Text

Initial Exam 2013-301 Final SRO Written Exam
	ML13191A891
Person / Time
Site:	Farley
Issue date:	07/10/2013
From:	; NRC/RGN-II
To:	; Southern Nuclear Operating Co
References
	ES-401
	Download: ML13191A891 (434)
	v • d • e

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: 06/25/13 Facility/Unit: FARLEY Unit 1 & 2 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete the combined examination, and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months if you are only taking the SRO portion.

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

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

1. Unit 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 the Channel 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 Bypass Breaker (2) light will be LIT.

(1) (2)

A. insert RED B. insert GREEN C. withdraw GREEN D. withdraw RED

2. Unit 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, Reactor Coolant System?

The 1C RCP oil lift pump handswitch white light indicates (1) .

The 1C RCP breaker closing operation (2) interlocked with a 2 minute time delay.

A. 1) BOTH the oil lift pressure has reached 600 psig AND 2 minutes have elapsed

2) is NOT B. 1) ONLY that the oil lift pressure has reached 600 psig

2) IS C. 1) ONLY that the oil lift pressure has reached 600 psig

2) is NOT D. 1) BOTH the oil lift pressure has reached 600 psig AND 2 minutes have elapsed

2) IS

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 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)

A. CLOSED LOWER B. CLOSED RISE C. OPEN LOWER D. OPEN RISE

4. The 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 be performed per AOP-12.0, Residual Heat Removal Malfunction?

A. No actions are required.

B. Secure the 1A RHR pump ONLY.

C. Secure the 1B RHR pump ONLY.

D. Secure BOTH 1A AND 1B RHR pumps.

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

6. Unit 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)

A. be affected NOT be affected B. NOT be affected be affected C. be affected be affected D. NOT be affected NOT be affected

7. Unit 2 was operating at 2% power with a plant startup in progress per 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 completes 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)

A. OPEN WILL B. OPEN will NOT C. CLOSED WILL D. CLOSED will NOT

8. Unit 1 is in Mode 5 and forming a pressurizer steam space (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 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)

A. charging flow remain constant B. letdown flow remain constant C. letdown flow rise D. charging flow rise

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 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)

A. 546°F rising B. 546°F lowering C. 320°F lowering D. 320°F rising

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)

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

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 level must be greater than (1) to (2) .

A. 1) 31%

2) ensure SG tubes are covered to promote reflux boiling B. 1) 48%

2) ensure SG tubes are covered to promote reflux boiling C. 1) 31%

2) ensure adequate SG inventory to provide a secondary heat sink D. 1) 48%

2) ensure adequate SG inventory to provide a secondary heat sink

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 describes Pressurizer Heater operation per ESP-0.1, Reactor Trip Response?

A. The 1A PZR Heaters will have power available, NO actions are required to energize them.

B. The 1A PZR Heaters will have power available AND manual actions on the MCB are required to energize them.

C. The 1A PZR Heaters will NOT have power available. Manual actions are required to align power to them on the EPB but NO other actions are required to energize them.

D. The 1A PZR Heaters will NOT have power available. Manual actions are required to align power to them on the EPB AND manual actions on the MCB are required to energize them.

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 Supervisor directs the OATC to perform ESP-1.4, Transfer to Simultaneous 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 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

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)

A. HOT COLD B. COLD COLD C. HOT HOT D. COLD HOT

14. Unit 1 is operating at 100% power when the 1E 4160V bus becomes de-energized due to an electrical fault.

Which one of the following completes the statement below?

Pressurizer heater groups have lost their normal power supply.

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

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.

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

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 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)

A. IS are NOT B. IS ARE C. is NOT are NOT D. is NOT ARE

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 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) the most reliable indication.

A. 1) exactly 100 steps

2) the group step counters B. 1) approximately 100 steps

2) the group step counters C. 1) exactly 54 steps

2) DRPI D. 1) approximately 54 steps

2) DRPI

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 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) .

A. 1) Motor Bearings

2) 2 minutes B. 1) Motor Bearings

2) 60 minutes C. 1) Pump Lower Radial Bearings

2) 2 minutes D. 1) Pump Lower Radial Bearings

2) 60 minutes

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 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 Degraded Core Cooling.

(1) (2)

A. HIGH hottest B. HIGH 5th hottest C. LOW hottest D. LOW 5th hottest

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 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)

A. ductwork MORE B. dropout plate MORE C. ductwork LESS D. dropout plate LESS

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 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 would decrease to a minimum flow rate of (1) , which is designed to prevent (2) .

A. 1) 18 gpm

2) flashing downstream of the letdown orifices B. 1) 18 gpm

2) overheating of the charging pumps C. 1) 40 gpm

2) flashing downstream of the letdown orifices D. 1) 40 gpm

2) overheating of the charging pumps

22. Unit 1 is operating at 100% 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 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)

A. IS is NOT B. is NOT is NOT C. IS IS D. is NOT IS

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 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 preferred method to re-establish core cooling is to establish (1) .

Core cooling is monitored using (2) .

(1) (2)

A. feed and bleed RCS cold leg temperatures B. a secondary heat sink RCS cold leg temperatures C. feed and bleed CETCs D. a secondary heat sink CETCs

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 completes the statements below?

CCW system automatic isolations are designed to occur at (1) in the CCW Surge Tank.

Using the NORMAL source of makeup water, the operator will open (2) to make up to the CCW Surge Tank.

(1) (2)

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

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)

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

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 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)

A. WILL will NOT B. will NOT will NOT C. WILL WILL D. will NOT WILL

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 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, the MINIMUM required SFP boron concentration is (2) ppm.

(1) (2)

A. boration 2000 B. boration 2200 C. dilution 2000 D. dilution 2200

28. Unit 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 are required to (2) .

(1) (2)

A. steam reduce turbine load B. steam trip the reactor C. feed reduce turbine load D. feed trip the reactor

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 reports 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 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)

A. AOP-49.3 make up to the SFP using the RWST B. AOP-30.0 make up to the SFP using the RWST C. AOP-49.3 ensure all SFP hatches and doors are closed D. AOP-30.0 ensure all SFP hatches and doors are closed

30. Unit 1 is performing the actions of AOP-2.0, Steam Generator Tube Leakage, due to a tube 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)

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

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 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)

A. has NOT Steam Generator B. has NOT Pressurizer C. HAS Steam Generator D. HAS Pressurizer

32. Concerning 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)

A. upstream 25%

B. downstream 25%

C. upstream 10%

D. downstream 10%

33. Which one of the following coincidences will cause an anticipated transient without 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)

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

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

A. Plant Trip CLOSED B. Plant Trip OPEN C. Loss of Load CLOSED D. Loss of Load OPEN

35. Unit 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, controllers are 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)

A. remain OPEN remain OPEN B. remain OPEN CLOSE C. CLOSE remain OPEN D. CLOSE CLOSE

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)

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

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 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 at the SG pressure specified in ECP-0.0 is to prevent (2) .

(1) (2)

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

38. Unit 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)

A. LIT LIT B. LIT NOT lit C. NOT lit LIT D. NOT lit NOT lit

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.

A. HV-7614A/B/C, 2A/B/C SGBD ISO B. FCV-1152, SGB INLET STOP VALVE C. RCV-023B, SGBD DISCH TO ENVIRONMENT D. HV-7697A/B, 7698A/B and 7699A/B, 2A/B/C SGBD ISO

40. Unit 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 are observed:

KB4, SGFP SUCTION PRESS LOW, comes into alarm and the operating crew observes the following on PR4039, SGFP SUCT PRESS:

Time 0 sec 10 sec 20 sec 30 sec 40 sec 300 psig 275 psig 265 psig 270 psig 285 psig At 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) .

A. 1) will NOT

2) rapidly reduce Turbine load using AOP-17.1, Rapid Turbine Power Reduction B. 1) will NOT

2) check SGFP suction pressure stabilizes C. 1) WILL

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

D. 1) WILL

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

41. Unit 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)

A. will NOT 350 B. will NOT 700 C. WILL 700 D. WILL 1050

42. Unit 1 was operating at 100% power when a Reactor Trip occurred and the following conditions 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 to address the cooldown?

A. Minimize total AFW flow.

B. Emergency borate the RCS.

C. Close all MSIVs and MSIV Bypass Valves.

D. Manually initiate SI and return to EEP-0.0, Reactor Trip or Safety Injection.

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 has dropped to 179 feet, 10 inches.

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

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

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

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

2) RCV-018, WMT DISCH TO ENVIRONMENT, will automatically close.

C. 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.

D. 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.

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.

A. 2C Instrument Air Compressor B. 2C Reactor Coolant Pump Motor Air Cooler C. 2C Component Cooling Water Heat Exchanger D. Steam Generator Blowdown Heat Exchanger

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 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 sequence needed loads if power is not restored to the 125V DC battery chargers on each train within a MINIMUM of (2) .

A. 1) drop slowly at first; then later drop rapidly as the battery nears exhaustion

2) 30 min B. 1) drop slowly at first; then later drop rapidly as the battery nears exhaustion

2) 90 min C. 1) drop at a constant, linear rate the entire time the battery discharges

2) 30 min D. 1) drop at a constant, linear rate the entire time the battery discharges

2) 90 min

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 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) .

A. 1) Mode 2

2) automatically start B. 1) Mode 2

2) have to be manually aligned C. 1) Mode 1

2) automatically start D. 1) Mode 1

2) have to be manually aligned

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 DGs required minimum time to reach rated speed and voltage is (2) seconds after receiving an emergency start signal.

(1) (2)

A. 200 7 B. 200 12 C. 350 7 D. 350 12

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 Instrument Air, which one of the following methods below will 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 VLV A. Place BOTH MDAFW pump MCB hand switches in the STOP position and release them.

B. Throttle HV-3228A / B / C on the MCB.

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

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

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 completes the statements below?

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

(1) (2)

A. MODE 3 WILL B. MODE 3 will NOT C. MODE 4 WILL D. MODE 4 will NOT

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 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)

A. WILL CANNOT B. WILL CAN C. will NOT CANNOT D. will NOT CAN

51. Unit 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) .

A. 1) HV-3328, HV-3329 AND HV-3330, STEAM GEN 1A/1B/1C SAMPLE ISO,

2) pull the INSTRUMENT power fuses for R-19 B. 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 C. 1) HV-3179A, 3180A, AND 3181A, STEAM GEN 1A/1B/1C LOWER BLOWDOWN,

2) pull the INSTRUMENT power fuses for R-19 D. 1) HV-3179A, 3180A, AND 3181A, STEAM GEN 1A/1B/1C LOWER BLOWDOWN,

2) place R-19 Operations Selector Switch to the RESET position

52. Unit 1 is operating at 100% power.

Which one of the following meets the MINIMUM reactor coolant leakage detection system(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)

A. R-11 ONLY

() R-11 AND R-12 B.

C. The CACCLMS ONLY D. R-12 AND the CACCLMS

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 completes the statements below?

The minimum GFFD reading that would cause FG5 to come into alarm is (1) above background.

At 1020, flow through the GFFD (2) be isolated.

(1) (2)

A. 1 X 104 cpm will NOT B. 1 X 104 cpm WILL C. 1 X 105 cpm will NOT D. 1 X 105 cpm WILL

54. Unit 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 to isolate 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 TRN A. MOVs 514 and 517 B. MOVs 514 and 516 C. MOVs 515 and 517 D. MOVs 515 and 516

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

1A Containment Cooler is isolated per SOP-12.1, Containment 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 containment pressure rises to 5 psig.

Which one of the following completes the statement below?

1A Containment Cooler service water flow will be .

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

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 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) .

A. 1) is required to be secured and aligned for AUTO START

2) because this places the 1B DG is in the most reliable condition B. 1) is required to be secured and aligned for AUTO START

2) because the LOSP Sequencer will not run LOSP loads if DG01 opens with 1B DG output breaker aligned to 1G 4160V bus C. 1) load is required to be raised to FULL LOAD and DG01 opened

2) to ensure adequate voltage for safety related equipment D. 1) load is required to be raised to FULL LOAD and DG01 opened

2) to prevent extended low load operation which would result in the buildup of combustion products in the engine exhausts

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 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)

A. OPEN OPEN B. OPEN CLOSED C. CLOSED OPEN D. CLOSED CLOSED

58. The 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) .

A. 1) IS

2) a High Steam Flow - Lo Lo Tavg Main Steam Isolation may occur B. 1) IS

2) Technical Specification cooldown limits may be exceeded C. 1) is NOT

2) a High Steam Flow - Lo Lo Tavg Main Steam Isolation may occur D. 1) is NOT

2) Technical Specification cooldown limits may be exceeded

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.

A. 1) 1 additional Containment pressure channel reaching 16.2 psig

2) TWO B. 1) 2 additional Containment pressure channels reaching 16.2 psig

2) ONE C. 1) 1 additional Containment pressure channel reaching 27 psig

2) TWO D. 1) 2 additional Containment pressure channels reaching 27 psig

2) ONE

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 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 to correct the 1000 log entry by which one of the following methods?

A.

Circle the incorrect entry in red.

Enter the correct information next to the incorrect information and record the date and initial.

B.

Circle the incorrect entry in red.

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

C.

Draw a single line through the incorrect entry.

Enter the correct information next to the incorrect information and record the date and initial.

D.

Draw a single line through the incorrect entry.

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

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. Note: The FPA is the designated Shift Communicator.

Which one of the following completes 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 Plant Operator may work in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is (2) per NMP-AD-016-003, Scheduling and Calculating Work Hours.

(1) (2)

A. 3 12 B. 3 16 C. 4 12 D. 4 16

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?

A. Obtain Shift Manager's permission, THEN take the handswitch to START a second time.

B. Take the handswitch to START a second time, THEN write a Condition Report to document the action.

C. Write a condition report documenting the event and contact Maintenance.

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

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

A. Protects the pressurizer safety valves against water relief.

B. Provides the primary protection for preventing RCS over pressurization.

C. Protects against loss of pressure control due to spray nozzle being submerged.

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

64. Unit 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) .

A. 1) > 100 mrem/hr

2) Health Physics Supervision B. 1) > 100 mrem/hr

2) the Shift Support Supervisor (SSS)

C. 1) > 1000 mrem/hr

2) Health Physics Supervision D. 1) > 1000 mrem/hr

2) the Shift Support Supervisor (SSS)

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 Control, the operator (1) required to log in on the normal Auxiliary Building Access Control System (ACS) terminal prior to entering the Dry Cask Storage Area RCA.

Upon exiting from the Dry Cask RCA, the operator is required to perform a 2 minute frisk and also use the (2) .

A. 1) IS

2) Primary Access Point (PAP) exit portal monitors B. 1) IS

2) Auxiliary Building RCA exit portal monitors C. 1) is NOT

2) Primary Access Point (PAP) exit portal monitors D. 1) is NOT

2) Auxiliary Building RCA exit portal monitors

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

An employee who is a fully documented radiation worker and DOES NOT declare her pregnancy has an annual FNP Administrative TEDE limit of .

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

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 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) .

A. 1) is NOT

2) ensure a secondary side heat sink B. 1) is NOT

2) prevent SG depressurization during the RCS cooldown C. 1) IS

2) ensure a secondary side heat sink D. 1) IS

2) prevent SG depressurization during the RCS cooldown

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

IF ECP-1.1, Loss of Emergency Coolant Recirculation, is in effect, THEN Containment Spray 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 .

