ML092260186

Initial Exam 2009-301 Final RO Written Exam

ML092260186
Person / Time
Site:	Surry
Issue date:	07/28/2009
From:	NRC/RGN-II
To:
References
50-280/09-301, 50-281/09-301
Download: ML092260186 (95)

Text

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Ap(>licant Information Name:

Date: Facility/Unit:

Region: I I I II I III IV Reactor Type: WI CE **1 BW ~E Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

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

Applicant's Signature Results RO/SRO-OnlylTotal Examination Values -- / -- / -- Points Applicant's Scores - - / - - / - - Points Applicant's Grade -- / -- / -- Percent ES-401, Page 31 of 33

Name: ____________________________ RO Portion of Exam

1. Unit 1 initial conditions:

Reactor power = 100%

=

Condenser vacuum 25" Hg degrading rapidly Current plant conditions:

1-ES-O.1 (REACTOR TRIP RESPONSE) in progress

=

Condenser vacuum 0" Hg Based on the above conditions, which ONE of the following states (1) the temperature at which the RCS would be maintained with no operator action and (2) what temperature the RCS is directed to be maintained by 1-ES-0.1?

A. (1) 550°F (2) 547"F B. (1) 550°F (2) 535°F C. (1) 556°F (2) 535°F D. (1) 556°F (2) 547"F 1

2. Unit 1 plant conditions:

Reactor power = 50% power Reactor Protection System testing in progress

'A' Reactor Trip Breaker is CLOSED

'A' Reactor Trip BYPASS Breaker is CLOSED

'B' Reactor Trip Breaker is CLOSED

'B' Reactor Trip BYPASS Breaker is OPEN Based on the above conditions, if an operator closes (attempts to close) 'B' Reactor Trip BYPASS Breaker which ONE of the following states (1) the status of the Reactor Trip BYPASS Breakers and (2) the status of the Reactor?

A. (1) Both Reactor Trip BYPASS Breakers will open (2) The reactor will trip B. (1) Both Reactor Trip BYPASS Breakers will open (2) The reactor will NOT trip C. (1) Only 'B' Reactor Trip BYPASS Breaker will open (2) The reactor will trip D. (1) Only 'B' Reactor Trip BYPASS Breaker will open (2) The reactor will NOT trip 2

3. A loss of which ONE of the following busses will result in an imminent Safety Injection?

A. Vital Bus '"

B. Vital Bus I C. 'A' DC Bus D. 'B' DC Bus 3

4. Unit 2 Initial Conditions:

• 100% Power.

• Chilled CC is in service to Containment.

Current conditions:

• 2-CD-REF-1, Unit 2 Turbine Building Chiller Unit, trips due to a fault.

Based on the current conditions, which ONE of the following describes (1) the effect on Unit 2 containment indicated partial pressure, AND (2) Unit 2 containment temperature?

A. (1 ) Indicated partial pressure will INCREASE.

(2) Containment temperature will INCREASE.

B. (1 ) Indicated partial pressure will DECREASE.

(2) Containment temperature will INCREASE.

C. (1 ) Indicated partial pressure will INCREASE.

(2) Containment temperature will DECREASE.

D. (1 ) Indicated partial pressure will DECREASE.

(2) Containment temperature will DECREASE.

4

5. Current conditions on Unit 1 are as follows:

• Reactor power is 100%.

• All plant systems and components are in a normal configuration.

• RCS hot and cold leg temperatures are stable.

• Containment pressure is 12.2 psia and rising slowly.

• Pressurizer pressure is 2005 psia and slowly lowering.

• Pressurizer level is 59% and rising.

• Charging flow is lowering in automatic.

Which ONE of the following containment leak locations correlates to the above indications?

A. RCS cold leg B. Main steam line C. Reactor vessel head D. Pressurizer vapor space 5

6. Unit 1 Initial Conditions:

• A small break loss-of-coolant accident (SBLOCA) occurred from 100% power.

• Operators are performing steps in 1-ES-1.2, "POST LOCA COOLDOWN AND DEPRESSURIZATION. "

• A controlled RCS cooldown has been initiated at approximately 90 °Flhr.

• All Steam Generator (S/G) narrow range (NR) levels are approximately 45% and STABLE.

• All S/G pressures are approximately 650 psig and DECREASING.

• Auxiliary Feedwater Flow to EACH steam generator is initially 200 gpm.

CUrrent conditions:

• Operators stopped the CHG pump flowing to the alternate header.

• Operators then "paused" for approximately five (5) minutes after stopping the CHG pump to allow RCS pressure to stabilize or increase before taking further actions to reduce SI flow.

• Following safety injection flow reduction, cooldown rate was calculated to be 77"F/hr.

• No operator actions were performed during the "pause."

Based on the conditions at the end of the "pause" (approximately five (5) minutes after stopping the CHG pump), which ONE of the following predicts:

(1) S/G NR level response AND (2) S/G pressure response?

A. (1 ) levels will be INCREASING.

(2) pressures will be DECREASING.

8. (1 ) levels will be INCREASING.

(2) pressures will be STABLE at a lower value.

C. (1 ) levels will be STABLE at the same value.

(2) pressures will be DECREASING.

D. (1 ) levels will be STABLE at the same value.

(2) pressures will be STABLE at a lower value.

6

7. Unit 1 initial plant conditions:

TIme = 0800

=

Reactor power 100%

Current plant conditions:

Time = 0845 RCS pressure = 700 psig decreasing RCS Sub cooling = 20°F decreasing

=

Safety Injection Flow 225 gpm to each loop Based on the above conditions after transition to 1-E-1 (Loss of Reactor or Secondary Coolant): (1) which ONE of the following actions are directed by 1-E-1 with regard to RCPs and (2) what is the reason for that action?

