ML033280742

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Initial Submittal of the Written Exam for the Prairie Island Initial Examination September 2003
ML033280742
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 09/09/2003
From: Cooper P
Nuclear Management Co
To:
Office of Nuclear Reactor Regulation
References
50-282/03-301, 50-306/03-301
Download: ML033280742 (56)
v • d • e


Text

Page 1 of 25

1. The following conditions are observed about 3 minutes after an automatic safety injection:
  • -Core exit T/Cs 540°F
  • -Pressurizer level 58% and increasing
  • -RCS pressure 1100 psig and decreasing
  • -Containment pressure 0 psig Based on these indications, it is likely that a
a. small break LOCA has occurred outside the containment.
b. a pressurizer PORV is stuck open.
c. steam generator tube rupture has occurred.
d. steam line break has occurred outside containment.

Page 2 of 25

2. You are responding to a small break LOCA in accordance with ES-1.1, Post-LOCA Cooldown and Depressurization when the Shift Manager provides the following update: RCS conditions have met ORANGE path conditions on the Integrity Critical Safety Function Status Tree.

What actions would be taken in response to an ORANGE path condition being met?

a. Restart any SI pump that was previously stopped.
b. The RCS cooldown should be suspended.
c. RCS pressure should be raised to maximize subcooling.
d. One RCP should be stopped if both RCPs are running.

Page 3 of 25

3. Why is RWST level of 33% a criterion for shifting ECCS into the recirculation mode?
a. Level below 33% does NOT provide adequate NPSH for 2 trains of ECCS and CS pumps.
b. Vortexing may occur in the RWST (below 33%) with full ECCS and CS system flow rates.
c. There will NOT be adequate level in containment sump B until the RWST level is below 33%.
d. With instrument uncertainty, 33% is lowest level reading that ensures adequate NPSH for one train of ECCS and CS pumps.

Page 4 of 25

4. If component cooling water flow is lost to an RCP, what parameters will change?

Temperatures will rise on the

a. RCP motor radial and thrust bearings and the pump radial bearing.
b. RCP motor stator and the motor radial and thrust bearings.
c. RCP motor stator and the pump radial bearing.
d. RCP motor radial and thrust bearings.

Page 5 of 25

5. The position of CV-31158, Charging Line Flow Control is changed to vary RCP seal injection flow. IF CV-31158 is closed slightly, THEN Charging Pump Header Pressure RCP Seal Water Injection Flow Charging Flow to Regen HX
a. Increases Increases Decreases
b. Increases Decreases Increases
c. Decreases Increases Decreases
d. Decreases Decreases Increases

Page 6 of 25

6. References available:
  • FIG C1-31, Boiling Curve
  • FIG C1-32, Boiling Curve
  • FIG C1-33, Time to Boiling After Pool Flood Given the following:
  • It is presently 9 am on Sept. 19, 2003
  • Unit 2 is in Mode 5
  • RCS temperature is 140°F.
  • The unit was shutdown following a 300-day continuous power run.
  • The timeline for recent Unit 2 operations includes:

- 09/14/03, 1000 Reactor shutdown.

- 09/14/03, 1100 Cooldown initiated for REFUELING outage.

- 09/14/03, 1900 Entered MODE 5.

- 09/16/03, 0800 Reactor vessel head removed.

- 09/16/03, 1330 Normal level established in refueling cavity.

If RHR cooling is lost, how much time (in minutes) is available before boiling will occur?

a. 12.5
b. 14
c. 180
d. 440

Page 7 of 25

7. If the pressurizer spray valves are not available for the RCS depressurization in E-3 for a steam generator tube rupture, why are the pressurizer PORVs the next preferred method of RCS depressurization ahead of using auxiliary spray?
a. Some of the excess RCS inventory would be removed.
b. RCS depressurization would be more controllable.
c. RCS boron concentration would not be diluted by pressurizer out surge.
d. Pressurizer spray nozzle failure would be precluded.

Page 8 of 25

8. Assume Unit 2 is operating in Mode 1 and the reactor trip breakers have failed such that NO automatic signal will cause them to open.

What initiating event would be considered an ATWS AND cause AMSAC/DSS to automatically actuate?

a. With reactor power at 8%, the main turbine inadvertently trips during turbine roll-up.
b. With reactor power at 28%, 11 SG level drops to 10% in the narrow range.
c. With reactor power at 8%, the RO inadvertently unblocks and energizes the Source Range NIs.
d. With reactor power at 28%, 22 RCP shaft seizes causing an overload trip of its supply breaker.

Page 9 of 25

9. Refer to Exhibit 1: Steam Dump Photos Reference available: 1E-3, Steam Generator Tube Rupture, Pages 7 and 8.

The LEAD has performed Step 7 of E-3 and the steam dump controls are positioned as shown in Figure 1. The MSIV for the unaffected SG is open. RCS Tavg is 538°F.

You have been asked to verify the setup. What would you report?

a. Setup correct. Maximum rate cooldown is in progress.
b. Disagree with setup. Steam Dump Mode switch should be in STM PRESS.
c. Disagree with setup. Lo-Lo Tavg Interlock switches should be in BYPASS INTLK.
d. Disagree with setup. Controller 1HC484 should be in AUTO with setpoint at 40%.

Page 10 of 25

10. If a pipe elbow blows out on the steam supply to the air ejectors near the inlet to the air ejector, which alarm would be unexpected for this steam line failure?
a. 47008-0209, CONDENSER HI PRESS
b. 47022-0204, TURBINE BUILDING STM EXCLUSION ACTUATED
c. 47022-0305, 122 FIRE PUMP (DIESEL) RUNNING
d. 47022-0611, FIRE DETECTION PANEL FP121 FIRE ALARM

Page 11 of 25

11. What is the basis for the S/G low level reactor trip?

The S/G low level reactor trip

a. prevents RCS and SG pressures from exceeding 110% of their design values.
b. ensures DNBR remains <1.3 during a loss of feedwater accident.
c. provides protection against a loss of heat sink.
d. limits thermal stresses on the SG shell and tube sheet when AFW flow is actuated.

Page 12 of 25

12. The coping study that was done to support the analysis for a loss of all AC power event at Prairie Island took credit for the condensate storage tanks (CSTs). The minimum volume of water in the CSTs ensures the affected unit can be
a. placed in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
b. maintained in HOT STANDBY for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c. cooled down to MODE 4 as long as the cool down rate does NOT exceed 25°F/hour.
d. cooled for at least 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> AND cooled down to MODE 4 during that 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> period.

Page 13 of 25

13. Refer to Exhibit 3: 121 Air Compressor Control Switch What would be the indications on this control switch if Bus 15 locked out?
a. There would be no change since this air compressor is ultimately powered by Bus 16.
b. The red light would be off with the green and white lights on.
c. The red and white lights would be off with the green light on.
d. All lights would be off.

Page 14 of 25

14. Given the following Unit 1 plant conditions:
  • The unit was at 50% for turbine valve testing
  • A load increase is in progress per 1C1.4 Power Operation
  • Power is currently at 72%
  • A lockout has occurred on 480V Bus 122
  • No operator actions have been taken.

What is the status of pressurizer heater groups?

a. All groups are energized.
b. All groups are de-energized.
c. Groups B, C, D, and E are energized AND group A heaters are de-energized.
d. Groups A, C, D, and E are energized AND group B heaters are de-energized.