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

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 alarms sounds and an announcement is made declaring 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) .

A. 1) after securing the #1 WMT release

2) Operations Support Center (OSC)

B. 1) after securing the #1 WMT release

2) Control Room C. 1) immediately

2) Operations Support Center (OSC)

D. 1) immediately

2) Control Room

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 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)

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

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 slowly Which one of the following completes the statement below?

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

A. 1) IS

2) 1B and 1C SG levels are sufficient B. 1) is NOT

2) Neither SG levels nor AFW flow capability is sufficient C. 1) IS

2) AFW flow capability is sufficient D. 1) is NOT

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

72. Unit 2 has experienced a large steam break inside containment and the following conditions 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 been entered 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 mitigation strategy of FRP-P.1?

The RCS cooldown must (1) .

An RCS pressure reduction (2) required.

(1) (2)

A. continue is NOT B. continue IS C. be stopped is NOT D. be stopped IS

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) that the applicable procedure(s) direct?

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

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

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

2) Transfer to Cold Leg recirculation ONLY.

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

2) Minimize HHSI flow to the minimum required to remove decay heat while attempting to restore at least one train of emergency recirculation.

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

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

74. The 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) .

A. 0 gpm to each SG to prevent excessive cooldown B. at least 20 gpm to each SG to prevent dryout of the SGs C. at least 20 gpm total AFW flow to minimize thermal stress to the SGs D. at least 395 gpm total AFW flow to maintain adequate heat sink

75. Which one of the following is the first Major Action Category in FRP-Z.2, Response To Containment Flooding, and reason for this in accordance with the background document?

A. Identify unexpected sources of water in the sump since flooding could damage critical plant equipment.

B. Evaluate the ECCS system status to determine a strategy to transition to simultaneous cold and hot leg recirculation.

C. Have chemistry evaluate sump level, chemistry, and activity level to determine a strategy to transfer excess water out of containment.

D. Notify the TSC of sump chemistry, and activity level to determine potential changes in the planned transition to simultaneous cold and hot leg recirculation.

76. Unit 1 is at 98% power.

The OATC pulls Control Rods 2 steps and the following indications occur:

Control Bank D Group Step Counters indicate 227 steps.

On DRPI, Control Bank D, rod P8 drops to 144 steps.

On DRPI, all other Control Bank D rods indicate 228 steps.

Other Control Room parameters indicate Control Rod P8 has dropped to 144 steps.

Which one of the following completes the statements below?

(1) , will come into alarm due to the rod malfunction.

Per the BASES of Tech Spec 3.1.4, Rod Group Alignment Limits, a power reduction to < 75% power is required to ensure (2) .

A. 1) FF5, COMP ALARM ROD SEQ/DEV OR PR FLUX TILT

2) total available rod worth is within safety analyses limits B. 1) FE2, CONT ROD BANK POSITION LO-LO

2) total available rod worth is within safety analyses limits C. 1) FF5, COMP ALARM ROD SEQ/DEV OR PR FLUX TILT

2) the local Linear Heat Rate increases will not exceed core design criteria D. 1) FE2, CONT ROD BANK POSITION LO-LO

2) the local Linear Heat Rate increases will not exceed core design criteria

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

AOP-16.0, CVCS Malfunction, is in progress due to a loss of Pressurizer level control.

At the step to "Check Pressurizer level", Pressurizer level is 76% and rising.

Which one of the following completes the statements below?

The Pressurizer is (1) per Tech Spec 3.4.9, Pressurizer.

Per AOP-16.0, the Control Room crew is required to (2) .

(1) (2)

A. NOT OPERABLE trip the Reactor and go to EEP-0.0, Reactor Trip or Safety Injection B. NOT OPERABLE reduce Reactor power/TAVG as necessary using UOP-3.1, Power Operation C. OPERABLE trip the Reactor and go to EEP-0.0, Reactor Trip or Safety Injection D. OPERABLE reduce Reactor power/TAVG as necessary using UOP-3.1, Power Operation

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

RCS pressure is 1800 psig.

An RCS heatup is in progress.

1A Accumulator has been declared INOPERABLE.

1A Accumulator pressure is 655 psig.

1A Accumulator level is 41%.

Per Tech Specs 3.5.1, Accumulators, which one of the following completes the statements below?

Restore 1A Accumulator to OPERABLE by (1) to within the Tech Spec required range.

Tech Spec 3.5.1 BASES states that (2) .

A. 1) lowering 1A Accumulator pressure

2) the maximum pressure limit prevents injecting nitrogen into the RCS during a LOCA B. 1) lowering 1A Accumulator pressure

2) the maximum pressure limit prevents the accumulator relief valve from actuating C. 1) raising 1A Accumulator level

2) the minimum level limit prevents injecting nitrogen into the RCS during a LOCA D. 1) raising 1A Accumulator level

2) the minimum level limit ensures peak clad temperature remains below 2200°F during a LOCA

79. Unit 1 is in Mode 2 at 3% power when the following occurs:

PCV-444B, PRZR PORV, fails open.

The OATC performs actions of AOP-100, Instrumentation Malfunction.

PCV-444B, PRZR PORV, handswitch RED light remains lit.

The OATC closes MOV-8000B, PRZR PORV ISO.

RCS pressure is rising.

Which one of the following completes the statements below per Tech Spec 3.4.11, Pressurizer Power Operated Relief Valves (PORVs)?

PRZR PORV ISO, MOV-8000B is required to be closed with power (1) .

Entry into Mode 1 (2) allowed by Tech Specs.

(1) (2)

A. removed is NOT B. maintained is NOT C. removed IS D. maintained IS

80. Unit 1 is in MODE 4, with the following plant conditions:

LCV-115B, RWST TO CHG PUMP, will not open due to a malfunction within the Limitorque Motor.

1B Charging pump is aligned to B train.

1B BAT is on service.

1A BAT is on standby.

The 1B DG is declared INOPERABLE.

Which one of the following describes the impact of these conditions on the number of FUNCTIONAL Boration Flowpaths, per TRM 13.1.3, Boration Flow Paths-Operating?

There is(are) FUNCTIONAL boration Flowpath(s) .

A. NO B. ONLY one C. ONLY two D. three

81. Unit 2 is in Mode 5 with the following conditions:

CTMT Purge was secured for a Local Leak Rate Test (LLRT) and the LLRT Tagout has been cleared.

R-24A, Containment Radiation Monitor, has been declared INOPERABLE.

Prior to placing the CTMT Purge system in operation per SOP-12.2, Containment Purge and Pre-Access Filtration System, which one of the following requirements must be met?

A. R-24A is required to be restored to OPERABLE status.

B. Gaseous Release Permits are required to be issued.

C. Alternate sampling is required to be performed per the ODCM.

D. CTMT Pre-Access Filtration must be placed in service prior to placing CTMT Purge in service.

82. The following conditions exist on Unit 1:

At 1000:

A Reactor Trip was initiated, but was unsuccessful.

The crew entered FRP-S.1, Response to Nuclear Power Generation/ATWT.

The Rover was dispatched to open Reactor Trip breakers.

At 1005:

Per FRP-S.1, an Emergency Boration is in progress from the 1A BAT (Boric Acid Tank).

FK-122, CHG FLOW, is in AUTO.

FI-122A, CHG FLOW, indicates 35 gpm.

All Reactor Trip and Reactor Trip bypass breakers are open.

The Shift Manager is evaluating emergency classifications.

Which one of the following completes the statements below?

At 1005, Charging flow (1) adequate for the Emergency Boration.

The Shift Manager is required to declare a(n) (2) emergency classification.

Reference Provided (1) (2)

A. IS Alert B. is NOT Alert C. IS Site Area D. is NOT Site Area

83. Unit 1 is at 28% power with the following conditions:

At 1000:

PR-4029, CONDENSER PRESSURE, indicates as follows:

- PT0501 is 1.7 psia and slowly rising.

- PT0502 is 1.7 psia and slowly rising.

KK1, TURB COND VAC LO, is in alarm.

HP Gland steam supply pressure is 3.5 psig.

Main Turbine LP Gland pressures are as follows:

- #3 LP Gland pressure is 2.0 psig.

- #4 LP Gland pressure is 4.8 psig.

- #5 LP Gland pressure is 2.5 psig.

- #6 LP Gland pressure is 0.5 psig.

At 1015:

PR-4029, CONDENSER PRESSURE, indicates as follows:

- PT0501 is 2.1 psia and slowly rising.

- PT0502 is 2.1 psia and slowly rising.

KK2, TURB COND VAC LO-LO, is in alarm.

Which one of the following completes the statements below per AOP-8.0?

The action required at 1000 to stabilize Condenser pressure is to (1) .

The action required at 1015 is to (2) .

Procedure titles are as follows:

AOP-3.0, Turbine Trip Below P-9 Setpoint AOP-8.0, Partial Loss of Condenser Vacuum A. 1) bypass the #6 Gland Seal Regulator

2) perform AOP-3.0 in parallel with AOP-8.0 B. 1) bypass the #6 Gland Seal Regulator

2) perform AOP-3.0 ONLY C. 1) throttle closed V528, SPILLOVER VALVE BYPASS

2) perform AOP-3.0 in parallel with AOP-8.0 D. 1) throttle closed V528, SPILLOVER VALVE BYPASS

2) perform AOP-3.0 ONLY

84. Unit 1 is at 37% power with the following conditions:

FCV-488, 1B SG FW FLOW, fails open and sticks open.

Given the following Tech Spec title:

3.7.3, Main Feedwater Stop Valves and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves With NO OPERATOR ACTIONS, which one of the following completes the statements below?

A Reactor Trip (1) occur.

Per the BASES of Tech Spec 3.7.3, the primary reason for automatic closure of the Main Feedwater Regulating valves is to prevent (2) .

(1) (2)

A. will NOT an excessive mass addition to the Steam Generator during a Steamline Break event B. will NOT a radiological release through the feedline penetration during a Steam Generator Tube Rupture event C. WILL an excessive mass addition to the Steam Generator during a Steamline Break event D. WILL a radiological release through the feedline penetration during a Steam Generator Tube Rupture event

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

A Grid disturbance caused a loss of all offsite power.

During the event, the 1-2A DG started and then tripped.

Operating DG status is as follows:

- 1C DG is supplying Unit 1.

- 1B DG is supplying Unit 1.

- 2B DG is supplying Unit 2.

ACC reports that the following lines have been restored:

- Webb 230 KV line

- Snowdoun 500KV line

#1 Auto Bank Transformer is out of service.

Which one of the following completes the statements below for an evaluation of Unit 1 Tech Specs for the current conditions?

Perform REQUIRED ACTION of Tech Spec 3.8.1, AC Sources -

Operating, (1) .

A 25% extension of the COMPLETION TIME (2) allowed for the initial performance of SR 3.8.1.1 (STP-27.1, A.C. Source Verification).

Reference Provided (1) (2)

A. CONDITION A IS B. CONDITION B IS C. CONDITION A is NOT D. CONDITION B is NOT

86. Unit 1 is at 100% power when the following conditions occur:

VC4, 1B BATT CHG FAULT OR DISC, alarms.

EEO5, 1B BATTERY CHARGER SUPPLY BREAKER, has tripped open.

1B DC Bus Voltage is 130 Volts.

Given the following Tech Spec titles:

3.8.4, DC Sources-Operating

3.8.9, Distribution Systems-Operating Which one of the following completes the statement below?

is(are) required to be implemented.

Reference Provided A. NEITHER Tech Spec 3.8.4 nor Tech Spec 3.8.9 B. ONLY Tech Spec 3.8.4 C. ONLY Tech Spec 3.8.9 D. BOTH Tech Spec 3.8.4 and Tech Spec 3.8.9

87. Unit 1 tripped from 100% power with the following conditions:

A Safety Injection occurred due to a Steam Dump malfunction.

Status of AFW pumps is as follows:

- 1A MDAFW pump is Tagged Out.

- 1B MDAFW pump is tripped.

- The TDAFW pump trip/throttle linkage is broken.

The Safety Injection signal has been reset.

FRP-H.1, Response to Loss of Secondary Heat Sink, is in progress with the crew attempting to restore SG level using a Condensate pump.

Which one of the following completes the statements below?

Implementation of FRP-H.1, Attachment 1, Main Feedwater Bypass Valves Automatic Closure Defeat, (1) required to open the Main Feedwater Bypass valves.

Criteria to exit FRP-H.1 is met (2) .

(1) (2)

A. IS ONLY after at least one SG NR level is > 31%

B. is NOT ONLY after at least one SG NR level is > 31%

C. IS when stable Condensate flow is established and SG wide range level is rising D. is NOT when stable Condensate flow is established and SG wide range level is rising

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

WE1, 1F, 1H, OR 1K 4KV BUS BKR AUTO TRIP, is in alarm.

WE3, 1H 4KV BUS UV OR LOSS OF DC, is in alarm.

The following indications are noted on the handswitch for 4160V Breaker DF-13-1, SUPPLY TO 1H 4160V BUS:

- The amber light is lit.

- The green light is lit.

The AC PWR AVAIL lights for 1F 4160V bus are illuminated.

The AC PWR AVAIL lights for 1H 4160V bus are NOT illuminated.

Which one of the following completes the statements below?

The 1C DG (1) autostart due to the DF-13-1 malfunction.

For Unit 1, a REQUIRED ACTION statement of Tech Spec 3.8.1, AC Sources - Operating, (2) required to be implemented.

(1) (2)

A. WILL is NOT B. WILL IS C. will NOT is NOT D. will NOT IS

89. Which one of the following describes TRM 13.12.4, Gas Storage Tanks, limitations on the quantity of radioactivity permitted in the Waste Gas Decay Tank(s) and the basis for the limit?

The quantity contained in each Waste Gas Decay Tank shall be < (1) to ensure an uncontrolled release will not exceed a whole body dose of 0.5 REM to (2) .

A. 1) 10 curies, excluding tritium and dissolved or entrained noble gases

2) an operator at the Waste Gas Control Panel B. 1) 70,500 curies of noble gases (considered as Xe-133)

2) an operator at the Waste Gas Control Panel C. 1) 10 curies, excluding tritium and dissolved or entrained noble gases

2) an individual at the exclusion area boundary D. 1) 70,500 curies of noble gases (considered as Xe-133)

2) an individual at the exclusion area boundary

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

At 1000:

#4 Waste Gas Decay Tank (WGDT) release is in progress.

CTMT Mini-purge supply and exhaust fans are running.

The Radwaste Ventilation system is running.

At 1015:

R-14, PLANT VENT, radiation monitor is oscillating erratically.

The HIGH ALARM and LOW ALARM lights are LIT.

The Shift Supervisor has declared R-14 INOPERABLE.

Which one of the following completes the statements below?

The WGDT release (1) terminated.

With R-14 INOPERABLE, to continue all the releases that were in progress at 1000, the ODCM requires sampling by obtaining (2) .

REFERENCE PROVIDED (1) (2)

A. IS at least two independent samples B. IS grab samples at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C. is NOT at least two independent samples D. is NOT grab samples at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

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

Several substations are separated from the grid resulting in the following plant conditions:

Unit 1 Generator Voltage is 20.45 kV.

The following alarms have actuated:

WE2, 1F 4KV BUS OV-OR-UV OR LOSS OF DC.