A. (1) Secure RCPs (2) To reduce the depletion of RCS water inventory.

B. (1) Secure RCPs (2) To prevent the possibility of flywheel fracture if the pump continues to operate without coolant.

C. (1) Maintain RCPs operating (2) They provide core cooling by pumping a 2 phase mixture through the core and loops.

D. (1) Maintain RCPs operating (2) To prevent phase separation in the core region which could lead to core uncovery.

7

8. Current Unit 1 plant conditions:

- Reactor power is at 33%

The latest temperature readings obtained from TR-1-448 are as follows:

TemReratures {OF) Rep 'A' Rep's' Rep 'c' Upper thrust bearing 181 178 163 Lower thrust bearing 173 183 172 Upper radial bearing 143 163 146 Lower radial bearing 172 189 158 Motor stator 285 273 302 Lower bearing seal water 153 183 167 Seal water 195 184 185 Given the above temperature readings, which one of the following correctly states the RCP, if any, that exceeds an ACTION LEVEL limit in Attachment 2, RCP Parameters, of 1-AP-9.00, RCP Abnormal Conditions?

A. RCP 'A'.

B. RCP 'B'.

C. RCP'C'.

D. No RCPs are exceeding an ACTION LEVEL limit.

8

9. Initial plant conditions are as follows:

• Unit 1 is at 100% power.

• Unit 2 is at 100% power.

• 1-CH-P-1 B, Unit 1 B Charging pump, is out of service with motor removed.

• 1-CH-P-1 C, Unit 1 C Charging pump, is running on its alternate power supply.

Current plant conditions are as follows:

• 1-CH-P-1 C has tripped.

• Attempts to start 1-CH-P-1A have been unsuccessful.

• The crew has entered 1-AP-8.00 "Loss of Normal Charging Flow".

• No indications of Unit 1 charging system leakage are observed.

• All Unit 2 Charging pumps are operable.

• RCP seal injection flow is zero.

• Component cooling water flow to the thermal barrier is normal.

Given the above conditions, which ONE of the following would be consistent with the actions required by 1-AP-8.00?

A. Trip Unit 1 reactor. Do NOT trip Unit 2 reactor.

Cross-connect charging with Unit 2 per 1-AP-8.00.

B. Trip Unit 1 reactor. Do NOT trip Unit 2 reactor.

Cross-connecting charging with Unit 2 is NOT permitted per 1-AP-8.00.

C. Trip Unit 1 and Unit 2 reactors.

Cross-connect charging with Unit 2 per 1-AP-8.00.

D. Trip Unit 1 and Unit 2 reactors.

Cross-connecting charging with Unit 2 is NOT permitted per 1-AP-8.00.

9

10. Unit conditions are as follows:

- A unit shutdown is in progress due to increased RCS leakage

- RCS Temperature is 300°F

- RCS Pressure is 300 psig

- The RCS is solid RH-P-1A ("A" RHR Pump) is in service with 1-RH-E-1A ("A" RHR Heat Exchanger)

Which ONE of the following would INITIALLY occur if 1-RH-P-1A were to trip on overcurrent?

A. RCS Pressure would increase B. 1-CH-PCV-1145 (Letdown Pressure Control Valve) would open C. Charging flow would increase D. CC head tank level would increase 10

11. Unit 1 initial conditions:

=

TIme 1000 Reactor power = 100%

1-RC-PCV-14SSB (pzr Spray Valve) fails open RCS pressure = 2100 psig decreasing Current conditions:

TIme = 1001 RCS pressure = 1900 psig decreasing 1-E-O REACTOR TRIP OR SAFETY INJECTION in progress Based on the above conditions: (1) state which ONE of the following actions is directed by step 7 of 1-E-0 if the Pzr Spray Valve can not be closed and (2) state the reason why?

A. (1) Secure RCP A (2) To stop the RCS depressurization.

B. (1) Secure RCP A (2) To prevent inadvertent Safety Injection.

C. (1) Secure RCP C (2) To stop the RCS depressurization.

D. (1) Secure RCP C (2) To prevent inadvertent Safety Injection.

11

12. Unit 1 Initial Conditions:

• An Anticipated Transient Without SCRAM (ATWS) occurred at 100% power.

• The reactor remains at power. An operator is inserting control rods in manual.

• Safety Injection is NOT actuated.

• Charging flow was verified to be 77 GPM, and the BATP was placed in FAST.

Current conditions:

• 1-CH-MOV-1350, Emergency Borate MOV, will not open. An operator reports that the valve appears to be mechanically bound.

• Neither Pressurizer PORV automatically opened when RCS pressure rose above 2335 psig. An operator was able to manually open ONLY one Pressurizer PORV to control RCS pressure.

Based on the current conditions, which ONE of the following correctly identifies (1) the next required action to initiate emergency boration, in accordance with 1-FR-S.1, "RESPONSE TO NUCLEAR POWER GENERATION/ATWS," AND (2) the required action to operate the PORV as specified in FR-S.1?

A. (1) Manually actuate SI to provide for maximum flowrate injection into the RCS.

(2) Allow the RCS pressure to lower to 2210 psig before closing the PORV.

B. (1) Place switches for CH-MOV-1115B and -11150 to OPEN and switches for CH-MOV-1115C and -1115E to CLOSE.

(2) Allow the RCS pressure to lower to 2210 psig before closing the PORV.

C. (1) Manually actuate SI to provide for maximum flowrate injection into the RCS.

(2) Close the PORV when RCS pressure equals 2335 psig and lowering.