Page 15 of 25

15. A NOTE in 2C20.9 AOP1, Loss of Unit 2 Train A DC states that the Main Generator Output Breakers will NOT open automatically following the turbine trip. What prevents this normally automatic action AND is a compensatory manual action required?
a. No breaker control power but no compensatory action is required due to breaker failure logic.
b. No breaker control power so the breakers must be opened locally.
c. Generator lockout relay does not actuate so the breakers must be opened manually.
d. Generator lockout relay does not actuate but no compensatory action is required due to breaker failure logic.

Page 16 of 25

16. You are assigned to a Unit which is operating at 100% power.

IF a rupture occurs in the Instrument Air system, you should monitor plant conditions and initiate a manual reactor trip if plant conditions approach any automatic reactor trip setpoint.

Which plant parameter is going to reach its automatic reactor trip setpoint FIRST for this event?

a. Pressurizer level
b. Steam generator level
c. Pressurizer pressure
d. RCS loop DT (OTDT)

Page 17 of 25

17. A loss of all feed water has occurred. You are implementing FR-H.1, Response to Loss of Secondary Heat Sink. While depressurizing SGs to establish condensate feed, you receive alarm C47010-0101, Condensate Storage Tank Lo-Lo Level.

What impact does this have on your (the crews) actions?

a. We must align an alternate water supply to the AFW pumps using C28.1 AOP2, Loss of Condensate Supply to Aux Feed Pump Suction.
b. This alarm means we cant use the condensate pumps and must go on to the next mitigating strategy in the EOPs.
c. This has no immediate impact because we are trying to establish flow using the condensate pumps.
d. This has no impact because it is an expected alarm. The condensate makeup valve fails open dumping the CST to the main condensers.

Page 18 of 25

18. Identify the strategy that is NOT used in ECA-1.1, Loss of Emergency Coolant Recirculation to cope with a loss of emergency coolant recirculation capability during a loss of coolant accident.
a. Starting a reactor coolant pump to establish forced circulation in the RCS.
b. Starting all available CRDM fans to maximize cooling to the reactor vessel head.
c. Depressurizing intact steam generators by dumping steam,
d. Reducing ECCS injection flow from the RWST to match RCS coolant loss.

Page 19 of 25

19. If Unit 1 is operating at 75% power, what event will cause a continuous control rod withdrawal in automatic?
a. MV-32086, Emergency Boration to Charging Pump Suction valve partially (<5%) open.
b. Electrical fault which causes an RCS Loop B cold leg RTD in the bypass loop to fail high.
c. A continuous turbine load runback.
d. Detector failure resulting in N44, Power Range NI, failing high

Page 20 of 25

20. Unit 2 is performing a shutdown. Power is 3% when alarm 47013-0602, N36 Intermediate Range Loss of Comp Voltage comes in.

How does this affect the Nuclear Instrumentation?

a. N36 reading would immediately drop about 1 decade. During the subsequent shutdown, SR NIs will energize automatically when N35 drops below P-6 setpoint.
b. N36 reading would immediately rise about 1 decade. During the subsequent shutdown, SR NIs will NOT energize automatically because N36 reading will remain above the P-6 setpoint.
c. N36 reading would NOT immediately change. During the subsequent shutdown, SR NIs will energize automatically when N35 drops below P-6 setpoint.
d. N36 reading would NOT immediately change. During the subsequent shutdown, SR NIs will NOT energize automatically because N36 reading will remain above the P-6 setpoint.

Page 21 of 25

21. You are the LEAD on Unit 2 and had to evacuate the control room due to smoke accumulation.

After completing Attachment G of F5 Appendix B, you report to the Unit 2 SS at the Hot Shutdown Panel for further duties.

You are directed to verify Natural Circulation has been established. Conditions are:

  • RCS pressure is 2235 psig
  • Tcold is stable at 510°F
  • Thot is 565°F and increasing
  • Both 21 and 22 SG have pressure at 1005 psig
  • 21 SG WR level is 65%
  • 22 SG WR level is 64%

Natural circulation is

a. established and current steam and feed flows should be maintained.
b. established but steam release should be increased.
c. NOT established and steam release should be increased.
d. NOT established and feed flow should be increased to raise SG levels.

Page 22 of 25

22. If a valid R-9 alarm occurs due to excessive Reactor Coolant System activity, Tavg is reduced to below 500°F in order to...
a. increase the safety margin for fuel clad integrity during any design basis accident.
b. ensure RCS saturation pressure is below the setpoint of the SG safety valves.
c. reduce the atmospheric dispersion of fission products during a loss of coolant accident.
d. lower peak containment pressure and containment leakage if a loss of coolant accident occurs.

Page 23 of 25

23. Unit 1 was operating at 100% power when a spurious SI occurred.

All systems respond as designed with the exception of the B reactor trip breaker, which did NOT open and remains closed. The crew has transitioned to 1ES-0.2, "SI TERMINATION. The following conditions exist:

  • RCS pressure is 2200 psig and slowly rising.
  • Containment pressure is 0.1 psig.
  • MSIVs are open and SG pressure is 990 psig When step 1 is performed and both SI pushbuttons have been momentarily pushed in, what will be the status of SI?
a. Both trains of SI will be reset with automatic SI initiation blocked.
b. Both trains of SI will be reset but only the "A" train automatic SI initiation will be blocked.
c. Only the A train of SI will be reset with automatic SI initiation blocked.
d. Only the A train of SI will be reset but neither trains automatic SI initiation will be blocked.

Page 24 of 25

24. Reference available: F-0.2, Core Cooling A small break LOCA occurred about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ago and mitigating actions were performed as directed by the appropriate EOPs. The current plant conditions are:
  • RCS pressure is 280 psig
  • Pressurizer level is 45%.
  • Overpressure Protection System is enabled
  • Both RCPs are stopped
  • RCS cooldown rate is about 10°F/hr.
  • Procedure ES-1.1, Post LOCA Cooldown and Depressurization is still being used Assuming the Core Cooling Critical Safety Function Status Tree (CSFST) is currently GREEN, what failure could result in a Yellow path condition on the Core Cooling CSFST?
a. Pressurizer spray valve fails OPEN due to a controller failure
b. Fuse blows (opens) in DC power supply to CV-31226, LETDOWN LINE ISOLATION
c. RCS wide range pressure instrument PT-420 fails HIGH
d. Air supply line breaks to valve CV-31235, 11 RHR HX OUTLET FLOW CONTROL

Page 25 of 25

25. Given the following conditions:
  • The faulted SG is isolated.
  • RED PATH conditions exist on the Integrity Status Tree.
  • The actions of FR-P.1, "Response To Imminent Pressurized Thermal Shock Condition" are being performed
  • RCS temperature soak is required and has been initiated (Step 23)
  • NO RCPs are running Which evolution can be performed during the one hour soak period?
a. Place normal letdown in service.
b. Lower the non-faulted SG PORV controller pressure setpoint 25 psi.
c. Raise AFW flow to and establish SG blowdown from the non-faulted SG.
d. Raise RCS pressure to the middle of the pressure band allowed by Figure FRP1-1.

Page 26 of 25

26. FR-H.2, Response to Steam Generator Overpressure, may be implemented when SG pressure is above _____ psig because pressure is above the __________ setpoint.
a. 1005, steam dump controller
b. 1050, SG PORV controller
c. 1077, lowest SG safety
d. 1131, highest SG safety

Page 27 of 25

27. During a LOCA, at what containment pressure would we declare adverse containment conditions AND why is this a concern?