VE2, 1G 4KV BUS OV-OR-UV OR LOSS OF DC.

Grid frequency has fallen to 59.6 hertz and is stable.

4160V Bus voltages are 3840 Volts.

This condition has existed for the past hour.

Which one of the following completes the statements below?

The Generator temperatures will (1) .

AOP-5.2, Degraded Grid, will require the crew to (2) .

(1) (2)

A. RISE immediately enter AOP-17.1, Rapid Turbine Power Reduction B. LOWER immediately enter AOP-17.1, Rapid Turbine Power Reduction C. RISE place the unit in mode 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months using UOP-3.1, POWER OPERATION D. LOWER place the unit in mode 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months using UOP-3.1, POWER OPERATION

92. Unit 1 had an extended Loss of all AC power with the following conditions:

At 1000:

Power has been restored to the 1F and 1K 4160V busses.

The Shift Supervisor is preparing to exit ECP-0.0, Loss of ALL AC Power, and is at the step to "Evaluate plant conditions".

SCMM is 39°F.

Pressurizer level is 16% and slowly falling.

CTMT pressure is 5.8 psig.

SI equipment did NOT automatically actuate when power was restored.

At 1002:

ECP-0.0 has been exited and the applicable recovery procedure has been entered.

The Shift Supervisor is informed that a RED path exists for Heat Sink on the IPC CSF status trees.

Which one of the following completes the statements below?

At 1000, entry into (1) is required.

At 1002, per the applicable procedure in effect, the Control Room crew (2) required to enter FRP-H.1, Loss of Secondary Heat Sink.

Procedure titles are as follows:

ECP-0.1, Loss of All AC Power Recovery Without SI Required.

ECP-0.2, Loss of All AC Power Recovery With SI Required.

(1) (2)

A. ECP-0.1 IS B. ECP-0.1 is NOT C. ECP-0.2 IS D. ECP-0.2 is NOT

93. Unit 1 is at 75% power and ramping up with the following conditions:

A reactivity plan has been developed and approved for the reactivity additions.

A Reactivity Management SRO has been stationed due to frequent reactivity additions.

Which one of the following completes the statements below per NMP-OS-001, Reactivity Management Program?

During the ramp up, the limit for Control Rod withdrawal is a maximum of (1) per rod pull.

The Reactivity Management SRO (2) authorize changes to the reactivity plan.

(1) (2)

A. three steps CANNOT B. three steps CAN C. four steps CANNOT D. four steps CAN

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

The packing was replaced on MOV-8812B, CTMT SUMP TO 1B RHR PUMP.

The MOV was left in the closed position.

The Return to Service Tagout is in progress and power has been restored to the MOV.

Which one of the following states the minimum action(s) required to restore MOV-8812B to OPERABLE after the valve packing replacement?

MOV-8812B is OPERABLE when .

A. the MOV's auto open function is tested per its Surveillance Test Procedure B. the MOV has been satisfactorily time stroked per its Surveillance Test Procedure C. the MOV is stroked open and closed from its remote handswitch per the guidance of the Tagout D. the MOV is manually stroked open and closed per the guidance of the Tagout, with no leakage verified

95. Unit 2 is at 100% power with the following conditions:

The 2B RHR pump is being Tagged Out for a scheduled oil change and maintenance inspection.

Maintenance is expected to take 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to complete.

Which one of the following completes the statements below?

The LCO required is a (1) LCO.

In addition to a Control Room LCO Log entry, an LCO/TR Status Sheet (2) required to document the LCO.

(1) (2)

A. Voluntary IS B. Voluntary is NOT C. Mandatory IS D. Mandatory is NOT

96. Unit 1 is shutdown with the following conditions:

STP-18.4, Containment Mid-Loop and/or Refueling Integrity Verification and Containment Closure, has been completed to allow refueling operations.

The Containment Equipment Hatch is open.

During fuel movement, a spent fuel assembly has dropped from the Manipulator Crane to the bottom of the cavity.

Both R-24A & B, CTMT PURGE, are in alarm.

AOP-30, Refueling Accident, actions are in progress.

Which one of the following completes the statements below?

The ACCEPTANCE CRITERIA of STP-18.4 for Refueling Integrity states that the Containment Equipment Hatch is capable of being closed within (1) of notification.

Per AOP-30, actions are required to place (2) in service without delay.

(1) (2)

A. four hours PRF (Penetration Room Filtration)

B. four hours CREFS (Control Room Emergency Filtration System)

C. two hours PRF (Penetration Room Filtration)

D. two hours CREFS (Control Room Emergency Filtration System)

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

A release of the #2 Waste Monitor Tank (WMT) is planned.

R-18, LIQ WASTE DISCH, is INOPERABLE.

Chemistry has taken two independent samples of the #2 WMT and reports the activity is <1.4 x10-5 µCi/ml and is within the normal limits for a release.

Two Shift Radio-Chemists have verified the manual input for the computer generated release rate calculation.

Which one of the following completes the statement below?

Per the ODCM, a WMT release .

A. is NOT permitted until the activity is lowered to <1x10-7 µCi/ml B. is NOT permitted with R-18 INOPERABLE, but the tank can be transferred to Unit 2 for release C. IS permitted, but as a minimum two qualified plant personnel are required to verify the discharge lineup only D. IS permitted, but as a minimum two qualified plant personnel are required to verify the discharge lineup and an SRO is required to verify the entire release rate calculation

98. Unit 1 tripped from 100% power with the following conditions:

A Safety Injection was actuated due to a SG Tube Rupture on the 1B SG.

The crew is performing EEP-3.0, Steam Generator Tube Rupture.

The following results are reported:

HV3369B, 1B SG MSIV, will not close.

HV3370B, 1B SG MSIV, will not close.

1B SG pressure is 950 psig.

Which one of the following completes the statements below?

The Control Room crew is required to (1) .

The recovery strategy for the selected procedure is to (2) .

Procedure titles are as follows:

EEP-3.0, Steam Generator Tube Rupture ECP-3.1, SGTR with Loss of Reactor Coolant Subcooled Recovery Desired (1) (2)

A. continue in EEP-3.0 complete an RCS cooldown first; then perform an RCS depressurization B. transition to ECP-3.1 complete an RCS cooldown first; then perform an RCS depressurization C. continue in EEP-3.0 start an RCS cooldown and then perform an RCS depressurization during the cooldown D. transition to ECP-3.1 start an RCS cooldown and then perform an RCS depressurization during the cooldown

99. Unit 1 has tripped from 100% power with the following conditions:

A Safety Injection occurred after the Reactor Trip due to a Steam Dump malfunction.

All MSIV's are closed.

The crew is evaluating SI termination criteria in EEP-0.0, Reactor Trip or Safety Injection.

The SCMMs indicate 105°F in CETC mode.

All SG Narrow Range levels are 43% and rising.

RCS pressure is 1820 psig and slowly rising.

Pressurizer level is 5% and rising.

Which one of the following describes the required actions regarding SI Termination?

A. Immediately transition to ESP-1.1, SI Termination, to terminate SI flow.

B. Maintain SI flow until PZR level recovers, then transition to ESP-1.2, Post LOCA Cooldown and Depressurization.

C. Maintain SI flow until PZR level recovers, then terminate SI flow in EEP-0.0, Reactor Trip or Safety Injection.

D. Immediately terminate SI flow in EEP-0, Reactor Trip or Safety Injection, then transition to ESP-1.1, SI Termination.

100. A LOCA has occurred on Unit 1 with the following conditions:

There has been no HHSI or LHSI flow for an extended period of time.

ESP-1.2, Post LOCA Cooldown and Depressurization, was in progress when conditions required a transition to FRP-C.2, Response to Degraded Core Cooling.

A SG depressurization is in progress per FRP-C.2.

Subsequently, core cooling is re-established:

1A RHR pump is started.

LHSI flow is 1800 gpm.

Which one of the following completes the statements below?

The SG depressurization is performed at (1) per FRP-C.2.

Once the SG depressurization is complete and core cooling is re-established, a transition will be made to (2) .

A. 1) a maximum attainable rate

2) EEP-1.0, Loss of Reactor or Secondary Coolant B. 1) <100°F/hour cooldown rate

2) EEP-1.0, Loss of Reactor or Secondary Coolant C. 1) a maximum attainable rate

2) ESP-1.2, Post LOCA Cooldown and Depressurization D. 1) <100°F/hour cooldown rate

2) ESP-1.2, Post LOCA Cooldown and Depressurization

REFERENCES 10/18/12 9:25:43 FNP-1-AOP-2.0 UNIT 1 STEAM GENERATOR TUBE LEAKAGE Version 35.0 Step Action/Expected Response Response Not Obtained 33.2 34.2 [CA] WHEN one of the following 34.2 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 in CETC mode.

33.3 34.3 Verify both normal pressurizer spray 34.3 Perform the following.

valves - CLOSED.

33.3.1 34.3.1 Deenergize rod control system.

1A(1B) LOOP

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

[ ] PK 444D OR

[ ] Open 1A and 1B MG SET SUPP BKRs.

33.3.2 34.3.2 Stop associated 1A & 1B RCPs.

33.3.3 34.3.3 IF any rod bottom light not lit, THEN emergency borate RCS using FNP-1-AOP-27.0, EMERGENCY BORATION.

33.4q 34.4 Verify auxiliary spray valve - CLOSED. 34.4 Isolate auxiliary spray line.

RCS PRZR CHG PMPS TO AUX SPRAY REGENERATIVE HX

[ ] Q1E21HV8145 [ ] Q1E21MOV8107 closed

[ ] Q1E21MOV8108 closed 33.5 34.5 Verify both PRZR PORVs - CLOSED. 34.5 Close PRZR PORV ISO for any open PRZR PORV.

Step 38 continued on next page 38 ProcedureStepsMain Page 35 of 39

AC Sources Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and

b. Two diesel generator (DG) sets capable of supplying the onsite Class 1E power distribution subsystem(s); and

c. Automatic load sequencers for Train A and Train B.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

NOTE----------------------------------------------------------

LCO 3.0.4b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit A.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable. required OPERABLE offsite circuit. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no discovery of no offsite offsite power available power to one train inoperable when its concurrent with redundant required inoperability of feature(s) is inoperable. redundant required feature(s)

AND (continued)

Farley Units 1 and 2 3.8.1-1 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months circuit to OPERABLE status. AND 13 days from discovery of failure to meet LCO B. One DG set inoperable. ------------------NOTE-------------------

LCO 3.0.4c is applicable when only one of the three DGs is inoperable.

B.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the required offsite circuit(s). AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s) supported by discovery of the inoperable DG set Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is inoperable. redundant required feature(s)

AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG set is not inoperable due to common cause failure.

OR (continued)

Farley Units 1 and 2 3.8.1-2 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3.2 Perform SR 3.8.1.6 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE DG set.

AND B.4 Restore DG set to 10 days OPERABLE status.

AND 13 days from discovery of failure to meet LCO C. Two required offsite C.1 Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable. feature(s) inoperable discovery of when its redundant Condition C required feature(s) is concurrent with inoperable. inoperability of redundant required features AND C.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

Farley Units 1 and 2 3.8.1-3 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

DC Sources Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources Operating LCO 3.8.4 The Train A and Train B Auxiliary Building and Service Water Intake Structure (SWIS) DC electrical power subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Auxiliary Building DC A.1 Restore the Auxiliary 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power subsystem Building DC electrical inoperable. power subsystem to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for 1B OPERABLE status. Auxiliary Building DC electrical power subsystem inoperable due to inoperable battery for cycle 19 only B. One Auxiliary Building DC B.1 Restore the battery 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months electrical power subsystem connection resistance to with battery connection within limit.

resistance not within limit.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. One required SWIS DC D.1 Restore the battery 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months electrical power connection resistance to subsystem battery within the limit.

connection resistance not within limit.

E. One required SWIS DC E.1 Declare the associated Immediately electrical power Service Water System subsystem inoperable. train inoperable.

OR Required Action and associated Completion Time of Condition D not met.

Farley Units 1 and 2 3.8.4-1 Amendment No. 164 (Unit 1)

Amendment No. 137 (Unit 2)

Distribution Systems Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems Operating LCO 3.8.9 Train A and Train B AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more AC electrical A.1 Restore AC electrical 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months power distribution power distribution subsystems inoperable. subsystem(s) to AND OPERABLE status.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO B. One or more AC vital B.1 Restore AC vital bus 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months buses inoperable. subsystem(s) to OPERABLE status. AND 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO C. One Auxiliary Building DC C.1 Restore Auxiliary 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power distribution Building DC electrical subsystem inoperable. power distribution AND subsystem to OPERABLE status. 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO Farley Units 1 and 2 3.8.9-1 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

Distribution Systems Operating 3.8.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, or AND C not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. One Service Water Intake E.1 Declare the associated Immediately Structure (SWIS) DC Service Water train electrical power distribution inoperable.

subsystem inoperable.

F. Two trains with inoperable F.1 Enter LCO 3.0.3. Immediately distribution subsystems that result in a loss of safety function.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to In accordance with required AC, DC, and AC vital bus electrical power the Surveillance distribution subsystems. Frequency Control Program Farley Units 1 and 2 3.8.9-2 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

FNP-ODCM Table 3-1 Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirementsb Minimum Channels Instrument OPERABLE Applicability ACTION

1. Steam Jet Air Ejector Noble Gas Activity Monitor (RE-15) 1 MODES 1,2,3,4 37

2. Plant Vent Stack

a. Noble Gas Activity Monitor (RE-14 or RE-22) 1 At all times 37a

b. Iodine Sampler 1 At all times 39 c Particulate Sampler 1 At all times 39

d. Flowrate Monitor 1 At all times 36

3. GASEOUS RADWASTE TREATMENT SYSTEM Noble Gas Activity Monitor (RE-14), with Alarm and Automatic Termination of Release 1 At all times 35

a. For continuous releases.

b. All requirements in this table apply to each unit.

3-3 Version 24 01/10

FNP-ODCM Table 3-1 (contd) Notation for Table 3 ACTION Statements ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and (1) Verify the manual portion of the computer input for the release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 3-3.

3-4 Version 24 01/10

ANSWER KEY 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: 06/25/13 Facility/Unit: FARLEY Unit 1 & 2 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete the combined examination, and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months if you are only taking the SRO portion.

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

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

1. 001AA2.01 001 Unit 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 the Channel 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 Bypass Breaker (2) light will be LIT.

(1) (2)

A. insert RED B. insert GREEN C. withdraw GREEN D. withdraw RED

PT-447 failing low will cause rods to insert continuously and failing high will cause rods to withdraw.

Per FSD-A181007, Figure 2 Sheet 2, a manual reactor trip will open the 'B' Reactor Trip Bypass Breaker.

Distracter Analysis A. Incorrect. First part is incorrect (See C.1). Plausible if the applicant reverses the system response so that the rod control system would insert rods to suppress the indicated power rise due to the failure instead of trying to adjust actual reactor power and Tavg to match the failed indication.

Second part is incorrect (See C.2). Plausible if the applicant believes that during the performance of STP-33.0B, the 'B' Reactor Trip 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 calculate Tref. Since PT-447 failed high, Tref fails to the 100% power Tref and the rod control system will step rods out in order to raise current Tavg to match Tref. Impulse pressure is also used to determine turbine power as compared to reactor power in the rod control circuitry. When PT-447 fails high, the rapid rate of change of impulse power as compared to reactor power will also cause rods to step out while impulse pressure is changing.

