, aligning filters, etc.Importance Rating: 3.2 3.7Technical

Reference:

FNP-0-RCP-0, General Guidance and Special Instructions to Health Physics Personne, Ver 68 FNP-0-RCP-0.1, Key Control Program and Health Physics Guidance for Control of High Radiation areas, Locked High Radiation Areas, and very High Radiation Areas, Ver 18References provided: NoneLearning Objective:

List four types of areas posted based on radiation levels andthe radiation levels/distances that require them to be posted(OPS30401A22)Question History: MOD SUMMER 11 K/A match: Requires the applicant to know the radiological safetyprinciple (value at which the locked high radiation isposted) and the requirements to enter a locked highradiation area.SRO justification: N/A

65. G2.3.13 065Which one of the following completes the statements below for entry into the Dry CaskStorage Radiation Controlled Area (RCA)?Per AP-42, Access Control, the operator (1) required to log in on the normalAuxiliary Building Access Control System (ACS) terminal prior to entering the DryCask Storage Area RCA.

Upon exiting from the Dry Cask RCA, the oper ator is required to perform a 2 minutefrisk and also use the (2) .1) IS 2) Primary Access Point (PAP) exit portal monitors1) IS 2) Auxiliary Building RCA exit portal monitors 1) is NOT

2) Primary Access Point (PAP) exit portal monitors1) is NOT

2) Auxiliary Building RCA exit portal monitors A.B.C.D.AP-42 rev 49.2:6.0 ENTRY INTO RCAS

6.2 Radiation

workers authorized entry into any RCA will ensure they have on theirperson, personnel monitoring device(s) assigned to them by Health Physics(Dosimetry), that being their dosimetry badge, and a self-issued digital alarmingdosimeter prior to entry into that RCA. 6.3 Entry into any RCA requires a Radiation Work Permit a nd issued personneldosimetry. Routine access to the main RCA will be through the hallway adjacentto the Health Physics Office.

6.3.4 Prior

to entering any RCA, each individual is responsible for ensuring that theymeet the requirements of the RWP under which they are entering.6.3.5 Prior to entry into any RCA each worker will either log in on ACS terminal orlog in using an alternate method which will be determined by Health Physics(e.g., manually logging personnel into and out of the RCA).6.3.6 Upon exit from the RCA each worker will log out at a ACS terminal except asnoted below.

6.3.7 Personnel

who are required to enter other RCA's where no ACS terminal exist (e.g. outside RCA's, temporary RCA's in Turbine Building, etc.), will either be requiredto use the normal Auxiliary Building ACS terminal or if available, a terminal which ismore convenient to the outside RCA. This may require individuals to transit back andforth while logged inside a RCA.6.3.8 Health Physics will implement an al ternate method of control when the ACSis inoperable.

6.4 Entry

into the LLRB and other RCAs outside the Protected Area

6.4.1 Personnel with Vital Area access will log into the RCA per step 6.3.

6.4.2 Upon completion of radiation work, personnel must either return to the HPOffice, log out per step 6.3 and check out via the Auxiliary Building RCA exitportal monitor, leaving digital dosimeters at appropriate locations (e.g., at theRCA exit) and other personnel monitoring devices as directed in step 6.2 or theymust log out and be monitored as directed by Health Physics.* Per the sign on the Dry Cask Storage Area access, frisking is required upon exitfrom that RCA.Distracter analysisA. Incorrect. First part is correct (See B.1). Second part is incorrect (See B.2). Plausible if the applicantassumes they can use the Primary Access Point (PAP) portal monitors since everyone who leaves the protected area passesthrough them.B. Correct First part is correct. Per step 6.3.5, the worker will use the AuxBuilding ACS terminal. Second part is correct. Per step 6.4.2, personnel must log out ofthe RCA and use the Aux Bldg exit portal monitor.C. Incorrect. First part is incorrect (See B.1). Plausible if the applicant believesthat since they are not entering the Aux building RCA, the ACSterminal entry is not required. Second part is incorrect (See A.2).

D. Incorrect. First part is incorrect (See C.1). Second part is correct (See B.2).

K/A: 2.3.13 Knowledge of radiological safety procedures pertaining tolicensed operator duties, such as response to radiationmonitor alarms, containment entry requirements, fuelhandling responsibilities, access to locked high-radiation areas, aligning filters, etc. Importance Rating: 3.4 3.8Technical

Reference:

FNP-0-AP-42, Access Control, Ver 49.2References provided: None Learning Objective: Outline the requirements and process for entry into an RCA(OPS40502M03). Question History: NEW K/A match: Requires the applicant to know the radiological safetyprocedure requirements (in this case AP-42, accesscontrol) to enter and exit an out building classified as aradiation controlled area.