O. (1) Place switches for CH-MOV-1115B and -11150 to OPEN and switches for CH-MOV-1115C and -1115E to CLOSE.

(2) Close the PORV when RCS pressure equals 2335 psig and lowering.

12

13. Unit 1 initial plant conditions:

Reactor power = 100%

=

Pzr level 52 % and stable

=

VCT Level 40% and decreasing slowly 1-AP-16.00 (Excessive RCS Leakage) initiated Current Unit 1 conditions:

Leak determined to be Steam Generator Tube Leak in the 1A SG = 39 gpm and increasing slowly The team has initiated 1-AP-24.00 (Minor Steam Generator Tube Leak)

Based on the above conditions: (1) how will charging pump amps change as 1-CH-FCV-1122 (Charging Flow Control Valve) opens to maintain pressurizer level and (2) is a reactor trip required at this instant in time per 1-AP-16.00 or 1-AP-24.00?

A. (1) pump amps will increase (2) Yes B. (1) pump amps will increase (2) No

c. (1) pump amps will decrease (2) Yes D. (1) pump amps will decrease (2) No 13

14. Which ONE of the following completes the below statements?

(1) The parameters used for SI Termination criteria in 1-E-2, "FAULTED STEAM GENERATOR ISOLATION," are RCS subcooling AND _ _ _ _ _ _ __

AND (2) The parameters used for SI Termination criteria in 1-ECA-2.1, "UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS?" are RCS subcooling AND

---

?

A. (1) E-2: ONLY RCS pressure and PRZR level.

(2) ECA-2.1: RVLlS indication greater than values specified in a table (based upon number of running RCPs).

B. (1) E-2: ONLY RCS pressure and PRZR level.

(2) ECA-2.1: RCS pressure and PRZR level.

C. (1) E-2: Secondary heat sink, RCS pressure, and PRZR level.

(2) ECA-2.1: RVLlS indication greater than values specified in a table (based upon number of running RCPs).

D. (1) E-2: Secondary heat sink, RCS pressure, and PRZR level.

(2) ECA-2.1: RCS pressure, and PRZR level.

14

15. Unit 1 initial conditions:

Station Blackout occurs 1-ECA-0.O (LOSS OF ALL AC POWER) in progress SGs are to be depressurized to allow SI accumulators to inject into the RCS.

Based on the above conditions, which ONE of the following: (1) correctly states the maximum cooldown rate allowed during this depressurization and (2) the basis for that rate ?

A. (1) 25 oF/Hr (2) To prevent a steam bubble from forming in the reactor vessel head.

B. (1) 25 oF/Hr (2) To minimize RCS inventory loss.

C. (1) 100 oFlHr (2) To prevent a steam bubble from forming in the reactor vessel head.

D. (1) 100 oF/Hr (2) To minimize RCS inventory loss.

15

16. Initial Conditions:

• Surry Unit 2 is in day 25 of a scheduled refueling outage.

• A Loss of All AC Power occurred on Unit 1, which was operating at 100% power.

• Control room operators implemented 1-ECA-0.0, "LOSS OF ALL AC POWER."

• Safety Injection (SI) initiated on Unit 1.

• Power was restored to one Unit 1 safeguards power train from an Emergency Diesel Generator.

Current conditions:

• Unit 1 control room operators are implementing 1-ECA-0.2, "LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED," and are performing the step that directs restoration ofSWto CC HXs IAWO-AP-12.01, "LOSS OF INTAKE CANAL LEVEL."

• Service water is isolated to all recirculation spray (RS) heat exchangers on both units.

• 2 ESW pumps are running.

Note: In 0-AP-12.01, "time zero" is defined as that time intake canal level reaches 23.5 FT.

Based on the current conditions, which ONE of the following correctly identifies (1) the 0-AP-12.01 restriction, if any, on CC HX SW flow 15 minutes after "time zero" AND (2) the 0-AP-12.01 restriction on CC HX SWflow after nine hours have elapsed from "time zero?"

(Reference provided)

A. (1) There are no restrictions on CC HX SW flow.

(2) Crosstie CC. 3 CC HXs allowed with SW outlet valves 19 turns open for each HX.

B. (1) There are no restrictions on CC HX SW flow.

(2) 2 CC HXs allowed with SW outlet valves 19 turns open for each HX.

C. (1) Maximum allowable flow is one CC HX outlet SW valve fully open.

(2) Crosstie CC. 3 CC HXs allowed with SW outlet valves 19 turns open for each HX.

D. (1) Maximum allowable flow is one CC HX outlet SW valve fully open.

(2) 2 CC HXs allowed with SW outlet valves 19 turns open for each HX.

16

17. Unit 2 is performing a plant shutdown with the reactor at 5% power when a loss of Vital Bus 1 occurs.

Which one of the following correctly describes:

(1) the direct effect of the loss of Vital Bus I on the RPS system and (2) if a trip or shutdown were to occur, the effect on the re-instatement of the Source Range NI's?

A. (1) An RPS trip signal is generated (2) Re-instatement of SRNls will occur automatically.

B. (1) An RPS trip signal is generated (2) Re-instatement of SRNls will NOT occur automatically.

c. (1) An RPS trip signal is NOT generated (2) Re-instatement of SRNls will occur automatically.

D. (1) An RPS trip signal is NOT generated (2) Re-instatement of SRNls will NOT occur automatically.

17

18. Unit 1 Initial Conditions:

• 100% power.

Current conditions:

• A complete loss of DC bus 1B has occurred.

Based on the current conditions, which ONE of the following correctly identifies (1) the required action in 1-AP-10.06, "LOSS OF DC POWER," for the generator output breakers, AND (2) the subsequent impact on Reactor Coolant Pump (RCP) operations following completion of the immediate actions of 1-E-0 (Reactor Trip or Safety Injection)?