This scenario has been run on desktop simulator and the rods will step out (IC 058).

Second part is correct. FSD A18007, Figure 2 sheet 2, shows that a manual trip actuation will open the 'B' Reactor Trip Bypass Breaker.

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 and interpret the following as they apply to the Continuous Rod Withdrawal : Reactor tripped breaker indicator Importance 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 12 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if entry into AOP-100, Instrument Malfunction is required.

(OPS-52521Q02)

ANALYZE plant conditions and DETERMINE the successful completion 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 a continuous rod withdrawal is occurring due to the failure of PT-447. After the reactor is tripped, the applicant is then required to interpret the reactor trip bypass breaker indication as to whether or not it is open.

SRO justification: N/A

2. 003A3.05 002 Unit 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, Reactor Coolant System?

The 1C RCP oil lift pump handswitch white light indicates (1) .

The 1C RCP breaker closing operation (2) interlocked with a 2 minute time delay.

A. 1) BOTH the oil lift pressure has reached 600 psig AND 2 minutes have elapsed

2) is NOT B. 1) ONLY that the oil lift pressure has reached 600 psig

2) IS C. 1) ONLY that the oil lift pressure has reached 600 psig

2) is NOT D. 1) BOTH the oil lift pressure has reached 600 psig AND 2 minutes have elapsed

2) IS

SOP-1.1 3.6 DO NOT attempt to start a RCP unless its oil lift pump has been delivering oil to the upper thrust shoes for at least two minutes. Observe the oil lift pumps indicating lights to verify correct oil pump motor operation and oil pressure. The oil lift pumps should run at least 1 minute after the RCP's are started. An interlock will prevent 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 minutes prior 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 Analysis A. Incorrect. First part is incorrect (See C.1). Plausible since the procedure requires both conditions to be met before starting the RCP. The applicant could have the misconception that the light is associated with 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 thinks that the procedural requirement of allowing the oil lift pump to run for 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 600 psig for the white light to come on. This may occur in as little as 1 minute or as long as 6 minutes, but the white light is not time dependent, it is only pressure dependent.

Second part is correct. The RCP breaker interlock is with the oil pressure 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 monitor automatic operation of the RCPS, including: RCP lube oil and bearing lift pumps Importance Rating: 2.7* 2.6 Technical

Reference:

FNP-1-SOP-1.1, Reactor Coolant System, Ver 47.2 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Reactor Coolant Pumps, to include the following (OPS-40301D02):

Oil lift system Question History: MOD VOGTLE 12 K/A match: The only AUTO features of the RCP LO and Brg lift pump is the white light will come on when the pressure reaches 600 psig and then the permissive will clear at 600 psig allowing the RCP breaker to be closed. Applicant must be able to monitor the white indicating light for the RCP oil lift pump which indicates the discharge pressure of that pump is > 600 psig and have knowledge that when the light comes on, the permissive automatically allows the RCP circuit breaker being capable of closing due to oil pressure and is time.

SRO justification: N/A

3. 004K3.08 003 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 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)

A. CLOSED LOWER B. CLOSED RISE C. OPEN LOWER D. OPEN RISE

AOP-6.0 Table 1:

Component No. Name Failed Position Q1E21V347 (1-CVC-FCV-122) CHG FLOW REG OPEN ARP-1.4, DC4 - SEAL WTR INJ FLTR HI P Probable Cause.

2. High seal injection flow rate.

3. Chg Flow Q1E21FCV122. (Also labeled Q1E21V347) failed closed.

Distracter Analysis A. Incorrect. First part is incorrect (See C.1). Plausible if the applicant does recall the fail position of FCV-122.

Second part is correct (See C.2). Logical connection to the first part if the applicant believes that FCV-122 is upstream of the seal injection line and the closure of FCV-122 would stop seal injection flow.

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

Second part is incorrect (See C.2). Logical connection to the first part if the applicant thought that FCV-122 failed closed since it would 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 high seal injection flow and thus high filter DP due to all of the charging pump discharge flow being directed to the seal injection filter (See ARP-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 first part if the applicant assumes that more charging flow equates to more seal injection flow.

K/A: 004K3.08 Chemical and Volume Control System (CVCS) - Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: RCP seal injection Importance 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 53 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Chemical and Volume Control System, to include the components found on Figure 3, Chemical and Volume Control System and 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 has been severed. Applicant must know how this CVCS malfunction affects seal injection flow.

SRO justification: N/A

4. 005A2.02 004 The 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 be performed per AOP-12.0, Residual Heat Removal Malfunction?

A. No actions are required.

B. Secure the 1A RHR pump ONLY.

C. Secure the 1B RHR pump ONLY.

D. Secure BOTH 1A AND 1B RHR pumps.

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 PT403 for 8702A and B to prevent isolation of both trains of RHR due to a single pressure transmitter failing high.

AOP-12:

Entry Conditions 1.4 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 3.4.12 Technical Specification 3.4.12, LTOP (Low Temperature Over Pressure Protection System requires two RHR suction relief valve with setpoints < 450 psig when the temperature of one or more RCS cold legs is < 325°F. If one or more of the RHR Loop suction valves closed, then this Technical Specification would not be met and Low Temperature Over Pressure Protection would not be satisfied.

Distracter analysis A. Incorrect. See B. Plausible if applicant remembers that the RHR Loop suctions are opened and de-energized at some point but cannot recall that it is when RCS temperature is less than 180°F. If the valves were de-energized they would not shut so no action would be required. Also, the applicant could remember that PT-402 and 403 provide interlocks to OPEN the RHR loop suctions but NOT remember they also will close the valves on high pressure. This would make this a correct answer.

B. Correct. PT-402 failing high will close MOV-8701A and MOV-8701B which isolates the suction to the 1A RHR pump. AOP-12 requires the 1A RHR pump to be secured.

C. Incorrect. See B. Plausible if the applicant improperly believes that PT-402 affects the 1B RHR pump suction valves instead of 1A RHR pump suction valves. If PT-403 failed high, this would be the correct answer.

D. Incorrect. See B. Plausible since the RHR loop suction valves have interlocks to prevent opening them if certain parameters are not met (See reference material FSD A181002). PT-402 must be less than 402.5 psig in order to open MOV-8701A and MOV-8702A which are on OPPOSITE trains. If the applicant thought that the closing on high pressure works the same way as the opening interlocks this would be a correct answer since they would believe a suction valve in each train will close and both RHR pumps would be required to be secured.

K/A: 005A2.02 Residual Heat Removal System (RHRS) - Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Pressure transient protection during cold shutdown Importance Rating: 3.5 3.7 Technical

Reference:

FSD-A181002, Residual Heat Removal, Ver 44 FNP-1-AOP-12, Residual Heat Removal Malfunction, Ver 25 FNP Technical Specifications, Ver 190 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if entry into AOP-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 loop pressure transmitter affects the RHR system in that it removes one of the required Low Temperature Over Pressure Protection System reliefs from service. The applicant will apply that prediction to AOP-12 to get to the action required to mitigate this malfunction.

SRO justification: N/A

5. 006K5.06 005 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 A. Stable and on-scale Zero B. Dropping or off-scale low Zero C. Dropping or off-scale low Rising D. Stable and on-scale Rising Desktop simulator IC-73 (100% power) - 10,000 gpm LOCA produced 477 psig in the RCS 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 Pumps should be stopped.

9.1 Check RCS pressure - GREATER THAN 275 psig {435 psig}

EEB-1 ERP Step Text: Check if LHSI Pumps should be stopped.

Purpose:

To stop the low-head SI pumps if RCS pressure is above their shutoff head to prevent damage to the pumps Basis: Upon safety injection initiation all safeguard pumps are started regardless of the possibility of high RCS pressure with respect to the low-head safety injection pump shutoff head. On low-head systems where the pump recirculates on a small volume circuit there is concern for pump and motor overheating. Shutdown of the pump and placement in the standby mode, when the RCS pressure meets the criteria outlined in this step, allows for future pump operability. If SI has not been previously reset and the low-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 safeguards components to rapidly inject [...] whenever the RCS pressure decreases below the tank cover gas pressure of 601 - 649 psig due to a loss of coolant accident Distracter Analysis A. Incorrect. First part is incorrect (See B.1). Plausible if the applicant is unfamiliar with the injection pressures of the accumulators. The

applicant may believe that RCS pressure is high enough to prevent accumulator injection.

Second part is correct (See B.2). Logical connection to the first part if the applicant recognizes that RHR injects at a lower pressure than the accumulators.

B. Correct. First part is correct. FSD-A181009, 3.3.13. Safety injection accumulators shall function as passive safeguards components to rapidly inject [...] whenever the RCS pressure decreases below the tank cover gas pressure of 601- 649 psig due to a loss of coolant accident.

Second part is correct. At 500 psig in the RCS, the accumulators will have injected but the RCS pressure will be above the RHR shut off head pressure. EEP-1 uses 435 psig (Adverse, because containment pressure is > 4 psig) as the criteria for RHR pump shut 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 first part if the applicant is unfamiliar with the injection pressures of the RHR pumps. If the applicant knows that the accumulators are(have) injecting(ed), they may also assume that RCS pressure is 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 first part if the applicant is unfamiliar with the injection pressures of the RHR pumps. The applicant may believe that the RHR pumps inject before the accumulators.

The K/A: 006K5.06 Emergency Core Cooling System (ECCS) - Knowledge of the operational implications of the following concepts as they apply to ECCS: Relationship between ECCS flow and RCS pressure Importance 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 4 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Emergency Core Cooling System, to include the components found on Figure 2, Accumulators, Figure 3, Refueling Water Storage Tank, and Figure 4, Emergency Core Cooling System (OPS-40302C02).

Question History: SUMMER 11 K/A match: The applicant is required to know which ECCS components are injecting into the core (flow) based on RCS pressure.

SRO justification: N/A

6. 006K6.18 006 Unit 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)

A. be affected NOT be affected B. NOT be affected be affected C. be affected be affected D. NOT be affected NOT be affected

SOP-68:

3.2 The normal display mode for the SMM is the CETC mode. This displays the margin to saturation (°F) using the highest core exit thermocouple (excluding upper head) and the lowest pressure. The RTD mode displays the margin to saturation (°F) using the hottest reactor coolant system (RCS) RTD (Th or Tc) and the lowest pressure. The pressure inputs are from PT-402 and 403 and from PT-455 for A-train and PT-457 for B-train.

Distracter Analysis A. Incorrect. First part is incorrect (See B.1). Plausible if the applicant thinks that PT-457 inputs to 'A" train SMM.

Second part is incorrect (See B.2). Logical connection to the first part 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 is unaffected.

Second part is correct. PT-457 inputs to 'B' Train and the SMM uses 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 thinks that 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 common misconception.

Second part is correct (See B.2) A logical connection to the first part if the applicant thinks that PT-455 and 457 input to both trains of 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 that PT-455 inputs to 'A' Train and PT-456 inputs to 'B' Train instead of PT-457 which would make this a correct answer.

Second part is incorrect (See D.1). Logical connection to the first part based on D.1 discussion.

K/A: 006K6.18 Emergency Core Cooling System (ECCS) - Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Subcooling margin indicators Importance Rating: 3.6 3.9 Technical

Reference:

FNP-1-SOP-68.0, Inadequate Core Cooling Monitoring System, Ver 8.1 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Inadequate 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 to evaluate plant conditions to determine if ECCS flow can be terminated. Part of this determination is evaluating subcooling. The applicant has to know the effect of the loss of PT-457 on the subcooling margin monitors in order to be able to provide the Shift Supervisor the correct subcooling value.

SRO justification: N/A

7. 007EA2.06 007 Unit 2 was operating at 2% power with a plant startup in progress per 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 completes 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)

A. OPEN WILL B. OPEN will NOT C. CLOSED WILL D. CLOSED will NOT

FSD-181007 Figure 2 Sheet 2 shows all signals that open the reactor trip breakers and none are present 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 analysis A. Incorrect. First part is incorrect (See D.1). Plausible since the applicant may believe that the loss of power to both trains of safety related power causes the reactor trip breakers to open.

Second part is correct (See D.2). Logical connection to the first part 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 applicant believes 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 2A CRDM 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) is normally 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 first part for two reasons. If the applicant thought there was no trip (rods did not fall into the core) this would be the correct conclusion. If the applicant knew that the trip breakers would not open they could still believe the plausibility of B.2.

K/A: 007EA2.06 Reactor Trip - Ability to determine or interpret the following as they apply to a reactor trip: Occurrence of a reactor trip Importance Rating: 4.3 4.5 Technical

Reference:

FNP-2-SOP-41.0, Control Rod Drive and Position Indication System, Ver 35.1 A351199, Unit 2 Electrical Load List, Ver 61 References provided: None Learning Objective: RECALL AND DESCRIBE the operation and function of the following reactor trip signals, permissives, control interlocks, and engineered safeguards actuation signals associated with the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) to include setpoint, coincidence, rate functions (if any), reset features, and the potential consequences for improper conditions to include those items 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 and determine if a reactor trip has occurred.

SRO justification: N/A

8. 007K5.02 008 Unit 1 is in Mode 5 and forming a pressurizer steam space (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 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)

A. charging flow remain constant B. letdown flow remain constant C. letdown flow rise D. charging flow rise

UOP-1.1:

5.11 WHEN pressurizer temperature increases to the saturation temperature for 375 psig (approximately 442°F) as indicated by increasing RCS pressure or letdown flow, 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 VLV Q1E21LCV115A in the fully diverted position.

Distracter analysis A. Incorrect. First part is incorrect (See B.1). Plausible if the applicant thinks that FCV-122 is in auto and will open to raise charging flow in response to the RCS pressure rise. FCV-122 operates in automatic based 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 temperature increases to the saturation temperature for 375 psig (approximately 442°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 applicant improperly 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 following concepts as the apply to PRTS: Method of forming a steam bubble in the PZR Importance Rating: 3.1 3.4 Technical

Reference:

FNP-1-UOP-1.1, Startup of Unit from Cold Shutdown to Hot Standby, Ver 94.3 References provided: None Learning Objective: DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the Pressurizer System components and equipment, to include the following (OPS-40301E07):

Normal Control Methods Question 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. 008AK1.01 009 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 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)

A. 546°F rising B. 546°F lowering C. 320°F lowering D. 320°F rising

WOG Executive Guideline - During situations where a steam vent path is established from 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 from steam generated in the reactor vessel, passing through the pressurizer surge line and preventing the water inventory of the pressurizer from draining into the RCS loops. This holdup of water can result in a stable or even increasing indicated pressurizer level while RCS water inventory is actually decreasing. Pressurizer level should be relied on only 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 analysis A. Incorrect. First part is incorrect (See D.1). Plausible if applicant believes that the temperature of the steam in the Pressurizer is the same temperature as the steam entering the PRT. 546°F is the approximate saturation temperature for 1035 psia. This was the error 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 the scenario 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 since during a LOCA event the normal response is that Pzr level decreases.

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 the pressurizer, the pressurizer level will be rising. This scenario was run on the desktop simulator. Pressurizer level was rising at 1020 psig.