SRO justification: N/A

66. G2.3.4 066Which one of the following completes the statement below?An employee who is a fully documented radiation worker and DOES NOT declare herpregnancy has an annual FNP Administrative TEDE limit of

.450 mRem500 mRem2000 mRem5000 mRem A.B.C.D.HP manual Step 4.1.3.3Any employee who discloses that she is or may be pregnant will complete the electionform (DOS Form 931 in FNP-0-DOS-2) to accept or decline a prenatal radiationexposure limit of 500 mrem (0.5 rem) for the embryo or fetus for the term of thepregnancy as recommended in 10CFR20.1208.FNP Admin Annual Dose Guidelines -

Fully documented radiation worker - 2000 mRem per year.Distracter analysisA. Incorrect. See C. Plausible since the is the Admin limit for the woman duringthe term of the pregnancy of a declared pregnant woman.B. Incorrect. See C. Plausible since this is the Federal Limit for the Embryo forthe term of the pregnancy.C. Correct. Since the woman has not declared her pregnancy, her adminexposure limit is 2000 mRem.D. Incorrect. See C. Plausible because this is the federal annual limit for andundeclared pregnancy.

K/A: G2.3.4 Knowledge of radiation exposure limits under normal oremergency conditions.Importance Rating: 3.2 3.7Technical

Reference:

FNP-0-M-001, SNC FNP Health Physics Manual, Ver 18 References provided: NoneLearning Objective:

List FNP Admin Limits for various categories of dose(OPS30401A20).Question History: NEW K/A match: Requires the applicant to know the normal exposurelimits for an un-declared pregnant woman.SRO justification: N/A

67. G2.4.20 067Unit 1 is performing the actions of EEP-3.0, Steam Generator Tube Rupture, due to atube rupture in the 1B SG.

• The 1B SG Narrow range level is 36% and rising.Which one of the following completes the statements below? The 1B SG narrow range level (1) adequate to begin the initial RCS cooldown. The operational implication of having sufficient level in the 1B SG prior to the cooldown is to (2) .1) is NOT2) ensure a secondary side heat sink1) is NOT2) prevent SG depressurization during the RCS cooldown1) IS2) ensure a secondary side heat sink1) IS

2) prevent SG depressurization during the RCS cooldown A.B.C.D.

EEP-3 Note prior to Step 4[CA] Maintaining ruptured SG(s) narrow range level greater than 31%{48%} preventsSG depressurization during RCS cooldown.FNP-0-EEB-3.0 version 2ERG Step Text:

Check Ruptured SG(s) LevelPurpose: 1. To reduce feed flow to the ruptured steam generators to minimize thepotential for steam generator overfill.

2. To establish and maintain a water level in the ruptured steam generators abovethe top of the U-tubes in order to promote thermal stratification to preventruptured steam generator depressurization.Basis: It is also important to maintain the water level in the ruptured steam generator abovethe top of the U-tubes. When the primary system is cooled in subsequent steps, thesteam generator tubes in the ruptured steam generator will approach the temperatureof the reactor coolant, particularly if reactor coolant pumps continue to run. If the steamspace in the ruptured steam generator expands to contact these colder tubes, condensation will occur which would decrease the ruptured steam generator pressure.

As previously demonstrated (see Step 3), this would reduce the reactor coolantsubcooling margin and/or increase primary-to secondary leakage, possibly delaying SItermination or causing SI reinitiation. Consequently, the water level must be maintainedabove the top of th e tubes to insulate the steam space.

In addition to insulating thesteam space, this ensures a secondary side heat sink in the event that no intactsteam generator is available and also provides protection against misdiagnosis of theruptured steam generator due to an imbalance of feed flow.Distracter analysisA. Incorrect. First part in incorrect (See D.1). Plausible since the applicant mayapply adverse numbers of 48% which would make this part correct. Second part in incorrect (See D.2). Plausible since this is anotherreason for having sufficient level in the ruptured SG ONLY if thereare NO intact SGs available. This is not the case in this question.B. Incorrect. First part in incorrect (See A.1). Second part is correct (See D.1).C. Incorrect. First part in correct (See D.1). Second part in incorrect (See A.2).

D. Correct. First part in correct. Ruptured SGWL must be > 31%. Second part is correct. Prevents SG depressurization during RCScooldown.

K/A: G2.4.20

Knowledge of the operational implications of EOP warnings,cautions, and notes.Importance Rating: 3.8 4.3Technical

Reference:

FNP-1-EEP-3, Steam Generator Tube Rupture, Ver 27 FNP-0-EEB-3.0, Specific Background Document forFNP-1/2 EEP-3, Ver 2References provided: NoneLearning Objective:

STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with EEP-3, Steam Generator Tube Rupture. (OPS-52530D03).Question History: MOD FNP EXAM BANKK/A match: Requires the applicant to know the operationalimplications of not meeting the Note and Caution ofEEP-3 by having the improper SG water level prior toRCS cooldown during a tube rupture event.