Containment conditions FIRST become adverse when containment pressure is greater than

a. 4 psig because instruments inside containment become less accurate due to high ambient temperature associated with higher pressures.
b. 4 psig because instruments inside containment become less accurate due to the external pressure on their casings.
c. 5 psig because instruments inside containment become less accurate due to high ambient temperature associated with higher pressures.
d. 5 psig because instruments inside containment become less accurate due to the external pressure on their casings.

Page 28 of 25

28. What is the power supply to 11 Reactor Coolant Pump?
a. Bus 11
b. Bus 12
c. Bus 15
d. Bus 16

Page 29 of 25

29. If the water level drops in the Boron Concentration Measuring System shield tank, what type radiation hazard would exist in the vicinity of the tank?
a. Alpha
b. Beta
c. Gamma
d. Neutron

Page 30 of 25

30. On the current outage schedule, the RCS cooldown (from 340°F to <200°F) is supposed to occur in the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. You have just finished placing RHR in service for Phase II cooldown but during this evolution you determined the 11 RHR HX had a tube leak. The 11 RHR HX is currently isolated. How will this impact the RCS cooldown on the outage schedule?
a. The cooldown will be completed on schedule because the RHR system has two 100%

redundant trains.

b. The cooldown will be completed on schedule because the SGs will do most of the cooling until RCS temperature is below 212°F.
c. The RCS cooldown can not be completed until decay heat level drops below the capacity of the single RHR train.
d. The RCS cooldown will be completed but it will take about twice as long as scheduled.

Page 31 of 25

31. A small break LOCA is in progress on Unit 1 with the following conditions:
  • All ECCS Pumps are operating except 11 SI pump which failed to start
  • RCS pressure is 1700 psig and stable
  • Pressurizer level is 5% and stable.
  • RCS average temperature is slowly lowering about 5°F/hour
  • 11 SI pump is now available for operation IF 11 SI pump is started, THEN the new equilibrium RCS pressure will be ________ than before and break flow will be _________ than before.
a. higher; higher
b. higher; the same
c. the same; the same
d. the same; higher

Page 32 of 25

32. The following conditions occurred in Unit 1:

Time:

1000 hrs 1100 hrs PRT level 72%

78%

PRT temperature 96°F 96°F Pressurizer level 45%

43%

Tavg 570°F 569°F Containment temperature 102°F 108°F What is the MOST LIKELY cause of the PRT level increase?

a. Expansion due to containment heatup
b. Pressurizer PORV leakage
c. Letdown relief valve (inside containment) leakage
d. Seal return relief valve (inside containment) leakage

Page 33 of 25 Page 34 of 25

33. Both Units are operating in MODE 1 at 100% power.

The CC system in each unit is in its normal alignment.

Predict the change in surge tank levels if a single tube rupture occurs in the Unit 1 letdown heat exchanger.

Unit 1 Surge Tank Level Unit 2 Surge Tank Level

a. Increases Increases
b. Increases No Change
c. Decreases Decreases
d. Decreases No Change

Page 35 of 25

34. Reference Available: ES-0.3A, Natural Circulation Cooldown with CRDM Fans When preparing for a cooldown, the Unit is borated to establish adequate shutdown margin for cold shutdown. This concept is true for a normal cooldown AND a natural circulation (NC) cooldown guided by ES-0.3A, Natural Circulation Cooldown with CRDM Fans.

IF 400 gallons of boric acid must be added to the RCS in either cooldown, compare the relationship between RCS and pressurizer boron concentrations in each cooldown case after the FIRST 200 gallons of boric acid has been added. You can assume that RCS and pressurizer boron concentrations were the same at the start of the boration in each case.

a. NC cooldown boration will have less difference between RCS and pressurizer boron concentrations AND RCS boron concentration > pressurizer boron concentration.
b. NC cooldown boration will have less difference between RCS and pressurizer boron concentrations AND RCS boron concentration < pressurizer boron concentration.
c. NC cooldown boration will have more difference between RCS and pressurizer boron concentrations AND RCS boron concentration > pressurizer boron concentration.
d. NC cooldown boration will have more difference between RCS and pressurizer boron concentrations AND RCS boron concentration < pressurizer boron concentration.

Page 36 of 25

35. While establishing a bubble in the pressurizer per 1C1.2, Unit 1 Startup Procedure, CV-31203, Letdown Pressure Control valve opens progressively in AUTO.

CV-31203 slowly opens

a. because RCS temperature is slowly lowering.
b. due to thermal expansion of PRZR liquid.
c. because the PRZR spray valves are slowly closed while drawing a bubble.
d. due to the switchover of letdown to orifices from RHR-CVCS cross-connect.

Page 37 of 25

36. Which of the following Reactor Protection System trips is listed in the Technical Specifications Bases as protecting against Departure from Nucleate Boiling (DNB) accidents?
a. Source Range High Flux
b. High Pressurizer Pressure
c. High Pressurizer Level
d. Power Range Negative Flux Rate

Page 38 of 25

37. Reference available: 1C51.4 , Containment Pressure 1P-950 - High The yellow containment pressure channel 1PI-950 has failed high.

Step 2 places the toggle switch for bistable PC950B, Hi-Hi Containment Spray in the UP position.

What is the result of placing this switch to the UP position?

a. Energizes the bistable preventing the yellow channel from generating a 'P' signal input.
b. De-energizes the bistable causing a yellow channel input into the 'P' signal coincidence logic.
c. Energizes the input relays causing a yellow channel input into the 'P' signal coincidence logic.
d. De-energizes the input relays preventing the yellow channel from generating a 'P' signal input.

Page 39 of 25

38. During the month of August, the Unit 2 containment fan cooling units are being supplied from Containment Chilled Water. The Containment Chillers trip and the CFCUs switch to cooling water. The cooling water inlet temperature is 95°F.

a) What changes would you expect to see in the containment parameters?

b) Which parameter will be the MOST limiting in terms of continued operation of Unit 2?

Affected Parameters Most Critical

a. Increases in temperature and pressure pressure
b. Increases in temperature and pressure temperature
c. Increase in pressure and decrease in humidity pressure
d. Increase in temperature and decrease in humidity temperature

Page 40 of 25

39. Assume the associated Relay Logic Cabinet loses power due to a fault in 125v DC Panel 15 and that the loss of DC power affects ONLY the associated Relay Logic Cabinet.

HOW would this failure affect the operation of the containment spray system if a large break LOCA occurred inside the Unit 1 containment at the same time? WHAT manual action(s) should you take, if any?

a. Containment Spray Train A equipment would NOT actuate. The spray pump could be manually started but the discharge valve would have to be locally opened.
b. Containment Spray Train B equipment would NOT actuate. The spray pump could be manually started but the discharge valve would have to be locally opened.
c. Both trains of containment spray would be actuated. Train A would have actuated when power was lost. No immediate manual actions are needed.
d. Both trains of containment spray would be actuated. Train B would have actuated when power was lost. No immediate manual actions are needed.

Page 41 of 25

40. IF one containment spray pump is inoperable, THEN what is the MINIMUM containment cooling required to mitigate a design basis steam line break inside containment?

The other train of Containment Spray

a. OR BOTH trains of CFCUs.
b. AND ONE train of CFCUs.
c. AND three of the four CFCUs.
d. AND BOTH trains of CFCUs.

Page 42 of 25

41. What are the immediate actions of E-0 associated with the main turbine?

Verify Turbine Trip:

a. a. Both turbine stop valves - CLOSED a. Manually trip turbine.