A. (1) A control room operator is NOT required to manually open the Generator output breakers.

(2) 'A' RCP will stop, 'B' and 'c' RCPs will remain running.

B. (1) A control room operator is NOT required to manually open the Generator output breakers.

(2) 'A' RCP will remain running, 'B' and 'C' RCPs will stop.

C. (1) A control room operator is required to manually open the Generator output breakers.

(2) 'A' RCP will stop, 'B' and 'c' RCPs will remain running.

D. (1) A control room operator is required to manually open the Generator output breakers.

(2) 'A' RCP will remain running, 'B' and 'c' RCPs will stop.

18

19. Unit 1 initial conditions; Reactor power = 90% and decreasing Shutdown in progress for refueling Current plant conditions:

Reactor power = 70% and decreasing 1G-85 COMPUTER PRINTOUT ROD CONT SYS in alarm QPTR = 1.05% and increasing The power decrease is stopped (1) Based on the above indications, which ONE of the following conditions has occurred and (2) what is the minimum QPTR that requires operator action when exceeded lAW Tech Spec 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS?

A. (1) Stuck control rod (2) 2%

8. (1) Stuck control rod (2) 5%

C. (1) Control bank inserted past insertion limit (2) 2%

D. (1) Control bank inserted past insertion limit (2) 5%

19

20. Following a complete loss of instrument air and subsequent restoration, which ONE of the following components will require manual re-alignment to support the start of 1-RC-P-1C?

A. 1-CC-TV-105C, RCP C Cooler CC Return Trip Valve B. 1-CC-TV-120C, RCP C Thermal Barrier Component Cooling Water System Trip Valve C. 1-RC-HCV-1303C, RCP C Seal Leakoff Isolation Valve D. 1-CH-MOV-1381, RCP Seal Return Valve 20

21. Unit 1 Initial Conditions:

• 100% Power

• Both Megawatts and Megavars start to oscillate.

• System Operator reports that the regional grid is experiencing dynamic instabilities.

Current conditions:

• 230 KV system voltage is 212 KV

• 500 KV system voltage is 492 KV Based on the current conditions, which one of the following identifies (1) the required operator action for continued voltage regulator operation, as specified in O-AP-1 0.18, "RESPONSE TO GRID INSTABILITY," AND (2) the required entry into Technical Specification (TS) 3.16, "Emergency Power System?"

A. (1) Place the voltage regulator - in MANUAL.

(2) Entry into TS 3.16 is required ONLY for the 230 KV system. Entry into TS 3.16 for the 500 KV system is NOT required.

B. (1) Place the voltage regulator - in MANUAL.

(2) Entry into TS 3.16 is required for BOTH the 230 KV system AND the 500 KV system.

C. (1) Verify the voltage regulator - in AUTO, or place the voltage regulator in AUTO if possible.

(2) Entry into TS 3.16 is required for BOTH the 230 KV system AND the 500 KV system.

D. (1) Verify the voltage regulator - in AUTO, or place the voltage regulator in AUTO if possible.

(2) Entry into TS 3.16 is required ONLY for the 230 KV system. Entry into TS 3.16 for the 500 KV system is NOT required.

21

22. Unit 1 Initial Conditions:

o A loss-of-coolant accident (LOCA) occurred from 100% power.

o Operators are implementing 1-ECA-1.2, "LOCA OUTSIDE CONTAINMENT."

Current conditions:

o Operators have completed all the steps in ECA-1.2 that (a) attempt to verify proper valve alignment, and (b) locate and isolate the leak.

oRCS pressure continues to DECREASE.

o Annunciator 1B-F3, SFGS AREA SUMP HI LEVEL, is LIT.

o 1-VG-RM-110, VENT VENT 2 GAS, is below the HIGH setpoint, but is rapidly trending UP.

o RM-GW-130-1, PROCESS VENT STK PART, is below the HIGH setpoint, but is rapidly trending UP.

Based on the current conditions, which ONE of the following:

(1) is the NEXT overall mitigating strategy that should be implemented, as specified by ECA-1.2, AND (2) a filtered release to the environment _ _ _ _ _ _ _ _ exist?

A. (1) Depressurize S/Gs to inject accumulators.

(2) DOES B. (1) Conserve RWST inventory.

(2) DOES C. (1) Depressurize S/Gs to inject accumulators.

(2) DOES NOT D. (1) Conserve RWST inventory.

(2) DOES NOT 22

23. Unit 1 initial conditions:

Loss of Main and Auxiliary Feedwater EOP transition from E-O to 1-FR-H.1(RESPONSE TO LOSS OF SECONDARY HEAT SINK)

Current plant conditions:

The 1A SG is to be depressurized to establish condensate flow to the SG.

RCS Pressure is 2000 psig and slowly increasing Steam Generator Levels are as follows:

'A' SG Wide Range Level - 42% and slowly decreasing

'B' SG Wide Range Level- 41% and slowly decreasing

'C' SG Wide Range Level - 43% and slowly decreasing Based on the above conditions: (1) why is steam flow limited to 1.0 x 106 Lbm/hr during the depressurization and (2) what actions are directed by 1-FR-H.1 if adequate condensate flow to the 1A SG does not occur with SG pressure = 550 psig?