SRO justification: N/A

68. G2.4.23 068FRP-Z.1, Response to High Containment Pressure, has the following caution:

IF ECP-1.1, Loss of Emergency Coolant Recirculation, is in effect, THEN ContainmentSpray should be operated as directed in ECP-1.1.Which one of the following describes the bases for giving priority to ECP-1.1? ECP-1.1 directs the operation of the Containment Spray (CS) pumps to ensure

.RWST level is conservedadequate NPSH for the RHR pumps is availablethe maximum available Containment heat removal systems are runningautomatic swapover of the CS pumps to the Containment sump is prevented A.B.C.D.

FRP-Z.1 Caution prior to step 3IF FNP-1-ECP-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, is in effect,THEN containment spray should be operated as directed in FNP-1-ECP-1.1.FRB-Z.1 Step 3 BasisGuideline ECA-1.1 uses a less restrictive criteria, which permits reduced spray pumpoperation depending on RWST level, containment pressure and number of emergencyfan coolers operating. The less restrictive criteria for containment spray operation isused in guideline ECA-1.1 since recirculation flow to the RCS is not available and it isvery important to conserve RWST water, if possible, by stopping containment spraypumpsDistracter analysisA. Correct. Per above basis statement:

The less restrictive criteria forcontainment spray operation is used in guideline ECA-1.1 since recirculation flow to the RCS is not available and it is veryimportant to conserve RWST water, if possible, by stoppingcontainment spray pumpsB. Incorrect. See A. Plausible since ECP-1.1 is Loss of Emergency CoolantRecirculation and the applicant may think that ECP-1.1 operatesthe spray pumps to maximize sump level to allow the RHR pumps to get a proper suction for alignment to sump recirculation.C. Incorrect. See A. Plausible since this is the goal of FRP-Z.1 and the applicantcould confuse the two procedures. D. Incorrect. See A. Plausible since there is an auto swap over for RHR sumpsuction valves but not for Containment Spray pumps. The applicantcould confuse these and think that ECP-1.1's mitigation addressedthis issue.

K/A: G2.4.23 Knowledge of the bases for prioritizing emergencyprocedure implementation during emergency operations.Importance Rating: 3.4 4.4Technical

Reference:

FNP-1-FRP-Z.1, Response to High Containment Pressure,Ver 15 FNP-0-FRB-Z.1, Specific Background Document forFNP-1/2-FRP-Z.1, Ver 1References provided: NoneLearning Objective:

STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with (1) FRP-Z.1, Response to High Containment Pressure; [...]. (OPS-52533M03) Question History: FNP 08 K/A match: Requires the applicant to know the basis for operating theContainment Spray pumps per ECP-1.1 versus FRP-Z.1(prioritizing emergency procedure implementationduring emergency operations).SRO justification: N/A

69. G2.4.29 069Unit 1 is operating at 100% power and the following conditions exist: * #1 Waste Monitor Tank (WMT) release is in progress.

• The Unit 1 Rad Side SO is at the RCA exit preparing to enter the portal monitors.Subsequently, the plant emergency alarms sounds and an announcement is madedeclaring a Site Area Emergency.Which one of the following completes the statements below? The Rad Side SO will go to the designated assembly area (1) . The designated assembly area for the Rad Side SO is the (2) .1) after securing the #1 WMT release2) Operations Support Center (OSC)1) after securing the #1 WMT release2) Control Room1) immediately2) Operations Support Center (OSC)1) immediately

2) Control Room A.B.C.D.

EIP-0.0 4.4.5 System Operators (2), plant operations.

• Assigned assembly area is the Control Room.4.4.6 Other System Operators (as required by Technical Specifications), plantoperations.

• Assigned assembly area is the Control Room.EIP-104.10 During outages and normal Monday through Friday day shifts, individuals NOTdescribed in section 4.1 thru 4.8 above will report to their assembly area as follows:

[...]OPS Group on shift or qualified for a Shift position, and NOT in Training - Control RoomDistracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible since this would stop arelease to the environment. However, the release is monitored andwill stop if R-18 alarms and the pump will trip on low level which would render the system safe. Second part is incorrect (See D.2). Plausible since this is anassembly area per procedure and the applicant may not recall theproper assembly areas for on shift staff. The OSC is extremelyclose to the Control Room and where all other personnelassemble. Prior to the new protected area, onshift staffing didassemble in the OSC so this was a normal assembly location foron shift OPS personnel in the recent past.B. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2).