IF NOT, THEN manually close control valves.

IF NOT, THEN manually close MSIVs and bypass valves.

b. a. Both turbine stop valves - CLOSED a. Manually trip turbine.

IF NOT, THEN manually stop both EHC pumps.

IF stop valves are NOT closed, THEN manually close MSIVs and bypass valves.

c. a. All turbine control valves - CLOSED a. Manually trip turbine.

IF NOT, THEN manually close control valves.

IF NOT, THEN manually close MSIVs and bypass valves.

d. a. All turbine control valves - CLOSED a. Manually trip turbine.

IF NOT, THEN manually stop both EHC pumps.

IF control valves are NOT closed, THEN manually close MSIVs and bypass valves.

Page 43 of 25

42. Unit 2 is operating at 6% reactor power in preparation for rolling the main turbine.

Starting the turbine roll-up will have what affect on the steam dump system?

a. The steam dumps will open if the roll-up arms the loss of load permissive.
b. The steam dumps will open due to the change in RCS Tavg.
c. The steam dumps will close if RCS Tavg drops below 545°F.
d. The steam dumps will close due to the change in steam header pressure.

Page 44 of 25

43. Given the following conditions on Unit 1:
  • Reactor power is 80% (460MWe)
  • 13 Condensate Pump is in standby
  • Annunciator 47009-0103, 12 CONDENSATE PUMP LOCKED OUT alarms What is the plant response to this condition and what general action is appropriate?
a. 13 Condensate Pump AUTO STARTS. Power may be maintained at 80% (460MWe).
b. 11 Feedwater Pump TRIPS. Power must be reduced below 60% (330 MWe).
c. BOTH Feedwater Pumps TRIP. E-0, Reactor Trip or Safety Injection must be implemented.
d. CV-31087, Condensate Bypass to Feedwater Pump, AUTOMATICALLY OPENS. Power must be reduced below 50% (287MWe).

Page 45 of 25

44. Given the following:
  • Reactor power is 8%
  • Turbine is rolling at 1800 rpm but output breakers are OPEN.
  • 11 SG narrow range level is 70%
  • 12 SG narrow range level is 75%

List the AUTOMATIC actions that will occur from the above conditions.

a. Turbine trip, Reactor trip and FRV & bypass valves close.
b. Turbine trip, Reactor trip and Feedwater pumps trip.
c. FRV & bypass valves close and Feedwater pumps trip.
d. Turbine trip, Feedwater pumps trip, AFW pumps start and FRV & bypass valves close

Page 46 of 25

45. Given the following conditions on Unit 1:
  • 83% Reactor power.

Which ONE of the following failures will cause RCS Tavg to INITIALLY INCREASE?

a. 11 SG Red Channel Narrow Range Level (LI-461) fails HIGH.
b. A Bypass FW valve, CV-31369, fails OPEN.
c. 12 FWP Discharge Valve, MV-32324, spuriously closes.
d. 12 MD AFW Pump spuriously starts..

Page 47 of 25

46. Unit 2 conditions:
  • 40% power.
  • All equipment is OPERABLE IF a lockout occurs on 4KV Bus 25, THEN what would be the availability of the Unit 2 Auxiliary Feedwater pumps (#21 and #22)?
a. Neither pump would be available.
b. ONLY 21 AFW pump would be available.
c. ONLY 22 AFW pump would be available.
d. Both pumps would be available.

Page 48 of 25

47. The following Unit 1 conditions exist:
  • Unit 1 has tripped from 15% power due to loss of power to Buses 11 and 12
  • Prior to the trip, all equipment was operable
  • Prior to the trip, BOTH AFW pump were stopped with their control switches in AUTO.
  • Prior to the trip, #12 MFW pump was in service.
  • Prior to the trip, control power was lost to #12 MFW pump
  • Busses 11 and 12 are still de-energized.
  • 11 SG NR level has remained above 20% during this event.
  • 12 SG NR level has remained above 16% during this event.

With NO operator action since the trip, what is the status of the AFW pumps?

  1. 11 AFW pump #12 AFW pump
a. running NOT running
b. NOT running NOT running
c. running running
d. NOT running running

Page 49 of 25

48. IF an electrical fault causes a lockout on Bus 25, Emergency Diesel Generator D5 will
a. NOT start.
b. start but its output breaker will NOT attempt to close.
c. start and its output breaker will close then open and lockout.
d. start and its output breaker will close. D5 will then trip on phase overcurrent.

Page 50 of 25

49. How will a very low-resistance (hard) ground affect the Train A 125V DC Safeguards System?
a. A ground on the + distribution line will cause a short which will blow the fuse closest to the ground. At worst it could cause a complete loss of Train A DC power.
b. A ground on the - distribution line will cause a short which will blow the fuse closest to the ground. At worst it could cause a complete loss of Train A DC power.
c. The Train A DC system would be more susceptible to failure. If another ground should occur on the opposite polarity line, then a short would occur and blow a fuse in the system.
d. The Train A DC system would be more susceptible to failure. If another ground should occur on the same polarity line, then a short would occur and blow a fuse in the system.

Page 51 of 25

50. Refer to Exhibit 2: Bus 15 Load Sequencer Based on the load sequencer panel lights, Bus 15
a. is de-energized. D1 did not start.
b. is de-energized. D1 started and is ready to load.
c. is energized. D1 is supplying power and the sequencer has completed all steps.
d. is energized. D1 is supplying power but the sequencer has NOT run through its steps.

Page 52 of 25

51. You are performing the test of R-18 in preparation for a release of the 121 ADT Monitor Tank.

When you place the OPERATIONAL SELECTOR switch in the CHECK SOURCE position, what should happen?

First, alarm 47022:0209, Rad Monitor Check Source Actuated will come in, then

a. R-18 reading will go to 0 then rise to the check source reading.

No radiation monitor alarms are expected.

b. R-18 reading will go to 0 then rise to the check source reading.

Alarm 47022:0208, Rad Monitor Downscale Failure will come in.

c. R-18 reading will rise from background by an amount equal to the check source reading.

No radiation monitor alarms are expected.

d. R-18 reading will rise from background by an amount equal to the check source reading.

Alarm 47022:0108, Hi Radiation Train A will come in.

Page 53 of 25

52. What Cooling Water System evolution should be preceded by a PA announcement?
a. Removing a traveling screen from service.
b. Opening the emergency bypass gate in the intake canal.
c. Starting the diesel-driven cooling water pump for monthly surveillance.
d. Switching containment fan coolers to Cooling Water from Chilled Water.

Page 54 of 25

53. Following a Unit 1 safety injection actuation, which heat load is isolated from the Cooling Water System?

A. Standby Control Room Chiller Unit B. Unit 2 Containment Fan Cooler Units (switches to Chilled Water)

C. Unit 1 Containment/Auxiliary Building Chillers D. Non-Running Unit 2 Diesel Generators

Page 55 of 25

54. Assume the Unit 2 Main Steam Isolation Valves (MSIVs) are open. What will happen to the MSIVs if instrument air pressure is lost?

The MSIVs will

a. close when air header pressure drops to about 75 psig.
b. remain open if their associated air tanks do NOT have check valve leakage.
c. drift off their open seat but will NOT fully close without a manual or automatic close signal.
d. fail as is (open) and will NOT close on either a manual or automatic close signal.