A. (1) To prevent main steam line isolation (2) Continue to depressurize 'A' SG until condensate flow established B. (1) To prevent main steam line isolation (2) Commence RCS Bleed and Feed C. (1) To limit cooldown rate to < 100°FlHr (2) Continue to depressurize 'A' SG until condensate flow established D. (1) To limit cooldown rate to < 100°F/Hr (2) Commence RCS Bleed and Feed 23

24. Initial Unit 1 Conditions:

- Power Range Nls are: N-41=36.3%, N-42=36.2%, N-43=34.7%, N-44=36.6%

- Delta T Power is 35%

- All three RCPs are operating Present Unit 1 Conditions:

- RCP 1A frame vibrations indicate 10 mils

- 2 minutes after the frame vibrations increase to 10 mils, the speed sensing panel actuates for RCP 1A.

Based on the above conditions, which one of the following:

(1) correctly states the status of the 1C-H4, RCP FRAME DANGER, annunciator and (2) the impact on reactor operations from the actuation of the speed sensing panel?

A. (1) 1C-H4 is illuminated.

(2) Unit 1 reactor automatically trips.

B. (1) 1C-H4 is illuminated.

(2) Unit 1 reactor does NOT automatically trip.

C. (1) 1C-H4 is NOT illuminated.

(2) Unit 1 reactor automatically trips.

D. (1) 1C-H4 is NOT illuminated.

(2) Unit 1 reactor does NOT automatically trip.

24

25. A tube leak has developed in the thermal barrier heat exchanger for 1-RC-P-1A.

Once the indicated thermal barrier flow has exceeded 50 gpm, which ONE of the following (1) states the expected automatic plant response and (2) the associated time delay for that response?

A. (1) 1-CC-TV-120A (A RCP Thermal Barrier Isolation Trip Valve) will close (2) 50 seconds.

B. (1) 1-CC-TV-120A (A RCP Thermal Barrier Isolation Trip Valve) will close (2) 10 seconds.

C. (1) 1-CC-TV-140A (Thermal Barrier CC Return Trip Valve) and 1-CC-TV-140B (Thermal Barrier CC Return Trip Valve) will close (2) 50 seconds.

D. (1) 1-CC-TV-140A (Thermal Barrier CC Return Trip Valve) and 1-CC-TV-140B (Thermal Barrier CC Return Trip Valve) will close (2) 10 seconds.

25

26. Current plant conditions are as follows on Unit 2:

• Reactor power is 100%

• Pressurizer level is 50% and slowly lowering.

• Letdown flow is 104 gpm The team initiated 1-AP-16.00 (Excessive RCS Leakage) and completed the immediate actions of 1-AP-16.00. At this time, the reactor operator reports the following:

• Seal Injection Flow to each RCP is 7 gpm

• Seal Leak-off Flow for each RCP is 3 gpm

• Tave is stable at 573 of

• The leak rate was determined to be 75 gpm Which ONE of the following correctly describes what value the RO should set charging flow at to maintain pressurizer level stable?

A. 54 GPM B. 63 GPM C. 75 GPM D. 84 GPM 26

27. Unit 1 Conditions:

Cold Shutdown - 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after an extended full power run 1-RC-LJ-100A = 12.0 ft.

1A RHR pump operating and pump amps are oscillating 1-RH-FI-1605 = 2000 gpm and oscillating The team has increased RCS level back into the acceptable region and secured 1-RH-P-1A in accordance with 1-AP-27.00 (LOSS OF DECAY HEAT REMOVAL CAPABILITY).

Which ONE of the following actions are directed by 1-AP-27.00 to restore RHR flow?

A. Vent the 1A RHR pump, restart the 1A RHR pump and verify the RHR heat sink.

B. Vent the 1B RHR pump, start the 1B RHR pump and verify the RHR heat sink.

C. Close 1-RH-HCV-1758 and 1-RH-FCV-1605, then re-start the 1A RHR pump and throttle to the pre-event rate.

D. Close 1-RH-HCV-1758 and 1-RH-FCV-1605, then start the 1B RHR pump and throttle to the pre-event rate.

27

28. Plant conditions:

=

Unit 1 100%

=

Unit 2 shutdown for refueling MAIN CONTROL ROOM OXYGEN MONITOR alarms A hazardous substance has been spilled in the main control room O-AP-20.00, MAIN CONTROL ROOM INACCESSIBILITY, is initiated Operators proceed to the Auxiliary Shutdown Panel with FCA procedures Based on the above conditions, which ONE of the following states equipment that is operated from the Auxiliary Shutdown Panel as directed by O-AP-20.00?

A. CC pumps B. RHR pumps C. 1-CH-HCV-1311 (Aux Pressurizer Spray)

D. 1-CH-FCV-1122 (Charging Flow Controller) 28

29. Initial plant conditions on Unit 1 are as follows:

• Reactor tripped following an inadvertent SI.

Current plant conditions on Unit 1 are <IS follows:

• 1-ES-1.1, SI Termination, has been initiated.

• SI signal has been reset.

• Actions have been completed to re-establish charging and letdown.

• Sealleakoff flow from each RCP is -3 gpm.

• PRT level shows a slow rising trend.

• VCT level is 43% and decreasing.

• Reactor Pressure is stable at 2235 psig.

• Tailpipe temperatures are 105°F and stable.

Which ONE of the following would be consistent with the above conditions?

A. 1-CH-MOV-1381 "Seal Return Header Isolation Valve" is closed causing 1-CH-RV-1382B "Seal Return Heat Exchanger Relief Valve" to lift.

B. A component cooling water leak has developed on the Seal Return Heat Exchanger causing 1-CH-RV-1382B "Seal Return Heat Exchanger Relief Valve" to lift.

C. 1-CH-MOV-1381 "Seal Return Header Isolation Valve" is closed causing 1-CH-RV-1382A "Seal Return Relief Valve" to lift.

D. A component cooling water leak has developed on the Seal Return Heat Exchanger causing 1-CH-RV-1382A "Seal Retum Relief Valve" to lift.