C. Incorrect. First part is correct (See D.1). Second part is incorrect (See A.2).D. Correct. First part is correct.NMP-EP-111-001 (pg 17) page announcementhas a section to give directions to personnel out in the field if the intent is to allow them to remain in the field. The stem does notindicate that this specific direction is given so the SO willimmediately go to his/her assembly area. Second part is correct. The control room is the proper assembly area per EIP-0.0 and EIP-10.

K/A: G2.4.29 Knowledge of the emergency plan.Importance Rating: 3.1 4.4 Technical

Reference:

NMP-EP-111-001, Emergency Notification NetworkCommunicator Instructions - Farley, Ver 3.2 FNP-0-EIP-0.0, Emergency Organization, Ver 29References provided: NoneLearning Objective:

IDENTIFY AND EXPLAIN the actions to be taken by anindividual following an evacuation announcement (OPS40501B04). Question History: NEW K/A match: Require the applicant to have knowledge of anindividual's responsibilities when the Emergency Plan isactivated.SRO justification: N/A

70. W/E04EA1.1 070The crew has transitioned to ECP-1.2, LOCA Outside Containment.* Step 2 of ECP-1.2 is in progress and the first flow path has been isolated.The following conditions exist:* Aux Building radiation levels are rising slowly.

• Safety Injection flow is stable.

• Aux Building sump levels are rising slowly.

• PI-402 and 403, RCS 1C/1A LOOP RCS NR PRESS, are rising.Which one of the following completes the statements below per ECP-1.2? The first flow path that was isolated was (1) injection.The intersystem LOCA (2) been isolated.

(1)

(2)

RCP seal HAS RCP seal has NOT

RHR cold leg HAS RHR cold leg has NOT A.B.C.D.

ECP-1.2Step 2: Try to identify and isolate break.2.1 Isolate A train RHR cold leg injection path.

2.2 Check

RCS pressure - RISING.

2.5 Isolate

B train RHR cold leg injection path.

2.6 Check

RCS pressure - RISING.Distracter analysisA. Incorrect. First part is incorrect (See C.1). Plausible since this is isolatedduring ECP-1.2 but not first. Second part is correct (See C.2).

B. Incorrect. First part is incorrect (See A.1). Second part is incorrect (See C.2). Plausible if the applicant doesnot recall which parameter is used to check leak isolation. Auxbuilding sump levels and radiation levels could cause the applicantto believe that the leak is not isolated. Once the leak was isolated,sump levels could continue to rise as well as radiation levels as theisolated piping drains.C. Correct. First part is correct. Per step 2 of ECP-1.2, LHSI (RHR cold leginjection) is isolated first. Second part is correct. Per ECP-1.2, RCS pressure rising is theparameter monitored for verifying the leak is isolated.D. Incorrect. First part is incorrect (See A.1).

Second part is incorrect (See B.2).

K/A: W/E04EA1.1 LOCA Outside Containment - Ability to operate and / ormonitor Components, and functions of control andsafety systems, including instrumentation, signals,interlocks, failure modes, and automatic and manual features as they apply to the (LOCA Outside Containment)

Importance Rating: 4.0 4.0 Technical

Reference:

FNP-1-ECP-1.2, LOCA Outside Containment, Ver 8References provided: NONELearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing ECP-1.2,LOCA Outside Containment. (OPS-52532E06)

ANALYZE plant conditions and DETERMINE the successfulcompletion of any step in ECP-1.2, LOCA OutsideContainment. (OPS-52532E07)

Question History: NEWK/A match: Requires the applicant to know which components areoperated and be able to determine from listedinstrumentation (monitor) if the leak has stopped duringthe leak isolation phase of ECP-1.2, LOCA Outside ofContainment.SRO justification: N/A

71. W/E05EG2.4.2 071Unit 1 was operating at 100% power when a Reactor Trip and SI occurred due to asteam line break in containment. The following conditions exist:

• The operating crew is performing the actions of EEP-2.0, Faulted Steam Generator Isolation.

• The maximum total AFW flow rate that can be achieved is 350 GPM.

• Containment pressure is 6 psig and falling.

• SG Narrow range levels are:

- 1A - Off Scale Low

- 1B - 32% and decreasing slowly - 1C - 34% and decreasing slowlyWhich one of the following completes the statement below?Secondary heat sink (1) adequate because (2) .1) IS2)

1B and 1C SG levels are sufficient1) is NOT 2) Neither SG levels nor AFW flow capability is sufficient1) IS2) AFW flow capability is sufficient1) is NOT

2) 1B and 1C SG levels are sufficient but AFW flow capability is NOT sufficient A.B.C.D.