K/A: 008AK1.01 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) - Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: Thermodynamics and flow characteristics of open or leaking valves Importance Rating: 3.2 3.7 Technical

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 System instrument/equipment response expected when performing Pressurizer System evolutions, including the Normal Condition, the Failed Condition, Associated Alarms, Associated Trip Setpoints, to include the components found on Figure 3, Pressurizer and Pressurizer Relief Tank (OPS-52101E07)

Question History: MOD HARRIS 09 K/A match: Applicant has to determine the safety valve tailpeice temperature using steam tables (Thermodynamics and flow characteristics) and pressurizer level trend (operational implications) during a stuck open PORV condition.

SRO justification: N/A

10. 008K2.02 010 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)

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

2C CCW pump is power from the 2F 4160V bus which is powered from the 1C DG in the above scenario.

Distracter analysis A. Incorrect. First part is incorrect (See D.1). Plausible since, with the exception of spent fuel pool cooling, every other train related pump with an 'A' designation is an 'A' train pump. The applicant may not recall that the CCW system is "backwards" Second part is incorrect (See C.2). Plausible because if there was an LOSP on Unit 2 only, the 1-2A DG would be assigned to the Unit 2 'A' train busses. However, there is a DUAL UNIT LOSP and the 1C DG gets assigned to the Unit 2 'A' Train busses. Applicants often get confused as to the assignment of 'A' train DGs during various 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 offsite power available and with or without the presence of SIAS signal, the on-service train CCW pump shall trip, then both train CCW pumps start by the diesel generator loading sequencers (ESS or LOSP).

The 2C CCW pump is the 'A' train pump therefore it will be powered from the 1C DG.

Second part is correct. Under the conditions in the stem, the 1C DG 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 of bus power supplies to the following: CCW pump, including emergency backup Importance Rating: 3.0* 3.2*

Technical

Reference:

FSD-A181000, Component Cooling Water, Ver 24 FSD-A181005, Diesel Generators, Ver 44 References provided: None Learning Objective: NAME AND IDENTIFY the Bus power supplies, for those electrical components associated with the CCW System, to include the following: (OPS-40204A04):

Question History: NEW K/A match: Requires the applicant to know the normal bus power supply to the CCW pumps in order to know the correct DG that is its emergency backup power supply.

SRO justification: N/A

11. 009EK2.03 011 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 level must be greater than (1) to (2) .

A. 1) 31%

2) ensure SG tubes are covered to promote reflux boiling B. 1) 48%

2) ensure SG tubes are covered to promote reflux boiling C. 1) 31%

2) ensure adequate SG inventory to provide a secondary heat sink D. 1) 48%

2) ensure adequate SG inventory to provide a secondary heat sink

ESB-1.2 ERP Step Text - Check intact SG levels

Purpose:

To ensure adequate feed flow or SG inventory for secondary heat sink requirements ESP-1.2

8. Check any intact SG narrow range level - GREATER THAN 31% {48%}

Distracter analysis A. Incorrect. First part is correct (See C.1).

Second part is incorrect (See C.2). Plausible if the applicant does not understand the mechanism of reflux boiling and improperly assumes this leak rate is sufficient to allow reflux boiling to be the method of heat removal at some time during the event. Reflux boiling is mentioned in ECP-1.1, Loss of Emergency Cooling Recirculation as a method of cooling if the RCS is NOT full.

B. Incorrect. First part is incorrect (See C.1). Plausible if the applicant applies adverse 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, adverse numbers 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 maintaining SGWL above 31% (See ESB-1.2 above). Additionally, at this leak rate, 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 SI at 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 interrelations between the small break LOCA and the following: S/Gs Importance 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 2.1 References provided: None Learning Objective: STATE AND EXPLAIN the basis for all Cautions, Notes, and Actions associated with ESP-1.2, Post LOCA Cooldown and Depressurization. (OPS-52531F03)

Question History: VOGTLE 10 K/A match: The applicant must know how the SGs interrelate to the RCS during a Small Break LOCA in that they are required to provide a secondary heat sink for the RCS.

SRO justification: N/A

12. 010A2.01 012 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 describes Pressurizer Heater operation per ESP-0.1, Reactor Trip Response?

A. The 1A PZR Heaters will have power available, NO actions are required to energize them.

B. The 1A PZR Heaters will have power available AND manual actions on the MCB are required to energize them.

C. The 1A PZR Heaters will NOT have power available. Manual actions are required to align power to them on the EPB but NO other actions are required to energize them.

D. The 1A PZR Heaters will NOT have power available. Manual actions are required to align power to them on the EPB AND manual actions on the MCB are required to energize them.

ESP-0.1 1.10.4 WHEN pressurizer heater group 1A operation is

desired, THEN place HTR GRP 1A BLOCKING BYPASS SW to BYPASS.

1.10.5 IF required, THEN manually energize pressurizer heater group 1A.

Distracter analysis A. 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 is taken to BYPASS and the heater control switch placed in ON to energize the heaters.

B. Correct. The BIF LOSP sequencer reenergizes the emergency section of 600v LC A on an LOSP at step 6. By procedure, the heater switch is taken to off, then the blocking bypass switch is taken to BYPASS and 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 1B PZR heaters which require EPB alignment and fails to recall that by procedure, the heater switch is taken to off, then the blocking bypass switch is taken to BYPASS and the heater control switch is placed in ON to energize the heaters.

D. Incorrect. See B. Plausible if the if the applicant confuses the 1A with the 1B PZR heaters which require EPB alignment but recognizes the heater switch is taken to off, then the blocking bypass switch is taken to BYPASS and the heater control switch placed in ON to energize the heaters.

K/A: 010A2.01 Pressurizer Pressure Control System (PZR PCS) - Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Heater failures Importance Rating: 3.3 3.6 Technical

Reference:

FNP-1-ESP-0.1, Reactor Trip Response, Ver 32 FNP-1-EEP-0.0, Reactor Trip or Safety Injection, Ver 44 References provided: None Learning Objective: NAME AND IDENTIFY the Bus power supplies, for those electrical components associated with the Pressurizer Pressure 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 become unavailable for use (failed) until operator action is taken to mitigate their loss. The applicant must know how to re-energize the PRZR heaters when they are lost during an LOSP.

SRO justification: N/A

13. 011EK2.02 013 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 Supervisor directs the OATC to perform ESP-1.4, Transfer to Simultaneous 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 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

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)

A. HOT COLD B. COLD COLD C. HOT HOT D. COLD HOT

ESP-1.4 intends to align LHSI to HOT leg and leave HHSI aligned for Cold leg recirculation. However, during realignment, if any portion of the LHSI system cannot be reconfigured (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 the operator assess the re-alignment and if the final requirement is not met, they are directed to Step 1 and contacting the Technical Support Center for guidance We have recently developed a JPM that causes the alignment to be in a cold/cold or hot/hot alignment since some failures can lead you there. That is precisely the reason the procedure will direct you to the TSC staff if a final alignment other than cold/hot or hot/cold is reached by the end of the procedure.

The final alignment of LHSI and HHSI can be confusing when coupled with power losses and/or equipment failures.

Distracter analysis A. 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 to the Hot Leg. This would be correct if it were HHSI.

Second part is incorrect (See B.2). Plausible if the applicant believes 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 is lost or equipment malfunctions, there are allowances to come back to steps in the procedure and perform them when power is restored and/or equipment repaired such as in ESP-1.3 when the charging suction and discharge header MOVs are aligned. Any alignment is possible in this procedure once malfunctions occur.

D. Correct. First part is correct. Per ESP-1.4 Step 1, if both trains of LHSI cannot be aligned to Hot Leg recirc then both trains are left aligned to cold leg recirc.

Second part is correct. ESP-1.4 Step 2 will align the A train HHSI to Hot Leg recirc and Step 3 will leave B train in its original alignment.

K/A: 011EK2.02 Large Break LOCA - Knowledge of the interrelations between 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 Recirculation, Ver 16.

References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if any system components 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 BANK K/A match: The applicant is required to know the interrelation between the RHR/Charging Pumps and the RCS during a Large Break LOCA. Based on the scenario given, the applicant must determine the final pump alignment.

SRO justification: N/A

14. 011K2.02 014 Unit 1 is operating at 100% power when the 1E 4160V bus becomes de-energized due to an electrical fault.

Which one of the following completes the statement below?

Pressurizer heater groups have lost their normal power supply.

A. 1C and 1D ONLY B. 1A, 1C and 1D ONLY C. 1B and 1E ONLY D. 1B, 1D and 1E ONLY Load Pressurizer 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 D Pressurizer Heater Group E 600V LC N (Unit 1 and 2) 4160V E Distracter analysis A. Incorrect. See C. Plausible since these heaters are powered from the same LC. The applicant may believe that this is the only LC affected by the power loss. Although 1A heaters are also powered from the same 4160V Bus as these heaters, the applicant may believe the 1A 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 1D 4160V bus and the applicant may think theses heaters are supplied 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 heaters D. Incorrect. See C. Plausible if the applicant knows that 2 sets of heaters are powered from the same LC but cannot correctly recall which ones.

The 1B heaters is a partially correct answer and would be included if the applicant thinks these heaters are powered from the same 4160V bus as the 1B heaters.

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

Reference:

A506250, Unit 1 Electrical Load List, Ver 74.0 References provided: None Learning Objective: NAME AND IDENTIFY the Bus power supplies, for those electrical components associated with the Pressurizer Pressure 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 the pressurizer heaters in order to determine which ones have lost power. The power supply has to go back to the 4160V bus so the applicant also has to know the LC supplies as well.

SRO justification: N/A

15. 012A4.06 015 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.

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

FSD-A181007:

3.3.2 pg 3-10 The first method of tripping the breaker (i.e., reactor trip or bypass breakers) is by a loss or drop of rated voltage to the Undervoltage Relay (UV). The relay is normally energized from the 48 volt DC from the RPS. When the voltage is removed by an automatic 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 the normally de-energized shunt trip (SHTR) coil is energized. When energized, the SHTR coil is powered from the 125 volt DC system used to close the reactor trip and bypass breaker closing circuits.

Distracter analysis A. Correct. Without 'A' train DC, the UV coils from the "A' Train Reactor Protection System (RPS) will still open 'A' Trip and 'B' Bypass breakers. 'B' train RPS deenergizes the UV coils for 'B' Trip and 'A' Bypass breakers. 'B' Train Aux Building DC will open the 'B' Trip breaker.

B. Incorrect. See A. Plausible if the applicant did not recall that the UV coils from RPS will trip ALL Trip and Bypass breakers. Since both listed breakers are 'B' breakers, this adds to plausibility due to the applicant thinking the 'B' train is unaffected and still would cause a reactor trip if the system worked this way.

C. Incorrect. See A. Plausible since this is how the RPS opens the Trip and Bypass 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 on the RPS.

D. Incorrect. See A. Plausible if the applicant thinks that the Trip breakers are tripped by RPS and the Bypass breakers from Aux Building DC.

Since the Shunt trip coils on the Bypass breakers can ONLY be operated locally, the applicant may think that without DC the Bypass breakers will not open.

K/A: 012A4.06 Reactor Protection System (RPS) - Ability to manually operate and/or monitor in the control room: Reactor trip breakers Importance Rating: 4.3 4.3 Technical

Reference:

FSD-A181007 Reactor Protection System, Ver 18 D-177198, Sheet 2, Ver 3 References provided: None Learning Objective: RECALL AND DESCRIBE the operation and function of the following reactor trip signals, permissives, control interlocks, and engineered safeguards actuation signals associated with the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) to include setpoint, coincidence, rate functions (if any), reset features, and the potential consequences for improper conditions to include those items in the following tables (OPS-52201I07):

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

SRO justification: N/A

16. 013G2.1.19 016 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 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)

A. IS are NOT B. IS ARE C. is NOT are NOT D. is NOT ARE

EEP- 0.0 -

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

SI Signal Instrumentation Setpoint Coinc

1. Pressurizer pressure low PT 455, 456, 457 1850psig 2/3 FSD-A181007 - Pg 2-26 The Main Steam Line Isolation is initiated by the following:

b. Low steam pressure; = 585 psig on 2/3 S.G.

Distracter Analysis A. 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 indicate OPEN, the downstream steam header pressure is 6.4 psig while the upstream pressure is >900 psig on all SGs, MSIV-3370A/B/C must 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 to evaluate steam pressure and assumes that since MSIV-3369A/B/C are open then MSIV-3370A/B/C must also be open.

C. Incorrect. First part is incorrect (See A.1). Plausible if the applicant has the misconception that the control channels, PT 444A and 445A are used to evaluate pressure instead of the protection channels PT-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 or component status Importance 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 18 References provided: Screen Image of the Integrated Plant Computer Learning Objective: DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the Emergency Core Cooling System components and equipment, to include the following (OPS-40302C07):

[...]

Automatic actuation including setpoint ( example SI, Phase A, Phase B, MSLIAS, LOSP, SG level)

[...]

Actions needed to mitigate the consequence of the abnormality Question History: NEW K/A match: Applicant must evaluate a set of data from the plant computer and based on that determine if ESFAS system actuation is necessary.

SRO justification: N/A

013G2.1.19 P 016

17. 014K5.01 017 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 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) the most reliable indication.

A. 1) exactly 100 steps

2) the group step counters B. 1) approximately 100 steps

2) the group step counters C. 1) exactly 54 steps

2) DRPI D. 1) approximately 54 steps

2) DRPI The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/- ? inch). If a rod does not move one step for each demand pulse, the step counter will still count the 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 on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one system fails, the DRPI will go on half accuracy with an effective coil spacing of 7.5 inches, 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 and temperature tolerances), and the maximum uncertainty is +/- 10 steps (only one data system A or B coils operable). With an indicated deviation of 12 steps between the group step counter and DRPI, the maximum deviation between actual rod position and the demand position could be 22 steps.

Distracter analysis A. Incorrect. First part is incorrect (See D.1). Plausible if the applicant does not recall that the Step Demand Counter only counts electrical impulses 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 not recall that the Step Demand Counter only counts electrical impulses. Each bank has 2 Step Demand Counters that step in 1 step 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 may reason that this is a potential reason to call rods by step counter as approximate.

Second part is incorrect (See A.2).

C. Incorrect First part is incorrect (See D.1). Plausible since the DRPI lights change only every 6 steps and 54 steps is a DRPI display light location. The applicant may think that since a DRPI display light is lit, 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 position based on the location of the rod in reference to the measurement coils and the step counter only counts electrical pulses, the rod is at ~54 steps. Also, the accuracy of DRPI is + 4 steps so the rod height is approximate.

Second part is correct. DRPI is the most reliable because it actually senses the location of the rod using coils.

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

Reference:

FNP Technical Specifications Bases, Ver 58 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Digital Rod Position Indication System (OPS-52201F02):

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

SRO justification: N/A

18. 015/17AA2.02 018 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 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) .

A. 1) Motor Bearings

2) 2 minutes B. 1) Motor Bearings

2) 60 minutes C. 1) Pump Lower Radial Bearings

2) 2 minutes D. 1) Pump Lower Radial Bearings

2) 60 minutes

DD1 - CAUTION: RCPs with #1 Seal Leakoff less than 2.5 gpm may develop lower bearing and seal temperatures that exceed 225°F within 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a loss of seal injection.

DD3 - On a complete Loss of CCW Flow to RCP Motor Bearing Oil Coolers, the bearing 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 as time allows.