Page 56 of 25

55. The design of the containment equipment hatch...
a. is sized to allow reactor vessel head "O" ring passage.
b. will allow only 2 people to enter/exit at a time.
c. includes a pneumatic seal to minimize leakage.
d. uses an interlock to prevent both doors being open.

Page 57 of 25

56. Given the following conditions on Unit 1:
  • 60% power and stable
  • PRZR Spray Valves closed in AUTO
  • PRZR Pressure Control Selector Switch is selected to the "2-3 (White-Blue)" position
  • PRZR Pressure Indications are:

- PT-429 (Red) 2230# and rising

- PT-430 (White) 2235# and rising

- PT-431 (Blue) at 2185# and stable

- PT-449 (Yellow) at 2235# and rising Assuming NO operator action is taken and the trend is due to an instrument failure, what is the Pressurizer Pressure Control System response to these conditions?

a. PRZR spray valves will fully open and depressurize the RCS, causing a reactor trip.
b. PRZR pressure will oscillate between 2210 and 2250 psig by the cycling of the heaters.
c. PRZR spray valves will stabilize RCS pressure below 2310 psig.
d. PRZR pressure will oscillate between 2315 and 2335 psig by the cycling of PORV PCV-430.

Page 58 of 25

57. What are the normal power supplies for the Unit 1 charging pumps?

11 Charging Pump 12 Charging Pump 13 Charging Pump

a. Bus 11 Bus 12 Bus 13
b. Bus 11 Bus 12 Bus 11
c. Bus 15 Bus 16 Bus 15
d. Bus 16 Bus 15 Bus 16

Page 59 of 25

58. If power fuse blows (opens) for the rod position indication to the shutdown bank control rod E3, what will change on ERCS?

ERCS will

a. display the last valid rod position but change the color to cyan to indicate questionable data.

There will be no ERCS generated alarms.

b. display UNK for rod position and generate alarm 47013:0507, Computer Alarm Rod Deviation/Sequencing.
c. display 0 for rod position and generate alarm 47013:0507, Computer Alarm Rod Deviation/Sequencing.
d. display 228 for rod position and generate alarm 47013:0507, Computer Alarm Rod Deviation/Sequencing if the associated bank is actually below 216 steps.

Page 60 of 25

59. IF ERCS is unavailable, core exit thermocouple temperatures can be monitored from
a. any remote multiplexing unit using portable thermocouple bridge readers.
b. the hot shutdown panels in the auxiliary feed pump room.
c. the local display panel in the Relay Room.
d. the ICCM panel behind the main control board.

Page 61 of 25

60. Where are the containment pressure sensors located and how are they affected by changes in atmospheric pressure?

LOCATION EFFECT OF RISING ATMOSPHERIC PRESSURE

a. Inside containment containment pressure indication decreases
b. Inside containment containment pressure indication increases
c. Outside containment containment pressure indication decreases
d. Outside containment containment pressure indication increases

Page 62 of 25

61. With Unit 2 at 98% power near the end of a fuel cycle, what is the initial plant response to a Condenser Steam Dump valve failing open?
a. An increase in steam flow resulting in an increase in turbine load.
b. A decrease in Tavg resulting in CBD rods stepping IN.
c. A decrease in reactor power and an increase in SG levels.
d. An increase in reactor power and a decrease in PRZR level.

Page 63 of 25

62. C26, Air Removal System has a requirement to open SV-33341, AIR EJECTOR LOOP SEAL DRAIN, if condenser air leakage flow is greater than 8.5 scfm. Why is this action necessary?
a. This will remove excess moisture carried over from the condenser due to the high flow.
b. To provide adequate drainage for steam condensation in the discharge line.
c. To prevent loop seal failure, at high flow, which would make the air removal system ineffective.
d. This changes the d/p in the discharge line which acts to rescale the air leakage flow meter.

Page 64 of 25

63. During a liquid radwaste discharge, alarm C47022:0101, HI RADIATION TRAIN B PANEL ALARM, comes in due to a high reading on R-18 but the auto actions do not occur.

What initial operator actions are required?

a. Close CV-31256,Waste Liquid Common Discharge Header Valve and CV-31841, Waste Liquid Common Discharge Header Keylock Release Valve.
b. Close CV-31256,Waste Liquid Common Discharge Header Valve and switch the affected tank to recirculation.
c. Close CV-31841, Waste Liquid Common Discharge Header Keylock Release Valve and switch the affected tank to recirculation.
d. Place the affected tank on recirculation and resample. Flush the discharge line and the radiation monitor.

Page 65 of 25

64. 121 Instrument Air Compressor is going to be isolated for maintenance.

In order to align Station Air to supply Instrument Air upstream of the air dryers, the operator must...

a. Open cross connect valves SA-12-19 and SA-12-18 and verify dryer bypass valve MV-32363 is in automatic.
b. Open manual cross connect valve CP-40-7 and verify one station air compressor is in MANUAL, the other in STANDBY.
c. Open MV-32318, Service Air Header Isolation Valve, and verify station air pressure is greater than instrument air pressure.
d. Open MV-32321, Header Cross Connect, and verify Instrument Air pressure is greater than 85 psig.

Page 66 of 25

65. The cooling water system
a. can pressurize the fire water suppression system through any of 9 crosstie valves.
b. loads can be adequately cooled by the fire protection system.
c. can supply backup fire suppression water to all plant areas EXCEPT inside either containment.
d. and the fire suppression system can only be cross-connected when one unit is in MODE 5.

Page 67 of 25

66. Which event is required to be recorded in the Unit 1 Reactor Log?
a. Receipt of a fuel oil shipment.
b. Placing the #11 Boric Acid Tank on recirc.
c. Addition of oil to the main turbine lube oil reservoir.
d. Operation of the Water Treatment/Reverse Osmosis system.

Page 68 of 25

67. During an independent verification (IV) a valve is found out of position. How should the verifier handle the component out of position situation?
a. Do NOT change valve position. Notify the shift supervisor of the discrepancy.
b. Do NOT change valve position. Notify the initial valve positioner of the discrepancy.
c. Correct the valve position. Have shift supervisor obtain new verifier for that valve only.
d. Correct the valve position. Have the initial valve positioner perform the IV for that valve only.

Page 69 of 25

68. An illuminated status light on the "SI ACTIVE Panel with the Unit at FULL POWER means the associated component
a. is in standby alignment.
b. is NOT in standby alignment.
c. is in its normal alignment.
d. is in its S signal alignment.

Page 70 of 25

69. Reference available: Fig C1A-3, Estimated Critical Boron Concentration Based On Simulate Code.

We are performing a reactor startup on Unit 1 in accordance with C1.2, Unit 1 Startup Procedure and C1B, Appendix - Reactor Startup. Rod withdrawal to obtain initial criticality is in progress. The current conditions are:

  • Zero power rod insertion limit is control bank C @ 47 steps.
  • Control Bank D is at 195 steps
  • Rods have NOT been moved for the last 80 seconds
  • There is a stable SUR of +0.25 dpm What actions are required?
a. Declare the reactor critical and continue with Appendix C1B
b. Declare the reactor critical and borate to lower Control Bank D to <170 steps
c. Fully insert control banks and borate RCS to Mode 3 Hot Shutdown boron concentration.
d. Insert rods to the Estimated Control Rod Position of FIG C1A-3 and dilute to achieve criticality.