CSF-0/0.3 Heat Sink:

To have adequate heat sink -SG Narrow Range levels in at least ONE SG greater than 31%{48%}

OR Total AFW to all SG's > 395 gpmDistracter analysisA. Incorrect. See B. Plausible if the applicant does not recognize that adversenumbers apply in this scenario then this would be correct.B. Correct. One SG NR level must be >48% OR AFW flow must be >395 gpmto satisfy the heat sink criteria.C. Incorrect. See B. Plausible if the applicant does not recall the minimum AFWflow required to meet heat sink and confuses it with 350 gpm whichis the design flow rate of one AFW pump.D. Incorrect. See B. Plausible if the applicant believes that BOTH AFW flow andSG NR levels are required to meet heat sink and does notrecognize that adverse numbers apply in this scenario.

K/A: W/E05EG2.4.2 Loss of Secondary Heat Sink - Knowledge of system set points, interlocks and automatic actions associated withEOP entry conditions

.Importance Rating: 4.5 4.6Technical

Reference:

FNP-1-CSF-0, Critical Safety Function Status Trees, Ver 17References provided: None Learning Objective:

EVALUATE plant conditions and DETERMINE if entry into(1) FRP-H.1, Response to Loss of Secondary Heat Sink; [..]is required. (OPS-52533F02)

Question History: MOD FNP EXAM BANK K/A match: Requires the applicant to know the setpoints of CSF-0 andrecognize that heat sink does not exist and the setpointswhich are met for entry into FRP-H.1, Response to Loss of Secondary Heat Sink.SRO justification: N/A

72. W/E08EG2.4.6 072Unit 2 has experienced a large steam break inside containment and the followingconditions exist:

• 1A SG Wide Range level indicates 0%.* AFW flow to 1B and 1C SG is currently 500 gpm.* FRP-P.1, Response To Imminent Pressurized Thermal Shock, has beenentered on a RED Path.* RCS cold leg temperature continues to decrease slowly.

• RCS pressure is 1500 psig and stable. Which one of the following completes the statements below concerning the mitigationstrategy of FRP-P.1?The RCS cooldown must (1) .An RCS pressure reduction (2) required.

(1)

(2) continue is NOT continue IS be stopped is NOT be stopped IS A.B.C.D.

FRP-P.1:Step 3: Check cold leg temperature 3. [CA] Stop the cooldown. STABLE OR RISING.Step 18: Reduce RCS pressure.FRB-P.1Step 3: [..] It is important to terminate, if possible, any cooldown in progress to limit theextent of possible vessel damage due to excessive thermal stresses. [...]Step 18:

The RCS pressure reduction is intended to decrease pressure stress on the vesselwall as much as possible. [...]Distracter analysisA. Incorrect. First part is incorrect (See D.1). Plausible since a cooldown will beperformed after the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> soak but the overall strategy of FRP-P.1is to stop the cooldown. Second part is incorrect (See D.1). Plausible since FRP-P.1requires the RCS pressure stable for the soak and the applicantcould confuse this with current conditions in the stem.B. Incorrect. First part is incorrect (See A.1). Second part is correct (See D.2).C. Incorrect. First part is correct (See D.2).

Second part is incorrect (See A.2)D. Correct. First part is correct. Per the background document - [...] It isimportant to terminate, if possible, any cooldown in progress to limit the extent of possible vessel damage due to excessivethermal stresses. Second part is correct. Per the background document -

The RCS pressure reduction is intended to decrease pressure stress on thevessel wall as much as possible.

K/A: W/E08EG2.4.6 Pressurized Thermal Shock -

Knowledge of EOP mitigationstrategies.Importance Rating: 3.7 4.7Technical

Reference:

FNP-2-FRP-P.1, Response to Imminent PressurizedThermal Shock Conditions, Ver 23 FNP-0-FRB-P.1, Specific Background Document ForFNP1/2-FRP-P.1, Ver 2References provided: NONE Learning Objective:

STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with (1) FRP-P.1, Response to ImminentPressurized Thermal Shock Condition; [...] (OPS-52533K03) EVALUATE plant conditions and DETERMINE if anysystem components need to be operated while performing(1) FRP-P.1, Response to Imminent Pressurized ThermalShock Condition; [...] (OPS-52533K06). Question History: FNP 08 K/A match:

Requires the applicant to know the mitigation strategy ofFRP-P.1, Response to Imminent Pressurized Thermal Shock Conditions, and select the appropriate actions totake under the given conditions.SRO justification: N/A

73. W/E11EK2.1 073A Dual Unit LOSP with a LOCA on Unit 1 has occurred and the following conditionsexist:

• EEP-1.0, Loss of Reactor or Secondary Coolant, is in progress.At 1000:

• WA2, 1-2A DG GEN FAULT TRIP, comes into alarm.At 1015:

• The following alarms are received:

- CF3, 1A OR 1B RHR PUMP OVERLOAD TRIP - CH2, RWST LVL A TRN LO

- CH3, RWST LVL B TRN LOWhich one of the following states:1) the status of Unit 1 emergency recirculation capability and

2) the action(s) that the applicable procedure(s) direct?1) One train ONLY of emergency recirculation capability has been lost. 2) Transfer to Cold Leg AND Containment Spray recirculation at this time.1) One train ONLY of emergency recirculation capability has been lost. 2) Transfer to Cold Leg recirculation ONLY.1) Both trains of emergency recirculation capability have been lost. 2) Minimize HHSI flow to the minimum required to remove decay heat whileattempting to restore at least one train of emergency recirculation. 1) Both trains of emergency recirculation capability have been lost. 2) Secure HHSI pumps while attempting to restore at least one train of emergencyrecirculation.

A.B.C.D.This is not a true 2+2 question to improve distracter plausibility.EEP-1: 13.1 Verify cold leg recirculation 13.1 IF cold leg recirculationcapability - AVAILABLE. capability can NOT be verified, THEN go to FNP-2-ECP-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION. 13.1.1 Train A equipment available: - 2A RHR Pump- CTMT SUMP TO 2A RHR PUMP Q2E11MOV8811A

- CTMT SUMP TO 2A RHR PUMPQ2E11MOV8812A

- 2A RHR HX TO CHG PUMPSUCT Q2E11MOV8706A

- CCW TO 2A RHR HXQ2P17MOV3185A OR13.1.2 Train B equipmentavailable:

- 2B RHR Pump

- CTMT SUMP TO 2B RHR PUMPQ2E11MOV8811B

- CTMT SUMP TO 2B RHR PUMPQ2E11MOV8812B

- 2B RHR HX TO CHG PUMPSUCT Q2E11MOV8706B

- CCW TO 2B RHR HXQ2P17MOV3185BECP-1.1 Purpose -

This procedure provides actions to restore emergency coolant recirculationcapability, to delay depletion of the RWST by adding makeup and reducing outflow,and to depressurize the RCS to minimize break flow.Distracter AnalysisA. Incorrect. First part is incorrect (See C.1). Plausible if the applicant doesn'trecognize that the DG trip results in the loss of the 1A RHR pump.

Second part is incorrect (See C.2). Plausible if the applicant thinksthat one train of recirc capability is available because this would be partially correct. The containment spray is not transferred to sumprecirc until the RWST is less than 4.5 ft. The applicant could beunfamiliar with the procedure and believe that both cold leg and containment spray are required to be transferred to sump recircwhen RWST is at 12.5 ft.B. Incorrect. First part is incorrect (See A.1) Second part is incorrect (See C.2). Plausible since this would bethe correct answer if recirc capability existed.C. Correct. First part is correct. Since neither RHR pumps are available so there is no recirculation capability. Second part is correct. This is the correct strategy for ECP-1.1.D. Incorrect. First part is correct (See C.1).

Second part is correct (See C.2). Plausible if the applicantrecognizes that recirculation capability is lost but incorrectlybelieves that CH2 and CH3 being in alarm indicates that theRWST is less than 4.5 ft which would make this the correct answerper step 34 of ECP-1.1.

K/A: W/E11EK2.1 Loss of Emergency Coolant Recirculation -

Knowledge ofthe interrelations between the (Loss of Emergency CoolantRecirculation) and the following: Components, and functionsof control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic andmanual features.Importance Rating: 3.6 3.9Technical

Reference:

FNP-1-EEP-1.0, Loss of Reactor or Secondary Coolant, Ver 31.

FNP-1-ECP-1.1, Loss of Emergency Coolant Recirculation, Ver 30References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if entry into(1) ECP-1.1, Loss of Emergency Coolant Recirculation; and/or (2) ECP-1.3, Loss of Emergency CoolantRecirculation, Caused by Sump Blockage is required.(OPS-52532D02)

Question History: FNP 10 K/A match: Applicant is required to know the interrelation betweenfailure modes of the RHR pumps and the Loss ofEmergency Coolant Recirculation procedure.SRO justification: N/A

74. W/E12EK1.1 074The crew is responding to a Steam Line Break on Unit 1.

• Due to equipment failures, ECP-2.1, Uncontrolled Depressurization of All Steam Generators, has been entered.