29

30. Unit 1 Initial Conditions:

• 100% Power

• Pressurizer Relief Tank (PRT) level is 62% and STABLE

• A leaking pressurizer code safety valve has been identified.

Current conditions:

• A PRT annunciator has just alarmed

• PRT level is 76% and INCREASING

• PRTtemperature is 126 of and INCREASING

• PRT pressure is 3 psig and INCREASING Based on the current conditions, which one of the following identifies (1) the parameter that caused the PRT annunciator, AND (2) chemistry will sample the PRT gas space when PRT level INCREASES by 5% because large changes in PRT level can result in an undesirable atmosphere in the PRT gas space (2)?

A. (1) High level (2) due to PG O 2 off-gassing as PRT temperature changes.

B. (1) High temperature (2) due to PG O 2 off-gassing as PRT temperature changes.

C. (1) High level (2) due to H2 concentration potentially exceeding 4% in the process vent.

D. (1) High temperature (2) due to H2 concentration potentially exceeding 4% in the process vent.

30

31. Unit 1 plant conditions:

=

Reactor power 100%

Charging flow = 100 gpm increasing 1-CC-RI-105 (CC Heat Exchanger AlB Outlet Radiation Monitor) alarms HIGH CC surge tank level = 64% increasing 1-AP-16.00 (EXCESSIVE RCS LEAKAGE) is initiated Based on the above conditions, which ONE of the following describes where the excess volume in the CC system will go to and (2) what actions is directed first by 1-AP-16.00 to attempt to isolate the leak?

A. (1) The process vent system (2) Isolate letdown B. (1) The process vent system (2) Isolate thermal barrier on suspected RCP C. (1) The auxiliary building sump (2) Isolate letdown D. (1) The auxiliary building sump (2) Isolate thermal barrier on suspected RCP 31

32. Unit 1 Initial Conditions:

• 75% Power at Middle-of-Life (MOL) conditions.

• Rod Control is in AUTOMATIC.

• VCT automatic makeup controls are set to the current RCS boron concentration.

• Excess Letdown is in service in preparation for removing Normal Letdown from service.

Current conditions:

• Component Cooling (CC) surge tank level is slowly DECREASING at 1% every 5 minutes.

• Reactor power is slowly INCREASING.

Based on the current conditions and assuming NO other operator actions, which ONE of the following identifies (1) the location of the CC leak that would cause the current conditions, AND (2) the expected impact of the CC leak on rod control?

LEAKING COMPONENT ROD CONTROL A. (1) Seal Water Heat Exchanger (2) Rods will step IN.

B. (1) Seal Water Heat Exchanger (2) Rods will step OUT.

C. (1) Excess Letdown Heat Exchanger (2) Rods will step IN.

D. (1) Excess Letdown Heat Exchanger (2) Rods will step OUT.

32

33. In FR-C.2 (Response to Degraded Core Cooling), if attempts to establish adequate core cooling using the High Head Safety Injection System are ineffective, the intact Steam Generators are depressurized to 200 psig and subsequently to atmospheric pressure.

Which ONE of the following describes (1) the purpose of the depressurization of the steam generators and (2) why the steam generators are depressurization is stopped at 200 psig?

A. (1) To depressurize the RCS to increase accumulator and Low Head Safety Injection Flow.

(2) To isolate the Safety Injection Accumulators and prevent Nitrogen addition into the RCS.

B. (1) To depressurize the RCS to increase accumulator and Low Head Safety Injection Flow.

(2) To prevent the loss of RCP support conditions by maintaining adequate RCS pressure.

C. (1) To maximize Auxiliary Feedwater Flow and enhance RCS cooldown.

(2) To isolate the Safety Injection Accumulators and prevent Nitrogen addition into the RCS.

D. (1) To maximize Auxiliary Feedwater Flow and enhance RCS cooldown.

(2) To prevent the loss of RCP support conditions by maintaining adequate RCS pressure.

33

34. Concerning a pressurized thermal shock event, a _(1 )_ _ would cause the most significant Pressurized Thermal Shock (PTS) challenge to the plant, and in accordance with FR-P.1 (Response to Imminent Pressurized Thermal Shock), an appropriate response would be to _ _(2)_ _?

A. (1) Main Steam Line Break (2) Stop the RCS cooldown and reduce Safety Injection flow B. (1) Main Steam Line Break (2) Reduce the ReS cooldown rate to < 100 °Flhr and depressurize the RCS to establish 30 of subcooling C. (1) Steam Generator Tube Rupture (2) Stop the RCS cooldown and reduce Safety Injection flow D. (1) Steam Generator Tube Rupture (2) Reduce the RCS cooldown rate to < 100 °Flhr and depressurize the RCS to establish 30 OF subcooling 34

35. A Large Break LOCA has occurred on Unit 1.

Current conditions:

• All plant systems operated as designed.

• Steam Generator Levels are 14% Narrow Range and stable in all three Steam Generators.

• AFW flow to each Steam Generator is 125 gpm.

• Containment Pressure is currently 15 psia after peaking at 47 psia.

• Containment High Range Radiation Monitors (1-RM-RI-127 and 128) are currently indicating 1 x 104 after peaking at 4.3 x 106 .

Based on the current conditions, which ONE of the states (1) the status of Heat Sink and (2) why?

A. (1) Steam generator Heat Sink requirements are met.

(2) Adequate Steam generator Inventory.

B. (1) Steam generator Heat Sink requirements are met.

(2) Adequate AFW flow.

C. (1) Steam generator Heat Sink requirements are NOT met.

(2) Inadequate Steam generator Inventory and AFW flow is < the limit of 540 gpm D. (1) Steam generator Heat Sink requirements are NOT met.