Page 71 of 25

70. If the diesel generators automatically start due to a safety injection signal, what diesel generator condition would cause a trip of D1 BUT the same condition would NOT cause a trip of D5?
a. Engine overspeed
b. Generator differential current
c. Ground fault
d. Crankcase high pressure

Page 72 of 25

71. 10CFR20 limits the radiation exposure (dose) to a qualified radiation worker to _______ per year. NMC limits the radiation dose to a qualified radiation worker to _______ per year without special authorization.

10CFR20 limit NMC limit

a. 3000 mrem 1500 mrem
b. 3000 mrem 2000 mrem
c. 5000 mrem 2000 mrem
d. 5000 mrem 3000 mrem

Page 73 of 25

72. The unit is in Mode 1 at 75% power?

What is required if a containment entry must be made?

a. Containment area radiation monitor R2 must read less than 500 mrem/hr.
b. Containment airborne radiation on R11 must be less than 2,000 CPM.
c. Flux mapping can not be performed during the containment entry.
d. Reactor power must be reduced £ 50% rated thermal power.

Page 74 of 25

73. To place the Unit 1 containment in-service purge system in service,
a. the unit must be in MODE 5 or 6.
b. the R11/12 sample selector must be on CONTAIN.
c. containment pressure must be less than 1.3 psig.
d. the supply fan should be started before the isolation dampers are opened.

Page 75 of 25

74. A reactor trip has occurred. During the SS read-through of E-0 Step 3, an Orange Path condition on a Critical Safety Function (CSF) Status Tree occurs.

Transition to the Orange Path procedure should take place:

a. immediately after confirming the Orange Path condition.
b. when transitioning to another E-series procedure.
c. immediately after the SS completes reading step 4.
d. at the discretion of the SS.

ANSWER: B

Page 76 of 25

75. Unit 2 is at 85% power and has experienced an event with the following alarms and indications:
  • 47520-0106, 21 CC PUMP LOCKED OUT
  • 47520-0407, 22 CC PUMP DISCH HDR LO PRESS
  • 47520-0205, 21 CC SURGE TANK HI/LO LVL
  • 47520-0105, 21 CC SURGE TANK LO LO LEVEL
  • 47515-0506, 21 RCP BEARINGS/STATOR HI TEMP
  • 21 RCP radial bearing temperature is 205°F and rising What should be your first action as RO or LEAD?
a. Open MV-32374, REACTOR MAKEUP TO 21 CC SURGE TANK.
b. Trip 21 REACTOR COOLANT PUMP.
c. Manually start 22 CC PUMP.
d. Manually trip the reactor.

Page 77 of 25 Page 78 of 25

76. Unit 1 is operating at 100% power when a main generator lockout occurs.

The following indications are observed during the SS read through of Step 1 of E-0, Reactor Trip or Safety Injection."

  • RTA is open.
  • RTB is closed.
  • Three (3) rod bottom lights are NOT LIT.
  • Power range flux is less than 1% and decreasing.

These indications do NOT change after the manual reactor trip and DSS switches are operated.

What is the next action the Shift Supervisor should take?

a. Remain in E-0 and go to step 2.
b. Transition to FR-S.1, Response to Nuclear Power Generation/ATWS
c. Transition to FR-S.2, Response to Loss of Core Shutdown
d. Direct an operator to locally open RTB.

Page 79 of 25

77. Reference available: 2ES-1.2, Transfer to Recirculation, Page 3 of 12.

Assume all plant equipment was OPERABLE prior to the accident.

A large break loss of coolant accident has occurred on Unit 2. The Lead has just performed step 2 of ES-1.2, Transfer to Recirculation.

The following indications were observed during performance of step 2:

1. The LEFT SI reset pushbutton was depressed. (Train A)
  • The AUTOMATIC SI RESET blue light remained dark. (47014:0504)
  • The SI ACTUATED blue light remained LIT. (47014:0604)
2. The RIGHT SI reset pushbutton was depressed. (Train B)
  • The AUTOMATIC SI RESET blue light illuminated. (47014:0504)
  • The SI ACTUATED blue light remained LIT. (47014:0604)

Step 4 reads Check Both Trains Of Safeguards Pump(s) Available For Recirculation.

What is your assessment of the RHR system based on the indications provided?

a. Both trains of RHR are available for recirculation and ES-1.2 should work without the need to perform RNO actions.
b. Both trains of RHR are available for recirculation BUT the 11 RHR pump must be placed in PULLOUT during switchover.
c. Both trains of RHR are available for recirculation BUT the 12 RHR pump must be placed in PULLOUT during switchover.
d. Both trains of RHR are NOT available for recirculation and a transition should be made to ECA-1.1, Loss of Emergency Coolant Recirculation.

Page 80 of 25

78. IF Unit 2 is in MODE 1 at 100% power, which set of conditions will require notification of the NRC resident per SWI-O-28?
a. The WHITE pressurizer pressure transmitter (PT-430) fails off-scale high with the pressure control selector switch in the NORMAL position (WHITE-BLUE).
b. Seat leakage was verified through BOTH pressurizer PORVs. PORV block valves are now closed with power maintained per the ARP and the high temperature alarm has cleared.
c. A failure of a pressurizer low level bistable has de-energized all of the pressurizer heaters. The heaters will not turn on in auto or manual control.
d. The PRZR PRESS MASTER CONTROL fails with a constant 90% output. The RO has taken manual control of pressurizer heaters and sprays.

Page 81 of 25

79. Unit 2 was operating normally at 100% power with all equipment OPERABLE and operating in the normal preferred lineups when power was lost to Buses 21 and 22.

The following conditions now exist on Unit 2:

  • RCS Tavg is 530°F and lowering
  • 21 SG WR level is 25% and lowering
  • 22 SG WR level is 60% and lowering
  • Both RCPs are stopped
  • Both AFW pumps are running
  • 75 seconds have elapsed since the loss of power Based on the plant indications, the ________(1)________ has NOT tripped and accident analysis limits ________(2)________ be violated.

(1) (2)

a. turbine may
b. turbine will not
c. reactor may
d. reactor will not

Page 82 of 25

80. Given the following conditions:
  • A loss of all AC power has occurred.
  • T-hot is 470°F and stable.
  • RCS pressure is 915 psig.
  • Pressurizer level is 1%.
  • Containment temperature and pressure are 125°F and 2.0 psig.

What form of heat transfer is occurring in the reactor coolant system?

a. Subcooled natural circulation
b. Saturated natural circulation (reflux boiling)
c. Saturated steaming through the break
d. Feed and bleed cooling through the break

Page 83 of 25

81. Reference available: T.S. LCO 3.8.9, Distribution Systems - Operating.

On June 10 at 1000, Unit 1 was in Mode 1 at 98% power.

Sequence of events:

  • 06/10 @ 1000, 480v motor control center (MCC) 1X1 was declared inoperable due to a concern about the seismic mounting of the MCC. Unit 1 entered T.S. LCO 3.8.9.
  • 06/10 @ 1800, A power reduction was started per Action D.
  • 06/10 @ 2355, Unit 1 entered Mode 3.
  • 06/11 @ 0100, Unit 1 started a cooldown to Mode 5.
  • 06/11 @ 0615, Arcing was observed in 120v AC panel 113. The panel was de-energized and a status evaluation was started.
  • 06/11 @ 0730, 480v MCCs 1X1 and 1X2 were declared OPERABLE. Plant cooldown was stopped with Unit 1 at 390°F.

What action(s) is required for this event and when must the action(s) be accomplished?

a. The cooldown should resume immediately. Unit 1 must be in Mode 5 by 06/12/03 @ 0600.