MOV-3052 isolate CCW flow to the RCP oil coolers and the RCP thermal barrier heat exchanger which functions to cool the lower radial bearing on a loss of RCP seal injection flow.

Distracter analysis A. Correct. First part is correct. The RCP motor bearings are the most limiting components 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 will increase to 195°F with 2 minutes. The ARP has the operator trip the 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 is the time the lower radial bearing temperature will rise in 1-2 hours on a loss of RCPs with #1 Seal Leakoff less than 2.5 gpm following a loss of seal injection. Plausible since this is a time requirement for a RCP malfunction on the same Annunciator panel as DD3.

C. Incorrect. First part is incorrect (See A.1) RCP lower radial bearings are cooled from two sources. A loss of the CCW will not cause the lower radial bearing temperatures to rise. Plausible since CCW is normal cooling to components and this is one of a few components with 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 to determine and interpret Abnormalities in RCP air vent flow paths and/or oil cooling system as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

Importance Rating: 2.8 3.0 Technical

Reference:

FNP-1-ARP-1.4, DD1 and DD3, Ver 53 References provided: None Learning Objective: LIST AND DESCRIBE the sequence of major actions associated with AOP-9.0, Loss of Component Cooling Water. (OPS-52520I04).

EVALUATE plant conditions and DETERMINE if any system components 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 CCW to the RCP oil coolers and lower radial bearings has resulted, then must interpret how this malfunction affects the RCP components (oil coolers and seal) and the time required for action to be taken.

SRO justification: N/A

19. 017K6.01 019 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 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 Degraded Core Cooling.

(1) (2)

A. HIGH hottest B. HIGH 5th hottest C. LOW hottest D. LOW 5th hottest

CSF-0.2 5th hottest CETC <1200 °F? NO Go to FRP-C.1 YES RCS SUBCOOLING NO 5th hottest CETC <700 °F? NO Go To from CETC > FRP-C.2 16°F {45°F}?

YES CSF - SAT U263686 pg 3-5 The signal conditioning panel contains the open thermocouple detection circuitry, noise filtering capacitors, and the cold reference junction compensation circuitry. Cold junction compensation is accomplished by measuring the barrier temperature utilizing a semiconductor temperature sensor located on the signal conditioning panel. The temperature 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 the channels on the Analog Input Boards (DT1748 and DT1748-24EX boards'. The thermocouple signals are then compensated in the software by adding the value of the cold reference junction to the thermocouple signals. If any of the thermocouples are open or shorted the signal conditioning panels open thermocouple detection circuitry will cause the input to be driven down to 0V.

Distracter analysis A. Incorrect. First part is incorrect (See D.1). Plausible if the applicant does not recall if a thermocouple fails high or low when shorted. An RTD that experiences an open circuit will cause a high temperature reading. The applicant could confuse RTD and thermocouple operating theory.

Second part is incorrect (See D.2). Plausible if the applicant doesn't recall that the 5th hottest is selected to allow for failed high thermocouples. 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 of the effect of a loss or malfunction of the following ITM system components: Sensors and detectors Importance 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 1 U-263686, ICCMS Tech Manual Vol II, Ver 2 OPS-31701G, Sensors and detectors, Ver 4 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Inadequate Core Cooling Monitor System (OPS-52202E02):

CETC Monitor Question History: NEW K/A match: The applicant is required to have a knowledge of the effect of a shorted CETC on the incore temperature monitoring system.

SRO justification: N/A

20. 022A1.03 020 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 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)

A. ductwork MORE B. dropout plate MORE C. ductwork LESS D. dropout plate LESS Added that the Containment Coolers are running in Slow Speed prior to the LOCA. The FSD states that the design of the Containment Cooler motor is based on 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 Containment atmosphere increases to the point where the Slow speed fan would draw more current than Fast speed. With the Containment Cooler in Slow at the beginning of the event, the applicant will only have to evaluate the current drawn by the cooler 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 the CCS controls including: Containment humidity.

Tech Spec 3.6.6 Bases:

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

are operated at the lower speed during accident conditions to prevent motor overload from the higher mass atmosphere. In addition, if temperature at the cooler discharge reaches 135°F, fusible links holding dropout plates will open and the fan discharge will no longer be directed through the common discharge header. This function helps to protect the fans in a post-accident environment by reducing 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 brake horsepower requirement of 105 hp during low-speed operation following a LOCA.

During normal operation, the design brake horsepower of the fan in high-speed operation is 80 hp.

Distracter analysis A. Incorrect. First part is incorrect (See B.1). When containment temperature reaches ~135°F, the dropout plates fall open. Plausible if the applicant believes that the links melt at 160°F to 175°F which is when 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 in Containment due to the LOCA will cause the fan motors to draw more current.

C. Incorrect. First part is incorrect (See A.1).

Second part is incorrect (See B.2). Plausible if the applicant only thought that the increase in temperature caused air density to lower 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 predict and/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: None Learning Objective: DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the Containment Spray and Cooling System components and equipment, to include the following (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 containment humidity and based on this predict the changes in the parameters of the containment cooler fans in an environment with more humidity (water vapor) and will draw more current. The dropout plate will actuate at 135°F to reduce the back pressure on the fans.

SRO justification: N/A

21. 022AK1.04 021 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 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 would decrease to a minimum flow rate of (1) , which is designed to prevent (2) .

A. 1) 18 gpm

2) flashing downstream of the letdown orifices B. 1) 18 gpm

2) overheating of the charging pumps C. 1) 40 gpm

2) flashing downstream of the letdown orifices D. 1) 40 gpm

2) overheating of the charging pumps

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

Distracter analysis A. Correct. First part is correct. Minimum charging flow in AUTOMATIC is 18 gpm.

Second part is correct. Per P&L 3.3, 18 gpm ensures adequate cooling to the regenerative heat exchanger to prevent flashing downstream 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 thinks that reduced charging flow would equate to reduced mini-flow flow which is incorrect. Also, the charging miniflow goes through the seal water return HX and temperature would be unaffected by changing charging flows.

C. Incorrect. First part is incorrect (See A.1). Plausible since Figure 1 of SOP-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 the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Reason for changing from manual to automatic control of charging flow valve controller.

Importance Rating: 2.9 3.0 Technical

Reference:

FNP-1-SOP-2.1, Chemical and Volume Control System Plant Startup and Operation, Ver 131 References provided: None Learning Objective: RECALL AND DISCUSS the Precautions and Limitations (P&L), Notes and Cautions (applicable to the Reactor Operator) 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 controller failure. FK-122 has been placed in manual and is now being placed in AUTO. A reason for placing FK-122 in AUTO and not leaving it in MANUAL is to ensure adequate cooling to the regenerative heat exchanger to prevent flashing downstream of the letdown orifices with one 60 gpm orifice on service should the controller fail to minimum demand based on current plant conditions.

The operational implication would be that flashing would occur if Chg flow were to fall to <18 gpm and cause damage to the orifices and piping due to water hammer and the flashing of water to steam. A loss of letdown would be the result. This question meets the KA in that it asks the minimum flow rate for being in auto if a controller were to fail and the reason.

SRO justification: N/A

22. 022K4.03 022 Unit 1 is operating at 100% 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 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)

A. IS is NOT B. is NOT is NOT C. IS IS D. is NOT IS

EE5 Automatic Action:

1. The following valves close c) CCW FROM RCP THRM BARR Q1P17HV3045 f) CCW FROM RCP THRM BARR Q1P17HV3184 FSD A181003 3.3.1.2 The seal water injection lines to the RCP are considered as open flow paths post-LOCA. The high pressure inflow through these lines during the injection and recirculation phases precludes any containment to atmosphere leakage. In the event of a loss of seal water flow through these lines, a water seal in the charging pump suction and discharge piping precludes containment to atmosphere leakage.

Distracter analysis A. Correct. First part is correct. Phase B isolates CCW cooling to the RCP Thermal Barrier Heat Exchanger.

Second part is correct. Seal injection is NOT isolated by SI, Phase A or Phase B.

B. Incorrect. First part is incorrect (See A.1). Plausible if the applicant does not recall all the components isolated on a Phase B. Since CCW is water 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 return is isolated on an SI and the applicant could confuse the two.

This is a plausible combination if the applicant reasons that the shutdown seal will actuate and seal injection and CCW to the thermal barrier hx are no longer needed D. 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 CCS design feature(s) and/or interlock(s) which provide for the following: Automatic containment isolation.

Importance Rating: 3.6* 4.0 Technical

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: None Learning Objective: DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the Containment Structure and Isolation System components and equipment, to include the following (OPS-40302B07):

[]

[]

Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLIAS, LOSP, SG level)

Question History: MOD SUMMER 11 K/A match: Requires the applicant to know the design feature of the Phase B Containment Isolation that isolates CCW cooling to the RCP Thermal Barrier Heat Exchanger.

SRO justification: N/A

23. 025AA1.01 023 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 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 preferred method to re-establish core cooling is to establish (1) .

Core cooling is monitored using (2) .

(1) (2)

A. feed and bleed RCS cold leg temperatures B. a secondary heat sink RCS cold leg temperatures C. feed and bleed CETCs D. a secondary heat sink CETCs ARG-1 If the RCS is intact and the loops are not isolated with SG nozzle dams or loop isolation valves, a secondary heat sink using half or more SGs will be an effective alternate mode of decay heat removal that will last for several hours or longer. Since there would be no significant fluid inventory losses for this case, makeup requirements can easily be met with a minimum amount of charging flow or possibly RWST (or VCT) gravity feed if initiated early enough. For this situation, it should also be possible to refill and pressurize the RCS and then operate the RCPs to sweep the noncondensibles from the 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 handhole covers - INSTALLED.

NOTE: Establishing a secondary heat sink will reduce RCS heat up and pressurization rate to provide more time for recovery actions.

25. Verify secondary heat sink established.

25.1 Maintain wide range level in all available SGs greater than 75% using FNP-1-SOP-22.0, AUXILIARY FEEDWATER SYSTEM.

25.2 IF SG steam space intact, THEN open atmospheric relief valves to prevent SG pressurization.

1A(1B,1C) MS ATMOS REL VLV PC 3371A adjusted PC 3371B adjusted PC 3371C adjusted 25.3 IF SGBD system available, AND AFW system available, THEN establish blowdown from available 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 than 121 ft 11 in AND core exit T/Cs GREATER than 200°F.

Distracter analysis A. Incorrect. First part is incorrect (See D.1). Plausible if the applicant believes that establishing a secondary heat sink is not correct because RCS temperature is less than 200°F and so steaming the SG would not be 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 in other 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, establishing a 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 to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: RCS/RHRS cooldown rate Importance Rating: 3.6 3.7 Technical

Reference:

Background Information for WOG Abnormal Response Guideline ARG-1 Loss of RHR While Operating at Mid-Loop Conditions, Ver 2 FNP-1-AOP-12.0, RHR System Malfunction, Ver 25 References provided: None Learning Objective: LIST AND DESCRIBE the sequence of major actions associated with AOP-12.0, RHR System Malfunction and/or STP-18.4, Containment Closure. (OPS-52520L04)

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

SRO justification: N/A

24. 026AG2.4.50 024 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 completes the statements below?

CCW system automatic isolations are designed to occur at (1) in the CCW Surge Tank.

Using the NORMAL source of makeup water, the operator will open (2) to make up to the CCW Surge Tank.

(1) (2)

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

ARP AA5: 20 inches Automatic Action

1. Closes CCW Valves (Q1P17HV3096A&B) to isolate CCW to/from Evaporator Packages and H2 Recombiners. (Q1P17LSLL3027CD-A)

2. Trips closed Q1P17HV2229, CCW to Sample Cooler (Q1P17LSLL3027CD-A).

Operator Action 4.1 Attempt to fill CCW surge tank using Normal Make-up to maintain level above the lo level alarm point as follows; 4.2. IF unable to fill the CCW Surge Tank per the Normal Make-up method, THEN attempt to fill CCW surge tank using Emergency Make-up to maintain level above the lo level alarm point as follows;

5. IF a loss of CCW cooling has occurred, THEN refer FNP-1-AOP-9.0, LOSS OF COMPONENT COOLING WATER.

Distracter analysis A. Incorrect. First part is incorrect (See D.1). Plausible since this is the Surge tank LO Level alarm setpoint.

Second part is incorrect (See D.2). Plausible if candidate cannot recall 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 the automatic closure of HV3096A&B and HV2229.

Second part is correct. Demin water storage tank is the normal source for makeup to the CCW surge tank.

K/A: 026AG2.4.50 Loss of Component Cooling Water (CCW) - Ability to verify system alarm setpoints and operate controls identified in the alarm 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 53.1 References provided: None Learning Objective: SELECT AND ASSESS the following instrument/equipment response expected when performing CCW System evolutions including the fail condition, alarms, and trip setpoints (OPS-52102G07).

Surge Tank Level Question History: NEW K/A match: Requires the applicant to determine at which level the automatic isolations of the CCW system occur (verify system alarm setpoints which is when these valves close) and know what source of water is used to fill the surge tank (operate controls identified in the ARP to raise the CCW Surge Tank level).

SRO justification: N/A

25. 026K2.01 025 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)

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

In the LOSP the 1-2A DG will tie onto the unit 1 A Train busses. Then when the SI and subsequent phase B signal comes in the 1-2A DG and 1C DG will load shed, and then 1-2A DG will align to Unit 2 and the 2A CS pump will start at step 2.

FSD-A181008: 3.1.5.2 Without offsite power available, the CSS pumps shall start by the diesel generator ESS loading sequencer. Starting will occur at step two of the sequence if the "P" signal is present at that time. If the "P" signal occurs between the completion of step two and step six of the ESS sequence, then starting will occur at the completion of step six of the loading sequence. If the "P" signal occurs after the completion of step six, starting will take place immediately.

Pg 2-1 CSS initiation is automatic upon a containment pressure hi-3 signal ("P" signal)

FSD-A181005 LOSP on both units and LOCA on Unit 2:

For LOSP on both units and LOCA on Unit 2, the alignment of the diesel generators will be as follows:

1-2A Unit 2 Buses 2F and 2K 1C Unit 1 Buses 1F, 1K and 1H 1B Unit 1 Buses 1G, 1L and 1J 2B Unit 2 Buses 2G, 2L and 2J Distracter analysis A. Incorrect. First part is correct (See C.1).

Second part is incorrect (See C.2). Plausible since the 1C and 1-2A DGs align to either Unit's 'A' Train depending on the scenario and 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 not recall 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 is 2F 4160V AC bus.

Second part is correct. For LOSP on both units and LOCA on Unit 2, 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 bus power supplies to the following: Containment spray pumps.

Importance Rating: 3.4* 3.6 Technical

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 18 References provided: None Learning Objective: NAME AND IDENTIFY the Bus power supplies, for those electrical components associated with the Containment Spray and Cooling System, to include those items in Table 3- Power Supplies (OPS-40302D04).

Question History: NEW K/A match: Requires the applicant to know the normal power supply to the 2B CS pump and the 1-2A DG alignment and power supply to the 2A CS pump upon an LOSP with a subsequent SI.

SRO justification: N/A

26. 029A1.02 026 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 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)

A. WILL will NOT B. will NOT will NOT C. WILL WILL D. will NOT WILL

SOP-45:

3.5 The radiation monitors fail to a High Radiation condition on loss of instrument and/or control power that will result in actuation of associated automatic functions.