(2) Inadequate Steam generator Inventory and AFW flow is < the limit of 450 gpm.

35

36. Unit 1 plant conditions:

Plant runback occurs from 100% to 90%

RCS HI PRESSURE ALARM lit Master Pressure Controller output demand is currently 70%

Based on the above conditions, which ONE of the following states (1) the status of the Pressurizer Spray valves and (2) what the maximum Pressurizer spray flow rate is based on?

A. (1) Full Open (2) To prevent the PORVs from opening during a 10% step load decrease.

B. (1) Full Open (2) To prevent exceeding the capacity of the PORVs during a load rejection from 100% power.

C. (1) Modulated Open << 100%)

(2) To prevent exceeding the capacity of the PORVs during a load rejection from 100% power.

D. (1) Modulated Open << 100%)

(2) To prevent the PORVs from opening during a 10% step load decrease.

36

37. Unit 1 initial conditions:

- Reactor power = 5%

- Pressurizer Pressure Protection transmitter (1-RC-PT-1456) failed low (I&C investigating)

Current plant conditions:

- Pressurizer Pressure Protection transmitter (1-RC-PT-1455) subsequently failed low Based on the current plant conditions, which ONE of the following correctly describes (1) the effect on the reactor and (2) the effect these failures on subcooling indication on the Inadequate Core Cooling Monitor (ICCM)?

A. (1) A Reactor Trip will occur (2) No impact on subcooling indication B. (1) A Reactor Trip will NOT occur (2) Subcooling will indicate -35 of C. (1) A Reactor Trip will occur (2) Subcooling will indicate -35 of D. (1) A Reactor Trip will NOT occur (2) No impact on subcooling indication 37

38. Unit 1 plant conditions:

=

Reactor power 100%

Current conditions:

LBLOCA occurs Containment pressure = 25 pSia increasing 1-CS-MOV-101A (Containment Spray Pump 'A' Discharge Valve) does not open 1-CS-MOV-1 01 B (Containment Spray Pump 'A' Discharge Valve) does not open Based on the above conditions, which ONE of the following states:

(1) Which Recirculation Spray (RS) System pump suction(s) is/are being supplied by B Train of Containment Spray?

(2) If sufficient containment spray flow is being supplied to meet the design basis of the CS system?

A. (1) RS Train B ONLY (2) The CS design basis is being met.

B. (1) RS Train B ONLY (2) The CS design basis is NOT being met.

C. (1) RS Train A and B (2) The CS design basis is being met.

D. (1) RS Train A and B (2) The CS design basis is NOT being met.

38

39. Unit 1 Initial Conditions:

• Unit is shutdown for refueling operations.

• 1-RM-RM-159/160, Containment Particulate/Gas, both read 875 cps.

• 1-RM-RM-152, New Fuel Storage Area, reads 1.7 mr/hr.

Current conditions:

• Core off-load is in progress, when an event occurs.

• 1-RM-RM-159/160, Containment Particulate/Gas, both read 880 cps.

• 1-RM-RM-152, New Fuel Storage Area, reads 15.3 mr/hr.

Based on the current conditions, which ONE of the following describes REQUIRED operator actions, in accordance with 0-AP-22.00, "FUEL HANDLING ABNORMAL CONDITIONS?"

(1) Fuel handling operations MUST STOP _ _ _ _ _ _ _ __

AND (2) after dumping a train of MCR air bottles in accordance with 0-AP-22.00, THEN

---

?

A. (1) in the Fuel Building. Fuel Handling operations may continue in Containment.

(2) IMMEDIATELY start one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred)

B. (1) in the Fuel Building. Fuel Handling operations may continue in Containment.

(2) Wait 50 minutes before starting one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred)

C. (1) in BOTH the Fuel Building AND Containment (2) Wait 50 minutes before starting one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred)

D. (1) in BOTH the Fuel Building AND Containment (2) IMMEDIATELY start one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred) 39

40. Unit 1 is at 20% power during a power increase following a maintenance shutdown.

Initial conditions:

Time = 1000 An existing 2 gallon per day tube leak exists on the 1A SG CHG LINE FLOW = 97 gpm and increasing 1-AP-16.00 (EXCESSIVE RCS LEAKAGE) is entered Current conditions:

=

Time 1015 Main Steam Line Rad Monitor level increasing (NOT in alarm)

The NEW Steam Generator Tube leak rate is determined to be 6 gpm Based on the above conditions, which ONE of the following: (1) correctly states if 1-AP-24.00 (MINOR SG TUBE LEAK) is required to be initiated lAW 1-AP-16.00 and (2) what procedure is required to shut down the unit?

A. (1) Yes (2) 1-AP-23.00 (RAPID LOAD REDUCTION).

B. (1)Yes (2) 1-GOP-2.2 (UNIT SHUTDOWN, LESS THAN 30% TO HSD).

C. (1) No (2) 1-AP-23.00 (RAPID LOAD REDUCTION).

D. (1) No (2) 1-GOP-2.2 (UNIT SHUTDOWN, LESS THAN 30% TO HSD).

40

41. Unit 1 Initial Conditions:

• A Steam Generator (S/G) tube rupture caused an automatic reactor trip and SI from 100% power.

• Control room operators are implementing 1-ECA-3.1, "SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY."

Current conditions:

• A maximum-rate cooldown in accordance with ECA-3.1 was commenced at time 1500.

• The following data has been logged over the last hour: (consider that the time is currently 1600)

TIME RCS COLD LEG TEMP 1500 395 OF 1515 370 of 1530 346 of 1545 321°F 1600 296 of Based on the current conditions, which ONE of the following correctly describes the cooldown from 1500 to 16007 (1) The Technical Specification cooldown rate limit _ _ _ _ _ _ _ _ __

AND (2) The cooldown is _ _ _ _ _ _ _ _ _ __

A. WAS exceeded.

required to be temporarily stopped.