• All SG Narrow Range levels are 25% and lowering.Which one of the following completes the statement below?Per ECP-2.1, AFW flow will be adjusted to (1) .0 gpm to each SG to prevent excessive cooldown at least 20 gpm to each SG to prevent dryout of the SGsat least 20 gpm total AFW flow to minimize thermal stress to the SGsat least 395 gpm

total AFW flow to maintain adequate heat sink A.B.C.D.

ECP-2.1Step 4: [CA] Maintain at least 20 gpm AFW flow to SGs with narrow range level less than 31%{48%}.ECB-2.1Basis: If feed flow to a SG is isolated and the SG is allowed to dry out, subsequentreinitiation of feed flow to the SG could create significant thermal stress conditions onSG components. Maintaining a minimum verifiable feed flow to the SG allows thecomponents to remain in a "wet" condition, thereby minimizing any thermal shock effects if feed flow is increased.Distracter AnalysisA. Incorrect. See B. Plausible because stopping AFW flow would deplete thewater inventory and stop the cooldown. This is the strategy ofEEP-2.0, to stop all feed flow to the SG and allow it to blow down.This is not allowed by this procedure and would result in a loss ofheat sink.B. Correct. Per ECP-2.1, the operator is required to maintain at least 20 gpmAFW flow to SGs with narrow range level less than 31%{48%} to prevent dryout of the SGs. C. Incorrect. See B. Plausible if the applicant confuses AFW to EACH versusTotal AFW flow. If this were 20 gpm to EACH SG it would be acorrect answer.D. Incorrect. See B. Plausible since this meets the Heat Sink Critical SafetyFunction Status Tree. FRP-H.1, Response to Loss of SecondaryHeat Sink, has a caution that says the following: This procedureshould not be performed if total AFW flow is less than 395 gpmdue to operator action.

K/A: W/E12EK1.1 Uncontrolled Depressurization of all Steam Generators -Knowledge of the operational implications of the followingconcepts as they apply to the (UncontrolledDepressurization of all Steam Generators): Components:,capacity, and function of emergency systems.Importance Rating: 3.4 3.8 Technical

Reference:

FNp1-ECP-2.1, Uncontrolled Depressurization of All SteamGenerators, Ver 24 FNP-0-ECB-2.1, Specific Background Document forFNP-1/2-ECP-2.1, Ver 1References provided: NoneLearning Objective:

EVALUATE plant conditions and DETERMINE if any systemcomponents need to be operated while performing ECP-2.1,Uncontrolled Depressurization of All SGs. (OPS-52532F06)Question History: NEWK/A match: Requires the applicant to know the AFW flow rate forECP-2.1 and the operational implication of lowering AFWflow (AFW is an emergency system) to 20 gpm (reducingpump capacity) during an Uncontrolled Depressurization ofall Steam Generators. Each flow rate has a different operational implication to improve plausibility.SRO justification: N/A

75. W/E15EA1.3 075Which one of the following is the first Major Action Category in FRP-Z.2, Response ToContainment Flooding, and reason for this in accordance with the background document?Identify unexpected sources of water in the sump since flooding could damagecritical plant equipment.Evaluate the ECCS system status to determine a strategy to transition tosimultaneous cold and hot leg recirculation.Have chemistry evaluate sump level, chemistry, and activity level to determine astrategy to transfer excess water out of containment.Notify the TSC of sump chemistry, and activity level to determine potential changesin the planned transition to simultaneous cold and hot leg recirculation

.A.B.C.D.

FRP-Z.2Step 1: Try to identify source of water into sump.* Check indications forcomponents supplied with service water.* Check indications forcomponents supplied with CCW.* Check indication of ReactorMakeup Water Storage Tank level.* Check indication ofDemineralized Water Storage Tank level.FRB-Z.2

Background:

Step 1 Basis: This step instructs the operator to try to identify the unexpected source ofthe water in the containment sump. Containment flooding is a concern since criticalplant components necessary for plant recovery may be damaged and renderedinoperable.Distracter analysisA. Correct. Step 1 of FRP-Z.2 directs evaluating potential sources of flooding.The background document states - This step instructs the operator to try to identify the unexpected source of the water in thecontainment sump. Containment flooding is a concern since critical plant components necessary for plant recovery may be damagedand rendered inoperable.B. Incorrect. See A. Plausible since the ECCS system does enter containmentand the applicant may improperly think that this is a source offlooding. If the ECCS system were damaged, then determining astrategy for going on to simultaneous cold and hot leg recirculationwould be a plausible reason for this step. There is no step toevaluate ECCS as a source of flooding as it is designed to put water into the recirculation sump via the RCS break.C. Incorrect. See A. Plausible since this is Step 2 and the basis for this step inFRP-Z.2. The applicant may not be familiar with the procedure andbelieve that this is the first step.D. Incorrect. See A. Plausible since Step 3 does have the TSC evaluate sumpchemistry, and activity level but not for this reason.