Cooldown may be stopped when Panel 113 is again OPERABLE.

b. The cooldown should resume immediately. Unit 1 must be in Mode 5 by 06/12/03 @ 1400.

Cooldown may be stopped when Panel 113 is again OPERABLE.

c. The cooldown should resume immediately. Unit 1 must be in Mode 5 by 06/12/03 @ 1815.

Cooldown may be stopped when Panel 113 is again OPERABLE.

d. Panel 113 must be returned to an OPERABLE status by 06/11/03 @ 0815 or Unit 1 must start a cooldown and be in Mode 5 by 06/12/03 @ 2015.

Page 84 of 25

82. Following a LOCA, low head recirculation flow was established on both RHR trains. Several hours later, debris blocks the suction lines from Containment Sump B to both trains of RHR and both RHR pumps are cavitating. The RO suggests making a transition to procedure ECA-1.1, Loss of Emergency Coolant Recirculation SHOULD a transition be made to ECA-1.1 and WHY?
a. No, ECA-1.1 strategies would NOT provide a success path for these conditions.
b. Yes, ECA-1.1 would still be effective in restoring long-term cooling.
c. Yes, ECA-1.1 would provide temporary core cooling until the containment is flooded.
d. NO, instead start the SI pumps and return to E-1, Loss of Reactor or Secondary Coolant.

Page 85 of 25

83. Reference available: C12.5 AOP1, Emergency Boration of the Reactor Coolant System.

Due to a turbine load rejection, the control rods are currently about 10 steps below the rod insertion limit. The RO has attempted to establish the required boration flow using the normal boration flowpath but was unsuccessful due to apparent line blockage at the blender. The RO has just completed the actions of C12.5 AOP1, Emergency Boration of the Reactor Coolant System to establish emergency boration flow but has reported that there did NOT appear to be any flow in this flowpath either. The following conditions exist:

  • Steps 1 through 2.4.3 of the AOP were performed with expected main control board indications.
  • There is NO response on 1YIC113, EMERGENCY BORATION INTEGRATOR.
  • RCS temperature is stable.
  • Step 2.4.3 was performed about 1 minute ago.

What action do you take?

a. Direct the RO to monitor alternate indications such as VCT level for evidence of boration.
b. Immediately enter T.S. LCO 3.0.3 and commence an orderly plant shutdown.
c. Borate using C18 AOP1, Makeup or Boration of the RCS Using a Safety Injection Pump.
d. Reduce turbine load to allow manual rod withdrawal until rods are above the rod insertion limit.

Page 86 of 25

84. Refer to ERCS Printouts (Exhibits: SRO-1 through SRO-3).

Unit 1 was operating at 25% power near the end of cycle when RCS temperature dropped for no apparent reason. Rods initially stepped out in AUTO but the operators placed rod control in MANUAL after verifying that a turbine runback was NOT in progress. The transient was not particularly severe and a manual reactor trip was NOT performed. After the transient, the plant parameters were as shown on the referenced printouts.

What is the MOST LIKELY cause of the transient?

a. Xenon oscillation
b. Dropped control rod
c. Power range NI failure
d. RCS boron change

Page 87 of 25

85. What is the basis for T.S. LCO 3.3.3, Event Monitoring Instrumentation requirement to have four (4) core exit thermocouples (CETs) OPERABLE?

The CETs are required to be OPERABLE

a. to help monitor core peaking factors.
b. for primary calorimetric monitoring.
c. to monitor core bypass flow.
d. for subcooling monitoring in the EOPs.

Page 88 of 25

86. Reference available: ES-0.4, Natural Circulation Cooldown With Steam Void In The Vessel.

Unit 1 has tripped and both RCPs are stopped due to mechanical problems. A natural circulation cooldown is in progress but a steam void has formed in the reactor vessel. Circumstances are forcing us to continue the plant cooldown at an accelerated rate using the guidance of ES-0.4, Natural Circulation Cooldown With Steam Void In The Vessel. Step 9 has been completed and you are currently looping through Steps 3 through 9.

The following conditions exist:

  • RCS cooldown is in progress with a stable cooldown rate of 80°F/hour.
  • RCS cold leg temperatures are 480°F and decreasing.
  • RCS hot leg temperatures are 530°F and decreasing.
  • RCS pressure is 1200 psig and stable.
  • Letdown is in service with all orifice isolation valves open.
  • A portion of charging flow is going to pressurizer spray.
  • RCP seal injection flow is 8 gpm to each RCP.
  • Pressurizer level is 88% and stable.
  • RVLIS full range level is 78% and decreasing.

What action should you implement as Unit 1 Shift Supervisor?

a. Turn on available pressurizer heaters and reduce aux spray flow to raise RCS pressure.
b. Reduce steaming rate to drop the RCS cooldown rate to <25°F/hour
c. Reduce charging flow and increase letdown flow to lower pressurizer level.
d. Actuate safety injection and transition to E-0, Reactor Trip or Safety Injection.

Page 89 of 25

87. References available:
  • 2ES-1.1, Post-LOCA Cooldown and Depressurization, step 5.

The following conditions exist:

  • A small break LOCA has occurred on Unit 2.
  • ES-1.1,Post-LOCA Cooldown and Depressurization is being implemented.
  • RCS pressure is 850 psig.
  • RCS subcooling is 52°F.
  • Both RCPs are operating.
  • 21 SG has pressure at 1150 psig and narrow range level at 95% and stable.
  • 22 SG has pressure at 520 psig and narrow range level at 54% and slowly lowering.
  • The Heat Sink Critical Safety Function Status Tree is Yellow due to 21 SG pressure.
  • In ES-1.1, we are about to perform step 5.

The Shift Manager has asked you to make a recommendation on actions to be performed.

You recommend

a. remaining in ES-1.1 but NOT using 21 SG for the RCS cooldown in step 5.
b. transitioning EOP actions to FR-H.2, Response to Steam Generator Overpressure.
c. transitioning EOP actions to FR-H.3, Response to Steam Generator High Level.
d. transitioning EOP actions to E-3, Steam Generator Tube Rupture.

Page 90 of 25

88. References available: T.S. LCO 3.3.1, Reactor Trip System (RTS) Instrumentation and Bases pages 1-15.

For corrective maintenance, power was lowered to 8% on Unit 1. In this condition, I&C had to perform SP-1198, NIS Power Range Startup Test. At the end of the surveillance test on channel N42, the I&C Specialist informed you that the power range low setpoint reactor trip function on channel N42 was found set at 37% power. The I&C Specialist also told you the trip function setpoint could be lowered no further than 25.2% which was NOT within the desired range of 23.9% to 24.9% in the SP. The as left setpoint was recorded as 25.2%.

What was and what is the status of the power range low power reactor trip function?

AS FOUND AS LEFT

a. OPERABLE OPERABLE
b. Inoperable OPERABLE
c. OPERABLE Inoperable
d. Inoperable Inoperable

Page 91 of 25

89. The Engineered Safety Feature Actuation System (ESFAS) automatic logic for AFW actuation is required to be OPERABLE in MODES 1, 2 and 3. However, the auto-start from undervoltage (UV) on the associated 4KV buses is only required to be OPERABLE in MODES 1 and 2. Why?
a. In MODE 3, the thermal power is limited to decay heat only so the UV auto-start is not needed.
b. RCP UV trip is blocked below P-7 and this auto-start exists to promote NC when the RCPs trip.
c. This auto-start anticipates the loss of both MFW pumps which are not required in MODE 3.
d. The UV start is associated with the TDAFW pump which is NOT required in MODE 3.