FH4: CP RE 24A or B HI RAD PROBABLE CAUSE

1. High Radiation Level in the Containment Purge Exhaust Line.

2. The radiation monitors fail to a High Radiation condition on loss of instrument and/or control power that will result in actuation of associated automatic functions.

AUTOMATIC ACTION

1. 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 analysis A. Correct. First part is correct. Per SOP-45, a radiation monitor that has lost control power will initiate its automatic actions.

Second part is correct. R-24A will NOT automatically secure the main purge supply and exhaust fans.

B. Incorrect. First part is incorrect (See B.1). Plausible if candidate does not recall 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 seem logical to the applicant that when the main purge supply and exhaust 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 does not recall what auto functions are actuated by R-24A and believes that R-24A will trip the fans. Additionally, with the Aux Building main exhaust fan running, the applicant could reason that there is still 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/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment Purge System controls including: Radiation levels Importance 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 38.1 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Containment Ventilation and Purge System, to include those items 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 the off site radiation exposure limit from potentially being exceeded.

SRO justification: N/A

27. 033K4.05 027 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 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, the MINIMUM required SFP boron concentration is (2) ppm.

(1) (2)

A. boration 2000 B. boration 2200 C. dilution 2000 D. dilution 2200

Tech Specs 3.7.14 The 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 blender to indicate total makeup flow to the charging header and as input to the Reactor Makeup Control System.

5.30.5.2 Upstream boric acid flow measurement (FT-113) shall be provided to serve as input to the Reactor Makeup Control System.

5.78.1.1 This differential pressure transmitter (and associated orifice flow element) shall provide measurement of the total makeup flow from the boric acid blender. It shall also provide input to the Reactor Makeup Control System for regulation of RMW flow and shall alert the operator of a deviation from the selected flow setpoint.

Distracter analysis A. Incorrect. First part is incorrect (See C.1). Plausible if the candidate thinks that the lower signal results in less RMW added and therefore more 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 applicant could confuse the SFP minimum boron concentration with the accumulators minimum boron concentration which is 2200 ppm.

C. Correct. First part is correct. FT-168 will regulate total makeup flow to 120 gpm. This is a combination of acid flow and Reactor Makeup Water (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" total flow of 120 gpm resulting in more RMW than expected therefore a lower boron concentration in the makeup water supplied to the SFP. This will result in a dilution of the SFP.

Second part is correct. Tech Spec 3.7.14 requires the SFP boron concentration 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 of design feature(s) and/or interlock(s) which provide for the following: 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 22 References provided: None Learning Objective: RECALL AND APPLY the LCO and APPLICABILITY for Technical Specifications (TS) or TRM requirements, and the REQUIRED ACTIONS for 1 HR or less TS or TRM requirements, and the relevant portions of BASES that DEFINE the OPERABILITY and APPLICABILITY of the LCO 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 characteristics including design features, capacities and protective interlocks for the components associated with the Reactor Makeup 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 the applicant to evaluate knowledge of TS requirement for minimum boron concentration, which provides for adequate SDM. Also evaluates candidates ability to predict effect of an equipment malfunction which could adversely affect ability to maintain desired boron concentration.

SRO justification: N/A

28. 035A2.01 028 Unit 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 are required to (2) .

(1) (2)

A. steam reduce turbine load B. steam trip the reactor C. feed reduce turbine load D. feed trip the reactor AOP-14:

1. [CA] Evaluate plant status 1. Perform the following for safe operation.

Pressurizer level 1.1 Verify reactor tripped GREATER THAN 15%

AND Pressurizer pressure 1.2 IF reactor tripped, GREATER THAN 2000 psig THEN CLOSE SG AND main steam isolation and Steam generator pressure bypass valves GREATER THAN 650 psig AND Containment pressure LESS THAN 2 psig AND IF main generator on line, THEN (check reactor power) - (turbine power + any steam dump power) mismatch LESS THAN 10%.

AND IF 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 AFW system if possible and step 9 has the crew reduce power per UOP-3.1 and UOP-2.1 if the above trip criteria is not met.

Distracter analysis A. Incorrect. First part is correct (See B.1).

Second part is incorrect (See B.2). Plausible since UOP-3.1, Power Operation, requires a reduction in turbine load if 100% power is exceeded. This would be the correct thing to do if containment pressure did not meet the reactor trip criteria. Also Step 7 and 9 and note above step 9 addresses ramping the unit down to mode 2 if 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 a steam break where the steam is exiting the piping before reaching the turbine. The containment parameters show that the break is in containment.

Second part is correct. With containment pressure greater than 2 psig, reactor trip criteria is met.

C. Incorrect. First part is incorrect on a feedline break, Rx power would be stable, turbine MWe would be stable, and RCS pressure would be stable. 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 similar characteristics.

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 the impacts 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 malfunctions or operations Importance Rating: 4.5 4.6 Technical

Reference:

FNP-1-AOP-14.0, Secondary System Leakage. Ver 10.0 References provided: None Learning Objective: STATE AND EXPLAIN the operational implications for all Cautions, 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 occurred based on the impact to plant parameters (which is a backward logic way to meet the first part of the KA). The parameters include but are not limited to SGs since the trip criteria in AOP-14 is due to ctmt pressure. Applicant must determine the proper procedural response to mitigate a faulted SG inside ctmt.

SRO justification: N/A

29. 036AG2.1.7 029 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 reports 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 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)

A. AOP-49.3 make up to the SFP using the RWST B. AOP-30.0 make up to the SFP using the RWST C. AOP-49.3 ensure all SFP hatches and doors are closed D. AOP-30.0 ensure all SFP hatches and doors are closed AOP-30 Symptoms or entry conditions

1. This procedure is entered when a fuel handling accident causes damage to a fuel assembly 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 EXH Step 1.6. Dispatch personnel to close all spent fuel area fuel handling hatches.

Step 5. Verify all access doors to accident area - CLOSED Step 20 has the operator makeup to the refueling cavity from the RHR system if the cavity is low.

AOP-49.3 B. Symptoms or entry conditions

1. 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 result in a long term loss of Spent Fuel Pool cooling.

Distracter analysis A. Incorrect. First part is incorrect (See D.1). Plausible based on the name of the procedures. The applicant may believe that since there is no refueling occurring (Unit 1 at 100%), AOP-30 does not apply and AOP-49.3 applies since there is "an emergency" in the SFP.

Second part is incorrect (See D.2). Plausible since this is an action of AOP-49.3 to keep all assemblies covered with water and would seem logical since there was a SFP HI-LO level alarm and a damaged fuel assembly lying on the racks. The applicant may think that keeping the damaged assembly covered with water is a required action.

B. Incorrect. First part is correct (See D.1)

Second part is incorrect (See D.2). Plausible since AOP-30 directs filling the refueling cavity. The applicant could easily confuse this action with filling the SFP. It could seem logical since there was a SFP HI-LO level alarm and a damaged fuel assembly lying on the racks. The applicant may think that keeping the damaged assembly 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 to AOP-49.3 since during a SFP Emergency, the applicant could assume the affected area would be isolated as radiation levels are high.

D. Correct. First part is correct. This scenario meets the entry requirements of AOP-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 plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Importance Rating: 4.4 4.7 Technical

Reference:

FNP-1-AOP-30.0, Refueling Accident, Ver 19 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if entry into AOP-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 applicable procedure to enter and the appropriate action to take (operational judgment).

SRO justification: N/A

30. 037AK1.01 030 Unit 1 is performing the actions of AOP-2.0, Steam Generator Tube Leakage, due to a tube 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)

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

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 in CETC mode.

Distracter analysis A. Incorrect. First part is incorrect (See C.1). Plausible since the applicant may determine subcooling based on ruptured SG pressure vs. RCS pressure. This would be a common misconception since in a SGTR procedure the ruptured SG is the focus for determining SG pressure less than RCS pressure and is the pressure referred to when 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 than 16°F, however, RCS pressure is less than the affected SG pressure and pzr level is >15%.Therefore the pressure reduction is required to be stopped. Plausible since the subcooling is met and the Pzr level is <63%, so one more evaluation as to be done. This evaluation has two components and one of the two components is met. Also if the candidate were to compare RCS pressure vs one of 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 3 components, two of which do not require the RCS pressure reduction to be stopped. However, RCS pressure is less than affected SG pressure, (RCS pressure is 916 psig and SG pressure is 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 the operational implications of the following concepts as they apply to Steam Generator Tube Leak: Use of steam tables Importance Rating: 2.9* 3.3 Technical

Reference:

FNP-1-AOP-2.0, Steam Generator Tube Leakage, Ver 35 Properties of saturated and superheated steam, 1967 References provided: Steam tables and AOP-2.0 step 34.2 Ver 35.0 Learning Objective: ANALYZE plant conditions and DETERMINE the successful completion 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 determine current value of subcooling and the whether or not the RCS depressurization is required to be stopped during a SG tube leak scenario.

SRO justification: N/A

31. 038EK3.02 031 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 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)

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

20.1 Check SUBCOOLED MARGIN MONITOR indication - GREATER THAN 16°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-3 Step 20 Basis: [...] If SI flow is not terminated, leakage into the secondary will eventually fill the steam generator with water and lift the atmospheric relief valves. This could damage the relief valve and main steamline which would complicate subsequent recovery and aggravate the radiological consequences. Hence, SI must be terminated when the criteria in subsequent steps are satisfied to prevent steam generator overfill Distracter analysis A. Incorrect. First part is incorrect (See C.1). Plausible if the applicant believes that 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 the reason to terminate SI in EEP-0 and ESP-1.1 for a spurious SI.

The applicant could confuse the basis for these procedures with the 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 is not terminated, leakage into the secondary will eventually fill the steam generator with water and lift the atmospheric relief valves. This could damage the relief valve and main steamline which would complicate subsequent recovery and aggravate the radiological consequences. Hence, SI must be terminated when the criteria in subsequent steps are satisfied to prevent steam generator overfill D. 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 it makes 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 - GREATER THAN 31%{48%} but AFW flow is > 395 gpm. One does not meet SI termination and 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 the reasons for the following responses as the apply to the SGTR: Prevention of secondary PORV cycling Importance Rating: 4.4 4.5 Technical

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 2 References provided: None Learning Objective: STATE AND EXPLAIN the basis for all Cautions, Notes, and Actions associated with EEP-3, Steam Generator Tube Rupture. (OPS-52530D03)

Question History: NEW K/A match: Requires the applicant to know that preventing the SG PORVs from lifting due to filling the SGs solid will prevent a radiological release from the atmospherics.

SRO justification: N/A

32. 039K1.09 032 Concerning 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)

A. upstream 25%

B. downstream 25%

C. upstream 10%

D. downstream 10%

FSD-A181015 3.2.8 These detectors are located to monitor the main steam lines upstream of the safety relief valves for the presence of Nitrogen-16 activity in the steam lines and alert the operator when setpoints are exceeded.

SOP-69 Step 3.1 - The system receives a reactor power input from power range channel N-43. IF N-43 fails OR is in Test OR is less than 20% power, THEN the system cannot accurately estimate a leak rate in the AV mode, and the indicators will display "PN <20%". If desired, the Counting Room can configure the N-16 system in the 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 can be used to indicate if leakage is increasing based on the indication trending up. The AV mode is the preferred mode of operation above 20% reactor power. The ME mode should only be utilized below 20% reactor power.

Distracter analysis A. Correct. First part is correct. R-70s are located to monitor the main steam lines 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 not recall 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 fails to 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 of the physical connections and/or cause-effect relationships between the MRSS and the following systems: RMS Importance rating: 2.7 2.7 Technical

Reference:

FSD-A181015, Radiation Monitoring System, Ver 14 FNP-1-SOP-69, N-16 Primary to Secondary Leak Detection System, Ver 5 D-175033, SH1, Main and Aux Steam, Ver 38 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Radiation Monitoring System to include those items in Table 4-Remote and Local Indications and Controls (OPS-40305A02).

RECALL AND DESCRIBE the physical in-plant location of those components associated with the Radiation Monitoring System, to include those items in Table 4- Remote and Local 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 steam system and the cause-effect (power level) of when the Rad monitors go into alarm.

SRO justification: N/A

33. 045K1.19 033 Which one of the following coincidences will cause an anticipated transient without 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)

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

FSD- A181007 pg 2-37 C-20 Interlock. Control interlock C-20 is used to enable the Anticipated Transient Without Trip (ATWT) Mitigation System Actuation Circuitry (AMSAC) . When turbine load is > 40% on 2/2 turbine impulse channel detectors and steam generator narrow range water level decreases to <10% on 2/3 steam generators for 25 seconds, the AMSAC system will trip the main turbine and provide an auto start signal to all AFW pumps. There is a time delay drop out associated with the impulse pressure portion of the 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 and permissive interlocks use a 1 of 2 logic to enable or disable functions. 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% enables AMSAC.

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 Low SGWL is 1 of 3 SGWL less that 28% NR. The applicant could improperly believe that AMSAC is 1 of 3 as is the Low Low SGWL logic.

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 the physical connections and/or cause-effect relationships between the MT/G system and the following systems:

ESFAS Importance Rating: 3.4* 3.6 Technical

Reference:

FSD-A181007, Reactor Protection System, Ver 18 References provided: None Learning Objective: DEFINE AND EVALUATE the operational implications of normal / 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 the abnormality.

Question History: MOD CALLOWAY AUG 05 K/A match: AMSAC is listed as a back up to the reactor trip system and ESFAS in the FSAR. This question requires the applicant to know the cause and effect of relationship between AMSAC and the Main Turbine. Conditions which cause AMSAC to be enabled and produce a turbine trip.

SRO justification: N/A

34. 051AK3.01 034 Unit 1 is operating at 40% power when PR-4029, CONDENSER PRESSURE indicates 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)

A. Plant Trip CLOSED B. Plant Trip OPEN C. Loss of Load CLOSED D. Loss of Load OPEN

FSD-A181007 Pg 2-36/37 C-9 Interlock. C-9 is the condenser-available interlock. This interlock allows the steam dump valves to be armed if the condenser is available. It also prevents an overpressure condition which could damage the condenser. To activate C-9, both condenser pressures shall be < 8 inches Hg vacuum, and 1/2 circulation water pump motor breakers must be shut.

8 inches of Hg vacuum is 10.8 psia.

See references Figure 2, Sheet 10 of FSD-A181007.

Distracter analysis A. Correct. First part is correct. A turbine trip results which causes a reactor trip, thus enabling the plant trip controller.

Second part is correct. C-9 is NOT enabled at 12 psia therefore the steam 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 cannot recall that the vacuum setpoint for the C-9 interlock is <10.8 psia and believes that adequate condenser vacuum exists for steam dump operation.

C. Incorrect. First part is incorrect (See A.1). Plausible if the applicant fails to recognize that the turbine trip causes a reactor trip at this power. If rx power were less than 35% then a rx trip would not occur and the turbine 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 for the following responses as they apply to the Loss of Condenser Vacuum: Loss of steam dump capability upon loss of condenser vacuum Importance Rating: 2.8* 3.1 Technical

Reference:

FSD-A181007, Reactor Protection System, Ver 18 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Steam Dump System to include the components found on Figure 5, Steam-Dump Control (OPS-52201G02).