K/A: W/E15EA1.3 Containment Flooding - Ability to operate and / or monitorthe following as they apply to the (Containment Flooding):Desired operating results during abnormal and emergencysituations. Importance Rating: 2.8 3.0 Technical

Reference:

FNP-1-FRP-Z.2, Response To Containment Flooding, Ver 6 FNP-0-FRB-Z.2, Specific Background Document for FNP-1/2-FRP-Z.2, Ver 1References provided: None Learning Objective:

STATE AND EXPLAIN the basis for all Cautions, Notes, andActions associated with [...] ; (2) FRP-Z.2, Response to Containment Flooding; [...]. (OPS-52533M03)Question History: SUMMER 11 K/A match: Requires to applicant to monitor containment sump flooding sources and recognize the undesired operatingresults of not isolating flooding.SRO justification: N/A ANSWER KEY REPORTfor ILT-36 SRO NRC Exam Test Form: 0 Answers 0#IDPoints1001AA2.01 11.00C2003A3.05 21.00C 3004K3.08 31.00C 4005A2.02 41.00B 5006K5.06 51.00B 6006K6.18 61.00B 7007EA2.06 71.00C 8007K5.02 81.00B 9008AK1.01 91.00D 10008K2.02 101.00D 11009EK2.03 111.00C 12010A2.01 121.00B 13011EK2.02 131.00D 14011K2.02 141.00C 15012A4.06 151.00A 16013G2.1.19 161.00A013G2.1.19 P 16 17014K5.01 171.00D 18015/17AA2.02 181.00A 19017K6.01 191.00D 20022A1.03 201.00B 21022AK1.04 211.00A 22022K4.03 221.00A 23025AA1.01 231.00D 24026AG2.4.50 241.00D 25026K2.01 251.00C 26029A1.02 261.00A 27033K4.05 271.00C 28035A2.01 281.00B 29036AG2.1.7 291.00D 30037AK1.01 301.00C 31038EK3.02 311.00C 32039K1.09 321.00A 33045K1.19 331.00B 34051AK3.01 341.00A 35054AA2.05 351.00C 36055A3.03 361.00C 37056AK3.02 371.00A 38059A4.01 381.00A 39059AK2.01 391.00B 40059G2.2.44 401.00B 41061A3.01 411.00A 42061K5.01 421.00A 43062AA1.07 431.00A 44062K3.01 441.00B 45063A1.01 451.00A 46064K3.02 461.00D 47064K6.07 471.00D ANSWER KEY REPORTfor ILT-36 SRO NRC Exam Test Form: 0 Answers 0#IDPoints48065AG2.4.11 481.00C49068AK2.07 491.00D 50071K3.05 501.00A 51073A2.02 511.00B 52073G2.2.42 521.00B 53076AA1.04 531.00B 54076K2.08 541.00C 55076K4.03 551.00D 56077AK3.02 561.00A 57078K1.01 571.00A 58079G2.4.34 581.00D 59103K4.06 591.00C 60G2.1.18 601.00C 61G2.1.5 611.00B 62G2.2.20 621.00C 63G2.2.25 631.00A 64G2.3.12 641.00C 65G2.3.13 651.00B 66G2.3.4 661.00C 67G2.4.20 671.00D 68G2.4.23 681.00A 69G2.4.29 691.00D 70W/E04EA1.1 701.00C 71W/E05EG2.4.2 711.00B 72W/E08EG2.4.6 721.00D 73W/E11EK2.1 731.00C 74W/E12EK1.1 741.00B 75W/E15EA1.3 751.00ASECTION 1 ( 75 items)75.0076001A2.17 761.00C 77004A2.02 771.00A 78006G2.2.22 781.00B 79008AG2.2.22 791.00A 80024AA2.04 801.00B 81028G2.1.1 811.00B 82029EA2.04 821.00D 83051AG2.1.7 831.00A 84054AA2.01 841.00C 85055EG2.2.12 851.00D 86058AA2.02 861.00B 87059A2.12 871.00C 88064G2.2.44 881.00B 89071G2.2.25 891.00D 90073A2.01 901.00A 91077AG2.4.31 911.00C 92G2.1.20 921.00D ANSWER KEY REPORTfor ILT-36 SRO NRC Exam Test Form: 0 Answers 0#IDPoints93G2.1.37 931.00A94G2.2.21 941.00B 95G2.2.23 951.00A 96G2.3.14 961.00D 97G2.3.6 971.00C 98G2.4.6 981.00D 99WE02EA2.1 991.00C 100WE06EG2.1.27 1001.00BSECTION BREAK ( 25 items)25.00