Page 92 of 25

90. IF Unit 1 loses all AC power, ECA-0.0 will attempt to restore AC power to at least one Safeguard Bus. IF all possible sources are available, what is the PREFERRED BUS and POWER SUPPLY for power restoration?
a. Bus 15 supplied by D1.
b. Bus 15 supplied by 1RY.
c. Bus 16 supplied by D2.
d. Bus 16 supplied by 1RY.

Page 93 of 25

91. References available:

An operator investigating a D6 local alarm reports that the relief valve for 1A Starting Air Receiver has failed open. The control room directs the operator to isolate the 1A air receiver from D6.

What is the status of D6?

D6 is

a. OPERABLE
b. OPERABLE but degraded
c. Inoperable but available (for PRA purposes)
d. Inoperable and not available (for PRA purposes)

Page 94 of 25

92. Pressurizer level is used as a decision point for SI actuation or SI re-initiation in several procedures. In which circumstance should we NOT initiate Safety Injection flow and transition to a different procedure?
a. Mode 1, implementing 1C4 AOP2, Steam Generator Tube Leak, pressurizer level 30% and lowering with maximum charging.
b. Mode 3, implementing 1FR-P.1, Response to Imminent Pressurizer Thermal Shock, SI terminated, pressurizer level 4% and lowering.
c. Mode 3, implementing 1ES-0.1, Reactor Trip Response, pressurizer level is 3% and stable.
d. Mode 4, implementing 1ES-1.1, Post-LOCA Cooldown and Depressurization, SI terminated, pressurizer level 6% and lowering.

Page 95 of 25

93. Following a LOCA, both Hydrogen Recombiners were placed in service. For the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the containment hydrogen concentration continued to rise but at a slower rate Now, we are still implementing ES-1.1, Post-LOCA Cooldown and Depressurization.

Containment hydrogen concentration is 0.9% but the Train B hydrogen recombiner has lost power. The electricians believe there is a cable fault inside the containment such that the recombiner can NOT be returned to service.

What is the expected trend of the hydrogen concentration in containment?

a. Hydrogen concentration will remain below the 4% using only the Train A recombiner.
b. Hydrogen concentration will peak above the 4% but remain below the 6% flammability limit.
c. IF containment spray is NOT placed in service, THEN hydrogen concentration will exceed the 6% flammability limit.
d. Without using POST-LOCA hydrogen venting, hydrogen concentration will exceed the 6%

flammability limit.

Page 96 of 25

94. The Chemist reports that the lithium concentration in the RCS is approaching its upper limit.

What is the preferred action that should be taken to restore the lithium concentration to its proper level?

a. Letdown flow should be maximized to accelerate the cleanup using the CVCS mixed beds.
b. Letdown flow should be minimized to improve the DF of the CVCS mixed beds.
c. Letdown flow should be passed through the CVCS cation bed.
d. Letdown flow should be diverted to the HUT to establish a bleed and feed for the RCS.

Page 97 of 25

95. You are the Containment SRO during fuel handling when a fuel assembly is dropped on top of other fuel assemblies in the core. You note visible damage to the dropped fuel assembly and bubbles rising to the surface. What is your FIRST action?
a. Notify the control room of the conditions.
b. Actuate the containment evacuation alarm.
c. Use the plant page to direct containment evacuation.
d. Direct refueling personnel to don respirators.

Page 98 of 25

96. LCO 3.0.3 states: When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:
a. MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;
b. MODE 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and
c. MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

If we enter LCO 3.0.3 while already in MODE 3, how long do we have to enter MODE 5 from the time we declared our entry into the LCO?

a. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />
b. 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br />
c. 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />
d. 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br />

Page 99 of 25

97. IF the core loading pattern will be CHANGED during the next refueling outage to place the new fuel assemblies more toward the center of the core and the twice-burned assemblies more toward the periphery, THEN what affect would this loading pattern have on the unit?
a. The expected full power loop DT value should be significantly lower for this fuel cycle when compared to the value of full power loop DT for the previous fuel cycle.
b. The expected full power loop DT value should be significantly higher for this fuel cycle when compared to the value of full power loop DT for the previous fuel cycle.
c. IF the pre-startup PR NI channel calibration uses the present fuel cycles initial flux values, the PR NIs would read significantly below actual power when the first calorimetric is performed.
d. IF the pre-startup PR NI channel calibration uses the present fuel cycles initial flux values, the PR NIs would read significantly above actual power when the first calorimetric is performed.

Page 100 of 25

98. Given the following plant conditions:
  • Radiation Monitor 1R-19 has just lost power.

Which of the following actions should be taken?

a. Terminate discharge flow or obtain periodic effluent grab samples.
b. Reset blowdown in the Auxiliary Building and reopen the blowdown control valves.
c. IF R-18 discharge line monitor is operable, discharge may be resumed.
d. Terminate discharge flow because discharge is NEVER allowed with 1R-19 out of service.

Page 101 of 25

99. Given the following conditions on Unit 2:
  • A LOCA outside containment occurred at 0130
  • A Site Area Emergency was declared at 0140
  • The broken line was manually isolated locally, but the operator performing the task was injured and cannot leave the area on his own
  • Initial dose estimates are 90 R/hr gamma
  • The rescue time for a 2-man team is estimated to be 10 minutes with a maximum of 15 minutes Under these circumstances, a rescue attempt
a. is NOT allowed because whole body exposure would exceed the emergency limit.
b. may be made by qualified individuals selected and approved by the Shift Supervisor.
c. may be made by anyone since 10CRF20 exposure limits will NOT be exceeded.
d. by risk-informed volunteers may proceed ONLY with Emergency Director authorization.

Page 102 of 25 100 If there is a fire in the plant, the Shift Manager shall..

a. proceed to the scene and assume the role of Fire Brigade Chief.
b. proceed to the scene to determine if the Red Wing Fire Department should be called.
c. proceed to the control room and assume responsibility for coordinating fire fighting activities.
d. proceed to the technical support center and assume Emergency Director responsibilities.

Page 103 of 75 List of Attachments and References to be provided to the candidates:

Question:

Reference:

1. SRO-6 T.S. LCO 3.8.9, Distribution Systems - Operating.
2. SRO-9 Color copies of ERCS Printouts (Figures SRO-1 through SRO-3).
3. SRO-11 ES-0.4, Natural Circulation Cooldown With Steam Void In The Vessel.

2FR-H.2, Response to Steam Generator Overpressure

4. SRO-12 2FR-H.3, Response to Steam High Level 2ES-1.1, Post-LOCA Cooldown and Depressurization, step 5.
5. SRO-13 T.S. LCO 3.3.1, Reactor Trip System (RTS) Instrumentation.

T.S. LCO 3.8.1, AC Sources - Operating

6. SRO-16 AWI 5.13.5, Operability Determinations.

7.

Control of Exam Copies:

No. Printed for: Status NA Bob Gillespie for Management Review John Kempkes for Peer Review Destroyed on 06/27/03 except for Questions 1, 4, 8 1.

and 22.

2. Gene Dammann for Validation Destroyed on 06/27/03 except for Question 6.
3. Jeff Hawkenson for Validation Destroyed on 06/27/03 no exceptions.
4. John Kempkes for Quality Review 5.

6.

7.

8.

Prepared by: Paul Cooper Reviewed by: John Kempkes Validation by:

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