Question History: FNP EXAM BANK K/A match: Requires the applicant to know on a loss of vacuum which controller the steam dumps will operate on and the reason the steam dumps will not operate (loss of capability). On a loss of vacuum the reason is because the C-9 interlock (vacuum) is not met. This is not stated in the stem but is inherent to the question.

SRO justification: N/A

35. 054AA2.05 035 Unit 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, controllers are 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)

A. remain OPEN remain OPEN B. remain OPEN CLOSE C. CLOSE remain OPEN D. CLOSE CLOSE

For this event the applicant has to analyze the situation. For a SGFP trip, AOP-13 is required to be entered and a Rx trip is initiated >5% power. If the applicant thought the Rx was tripped, then the dumps would be controlling at 547°F and a FWI signal would be generated. This would directly affect the bypass valves. Since the bypass valves are 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 raise SGWL 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 two particular valves.

Distracter analysis A. Incorrect. First part is incorrect (See C.1). Plausible if the applicant thinks that 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 thinks that a FWI has occurred.

C. Correct. First part is correct. D175073, Sheet 1 shows that these valves close on a SGFP trip.

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

A FWI occurs with a P-4 signal (Rx Trip) coincident with a low Tavg, 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 and interpret the following as they apply to the Loss of Main Feedwater (MFW): Status of MFW pumps, regulating and stop valves Importance Rating: 3.5 3.7 Technical

Reference:

D-175073, SH 1, Main Feedwater System, Ver 18 FSD-181007, Reactor Protection System, Ver 18 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if entry into AOP-13, Loss of Main Feedwater is required.

(OPS-52520M02)

Question History: NEW K/A match: Requires the applicant to determine the status of feed system STOP valves and bypass FCVs upon a loss of Main Feedwater.

SRO justification: N/A

36. 055A3.03 036 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)

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

D170064/D-175027:

These drawings show that R-15A is upstream of the normally off service SJAE filtration system. R-15B is downstream of the SJAE filtration system. When the SJAE filtration sytem is placed on service, R-15B reading will decrease. The SJAE filtration system discharges directly to the turbine building roof and cannot be diverted elsewhere.

Distracter analysis A. Incorrect. First part is correct (See C.1).

Second part is incorrect (See C.2). Plausible since the turbine building 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 outside atmosphere. The SJAE filtration system discharges directly to the turbine building roof and cannot be diverted elsewhere. The Penetration Room Filtration system on the rad side does have recirc alignment MOVs and a student could confuse the two systems or apply the concepts from one system to the other.

B. Incorrect. First part is incorrect (See C.1). Plausible if the applicant cannot recall the location of R-15B and believes it is upstream of the SJAE filtration system. R-15A is located before the SJAE filtration system and 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 filtration system. When the SJAE filtration system is placed on service, R-15B reading will decrease.

Second part is correct. The SJAE filtration system discharges directly to the turbine building roof and cannot be diverted elsewhere.

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 monitor automatic operation of the CARS, including: Automatic diversion of CARS exhaust Importance Rating: 2.5* 2.7*

Technical

Reference:

D-170064, SH1, Condenser Vacuum System, Ver 19 D-175027, SH 1, HVAC: TUBINE BLDG, Ver 21 References provided: None Learning Objective: LABEL, DRAW AND ILLUSTRATE the Condensate and Feedwater System flow paths, to include the components on the 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 and determine its response when the SJAE filtration system is placed on service. FNP has no automatic diversion of the SJAE exhaust system. (10/24/12) Chief Examiner said using manual diversion based on our plant design is acceptable.

SRO justification: N/A

37. 056AK3.02 037 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 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 at the SG pressure specified in ECP-0.0 is to prevent (2) .

(1) (2)

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

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 pressure is above the minimum pressure to preclude injection of accumulator nitrogen into the RCS. The target SG pressure should be based on the nominal SG pressure to preclude nitrogen addition, plus margin for controllability (e.g., 100 psi).

Distracter analysis A. Correct. First part is correct. Per ECP- 0.0, Reduce intact SGs pressure to 260 psig.

Second part is correct: Per ECB-0.0, [...] Should ensure that RCS pressure is above the minimum pressure to preclude injection of accumulator nitrogen into the RCS.

B. Incorrect. First part is incorrect (See A.1). Plausible since the limit in the background document is 160 psig. The limit in the procedure adds a 100 psig for margin of controllability and the applicant could confuse these two numbers and believe that the depressurization must 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 the reason for the Tcold temperature limit of 280°F during the pressure reduction 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 the following responses as they apply to the Loss of Offsite Power: 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 3.1 References provided: None Learning Objective: STATE AND EXPLAIN the basis for all Cautions, Notes, and Actions 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 Offsite Power occurs and the Emergency DGs fail to energize the ESF busses. The Applicant is required to know the reason that the secondary depressurization is stopped at 260 psig (reasons for the actions contained in the EOP).

SRO justification: N/A

38. 059A4.01 038 Unit 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)

A. LIT LIT B. LIT NOT lit C. NOT lit LIT D. NOT lit NOT lit U-161792 - Tab 3, Section 5, Page 1 When a trip condition occurs, signals from the electronic controller close the steam valves.

Ran on desktop simulator. Inserted SGFP trip from 100% power and both governor valves went closed.

From OPS-52104C Ver 2 pg 15 -

Initially, as the feed pump turbine accelerates from operation on the turning gear to operating speed, both the LP and HP stop valves are open. The first governor valve to open on an increase speed signal from the control system is the LP governor. Since reheat steam is not available, the turbine speed does not increase. Once the LP governor 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 steam supply via the HP governor valve.

As main turbine load is increased, reheat steam pressure in the shell side of the MSRs also increases. At approximately 25 percent main turbine power, the reheat steam pressure is high enough to cause the feed pump turbine speed to increase. In an effort to maintain the desired feed pump turbine speed, the control system begins to shut the HP governor valve. Once the HP governor valve approaches the fully shut position, the control system starts closing the LP governor valves. During 100 percent power operation, 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 feed pump turbine speed.

Some validators selected the correct answer but stated that they struggled with determining the response of governor valves when the SGFP tripped.

Distracter analysis

A. Correct. First part is correct. The Servo controller shuts the HP governor valve when the SGFP turbine trips.

Second part is correct. The Servo controller shuts the LP governor valve when the SGFP turbine trips.

B. Incorrect. First part is correct (See A.1). Logical connection to the second part 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 thinks that the HP and LP Governor Valves remain in their pre-trip positions on a SGFP trip because the HP and LP STOP valves go shut.

C. Incorrect. First part in incorrect (See A.1). Plausible if the applicant thinks that the HP and LP Governor Valves remain in their pre-trip positions on a SGFP trip because the HP and LP STOP valves go shut. If the applicant thought the HP governor valve controlled speed 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 part if the applicant thinks that the HP governor valve controls speed at high power.

D. Incorrect. First part is incorrect (See A.1) Plausible if the applicant thinks that the HP and LP Governor Valves remain in their pre-trip positions on a SGFP trip because the HP and LP STOP valves go shut.

Logical connection to the second part if the applicant thinks that both the HP and LP governor valves are open at 100% power.

Second part is incorrect (See A.2) Plausible if the applicant thinks that the HP and LP Governor Valves remain in their pre-trip positions on a SGFP trip because the HP and LP STOP valves go shut. Logical connection to the first part if the applicant thinks that both the HP and LP governor valves are open at 100% power.

K/A: 059A4.01 Main Feedwater (MFW) System - Ability to manually operate and monitor in the control room: MFW turbine trip indication Importance Rating: 3.1* 3.1*

Technical

Reference:

U-161792, SGFP Drive Turbine and Accessories, Ver 12 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if entry into AOP-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 and determine the proper SGFP GOV valve positions on a SGFP trip.

SRO justification: N/A

39. 059AK2.01 039 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.

A. HV-7614A/B/C, 2A/B/C SGBD ISO B. FCV-1152, SGB INLET STOP VALVE C. RCV-023B, SGBD DISCH TO ENVIRONMENT D. HV-7697A/B, 7698A/B and 7699A/B, 2A/B/C SGBD ISO

FH1 Automatic Actions:

R23A :( Steam Generator Blowdown Processing) closes 2-BD-FCV-1152 S/G Blowdown Heat Exchanger Discharge Valve.

R-23A is in the SGBD line after FCV-1152 but before the SGBD Surge tank. R-23B is after the surge tank and is the last rad monitor and isolation signal before an accidental release would make it to the environment. R-23 A and B are often confused on the functions and locations. Two other sets of valves isolate SGBD due to other conditions, such as High Penetration room pressure and AFW autostart. All of these valves complete the same function but for different reasons.

Distracter analysis A. Incorrect. See B. Plausible since these valves will isolate SG Blowdown (SGBD) and automatically close on an AFW autostart. The applicant could believe they also close on a high radiation signal.

B. Correct. R-23A automatically closes FCV-1152. FCV-1152 also closes on SGBD high ST level, High pressure in the SGBD system and High flow.

C. Incorrect. See B. Plausible since R-23B automatically closes RCV-23B and the applicant could confuse which radiation monitor closes which valve. RCV-023B will isolate SGBD to the environment and is downstream of FCV-1152.

D. Incorrect. See B. Plausible since these valves are two series isolation valves located 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 the interrelations between the Accidental Liquid Radwaste Release and the following: Radioactive-liquid monitors Importance Rating: 2.7 2.8 Technical

Reference:

FNP-2-ARP-1.6, FH1 - RMS HI-RAD, Ver 70 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Radiation Monitoring 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 accidental liquid radwaste release. The applicant is required to know the interrelations between R-23A and the SGBD system that will terminate the accidental liquid radwaste release.

SRO justification: N/A

40. 059G2.2.44 040 Unit 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 are observed:

KB4, SGFP SUCTION PRESS LOW, comes into alarm and the operating crew observes the following on PR4039, SGFP SUCT PRESS:

Time 0 sec 10 sec 20 sec 30 sec 40 sec 300 psig 275 psig 265 psig 270 psig 285 psig At 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) .

A. 1) will NOT

2) rapidly reduce Turbine load using AOP-17.1, Rapid Turbine Power Reduction B. 1) will NOT

2) check SGFP suction pressure stabilizes C. 1) WILL

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

D. 1) WILL

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

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-13 6.1 Check SGFP suction pressure stabilizes above 275 psig.

6.1.2 RNO:

IF suction pressure still falling, THEN reduce turbine load rapidly using FNP-1-AOP-17.1, RAPID TURBINE POWER REDUCTION.

Distracter analysis A. Incorrect. First part is correct (See B.1).

Second part is incorrect (See B.2). Plausible since this is the action to 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 suction pressure 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 the SGFP from tripping at time 40 sec C. Incorrect. First part is incorrect (See B.1) Plausible if the applicant confuses the condensate pump autostart setpoint with the SGFP trip. The standby condensate pump, if in AUTO, would start 10 seconds after SGFP suction pressure falls below 275 psig.

OR plausible if the applicant thought that when the low pressure alarm comes in the SGFP would trip 30 sec later.

Second part is incorrect (See B.2) Plausible since this is the correct 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 was the correct actions 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 room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Importance Rating: 4.2 4.4 Technical

Reference:

FNP-1-AOP-13, Condensate and Feedwater Malfunction, Ver 33 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if any system components 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 SGFP suction 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 that the suction pressure is rising and not reduce power or trip the reactor.

SRO justification: N/A

41. 061A3.01 041 Unit 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)

A. will NOT 350 B. will NOT 700 C. WILL 700 D. WILL 1050 Not a true 2+2 question to improve distracter plausibility.

The applicant has to evaluate how the loss of the 1A Startup transformer (SUT) affects the TDAFW and MDAFW pumps auto starts. Since the 1A SUT powers the 1A bus and the 1B SUT powers the 1B and 1C busses, only the 1A bus is lost. The opposite is true on 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 powered from 1B SUT. The 1A MDAFW pump did lose power and will be sequenced on the bus when the DG starts and the LOSP sequencer runs.

Thirdly, once the applicant determines which AFW pumps are running, then they will have to recall design flow rates for each (350 gpm for the MDAFW and 700 gpm for the TDAFW pump) to determine total approximate flow.

Distracter analysis A. Correct. First part is correct. FSD-A181010 - 3.9.2.3 - The TDAFW pump shall start by opening the steam supply valves to the turbine drive on a loss of power signal, low-low water level signals from two out of three level transmitters of any two out of three steam generators, or an AMSAC signal. The loss of power signal comes from 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 and the 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 confuses the 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 applicant doesn't recall that the MDAFWP receives an auto start signal during an LOSP. This would make this a logical connection to the first part and a correct answer if the applicant thought that only the TDAFW pump started.

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

Second part is incorrect (See A.2).Plausible since this is the design flow for one MDAFW pump and the TDAFW pump and a logical connection to the first part if the applicant thought that the TDAFW pump started.

K/A: 061A3.01 Auxiliary / Emergency Feedwater (AFW) System - Ability to monitor automatic operation of the AFW, including: AFW startup and flows Importance 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 74 References provided: None Learning Objective: RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the AFW System to include the components found on Figure 2, Auxiliary Feedwater System, Figure 3, TDAFWP Steam Supply, and Figure 4, Air Supply to TDAFWP Steam Admission Valves (OPS-40201D02).

NAME AND IDENTIFY the Bus power supplies (Off-site sources and Emergency source-to- Load), for those electrical components associated with the AFW System to include those items in Table 3- Power Supplies (OPS-40201D04).

SELECT AND ASSESS the AFW System instrument/equipment response expected when performing auxiliary feedwater evolutions including (OPS-52102H05):

[...]

The Failed Condition

[...]

Associated Trip Setpoint(s)

[...]

Question History: NEW K/A match: The applicant is required to evaluate the loss of power and determine which AFW pump auto starts (monitor startup) and the resultant flow (monitor flow).

SRO justification: N/A

42. 061K5.01 042 Unit 1 was operating at 100% power when a Reactor Trip occurred and the following conditions 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 to address the cooldown?

A. Minimize total AFW flow.

B. Emergency borate the RCS.

C. Close all MSIVs and MSIV Bypass Valves.

D. Manually initiate SI and return to EEP-0.0, Reactor Trip or Safety Injection.

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 analysis A. Correct Step 1.1.4 RNO of ESP-0.1 has the operator minimize AFW to stop the cooldown.

B. Incorrect. See A. Plausible since this is an action in ESP-0.1 if Tavg falls below 525°F. This action is at step 4 and would not be required since Tavg is >525°F. The applicant could confuse this temperature 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 that equates to steam in the TB reduced, and stm dumps checked.

D. Incorrect. See A. Plausible because the Pzr level meets the SI reinitiation criteria (13%) for a number of other Emergency procedures (such as ESP-1.1) and the applicant could confuse it with the correct Pzr level 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 following concepts as the apply to the AFW: Relationship between AFW flow and RCS heat transfer Importance Rating: 3.6 3.9 Technical

Reference:

ESP-0.1, Reactor trip Response, Ver 32.

References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if any system components 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 the excessive RCS cooldown (operational implication), they must know that reducing AFW flow will reduce the heat transfer rate of the RCS.

SRO justification: N/A

43. 062AA1.07 043 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 has dropped to 179 feet, 10 inches.