B. WAS exceeded.

NOT required to be stopped.

C. was NOT exceeded.

NOT required to be stopped.

D. was NOT exceeded.

required to be temporarily stopped.

41

42. Unit 1 Initial Conditions:

• A reactor startup is in progress.

• All Steam Generator (S/G) Power Operated Relief Valves (PORV) are 10%

open, and the controllers are being operated in automatic.

Current conditions:

• The 1A S/G PORV controller setpoint is at 1000 psig AND the setpoint is continuously DECREASING.

Based on the current conditions, which ONE of the following correctly describes the effect of this failure if no operator action is taken?

A. If uncorrected, the 522 of minimum temperature for criticality Tech Spec limit may be violated.

B. If uncorrected, the 545 of minimum temperature for criticality Tech Spec limit may be violated.

C. 1Band 1C S/G PORVs will open to relieve more steam and maintain Tave within acceptable limits.

D. 1B and 1C S/G PORVs will not respond, and Tave will increase above the limit of 577 of.

42

43. Unit 2 plant conditions:

2-GOP 1.5 (UNIT STARTUP, 2% REACTOR POWER TO MAX ALLOWABLE POWER) is in progress Reactor Power = 12%

2-FW-P-1A CA' Main Feed Water Pump) is feeding all SGs at 1500 gpm each Based on the above plant conditions, which ONE of the following conditions will cause 2-FW-P-1A to trip?

A. MFW pump suction header pressure 65 psig for> 15 sec B. Bus voltage dips to 65% and returns to normal C. 'A' Main Feed Pump Recirc Valve closed for> 15 sec D. 2-FW-P-1A lube oil pressure decreases to 5 psig 43

44. Unit 1 plant conditions:

Reactor power = 50%

1-GW-RM-130A (Process Vent Particulate Radiation Monitor) high alarm sounds Based on the above conditions: (1) which ONE of the following valves are interlocked to reposition upon receiving the high alarm and (2) why does that action occur?

A. (1) GW-FCV-101 (Waste Gas Decay Tank bleed FCV)

(2) To isolate a potential release path B. (1) GW-FCV-101 (Waste Gas Decay Tank bleed FCV)

(2) To redirect flow through the waste gas charcoal filters C. (1) GW-FCV-100 (Process Vent Flow Control Valve)

(2) To isolate all potential release paths D. (1) GW-FCV-100 (Process Vent Flow Control Valve)

(2) To redirect flow through the waste gas charcoal filters 44

45. Unit 1 Initial Conditions:

• The plant operated continuously at 100% power for a period of time before the team manually tripped the reactor due to turbine high vibrations.

• All plant systems and component operated as designed.

Current conditions:

• Offsite power is NORMAL.

• Both motor-driven Auxiliary Feedwater (AFW) pumps 1-FW-P-3A and 1-FW-P-3B are running.

• All Steam Generator (S/G) narrow range levels are LESS THAN 12%.

• Control room operators have transitioned to 1-ES-O.1, "REACTOR TRIP RESPONSE."

Based on the current conditions, which ONE of the following:

(1) identifies the core bumup at time of the trip that will result in the GREATER required AFW system flowrate to maintain S/G levels stable, AND (2) is the MINIMUM AFW flowrate required by ES-0.1 for the current plant conditions?

CORE BURNUP MINIMUM REQUIRED AFW FLOWRATE A. (1) 1,000 MWO/MTU (2) 350 GPM B. (1) 1,000 MWO/MTU (2) 540 GPM C. (1 ) 10,000 MWOIMTU (2) 350 GPM D. (1 ) 10,000 MWO/MTU (2) 540 GPM 45

46. Unit 2 Initial Conditions:

• 100% Power.

• The MCR Undervoltage (UV) bypass switches for turbine-driven Auxiliary Feedwater (AFW) pump 2-FW-P-2 steam supply valves 2-MS-PCV-202A and 2-MS-PCV-202B are in "BYPASS" position for routine maintenance.

• At the completion of maintenance, the UV bypass switches are retumed to the "NORMAL" position, as required.

• Due to an electrical failure, all signals will respond as if the UV bypass switch was still in the "BYPASS" position. Operators are unaware of this condition.

Current conditions:

• A loss of all offsite power has occurred.

• #2 EDG loaded on 2H bus and #3 EDG loaded on 2J bus.

• Safety Injection (SI) is NOT actuated.

Based on the current conditions, which ONE of the following correctly identifies the expected plant response?

A. (1) 2-MS-PCV-202A and -202B will receive automatic open signals.

(2) AFW discharge valves 2-FW-MOV-251A through -251F will receive automatic open signals.

B. (1) 2-MS-PCV-202A and -202B will receive automatic open signals.

(2) AFW discharge valves 2-FW-MOV-251A through -251F will NOT receive automatic open signals.

C. (1) 2-MS-PCV-202A and -202B will NOT receive automatic open signals.

(2) AFW discharge valves 2-FW-MOV-251A through -251 F will receive automatic open signals.

D. (1) 2-MS-PCV-202A and -202B will NOT receive automatic open signals.

(2) AFW discharge valves 2-FW-MOV-251A through -251 F will NOT receive automatic open signals.

46

47. Unit 1 Initial Conditions:

• 100% Power.

Current conditions:

• Loss of letdown.

• Steam dump control NOT affected.

• Loss of Component Cooling to ALL RCP thermal barrier heat exchangers.

• Component Cooling to ALL other RCP heat exchangers is NOT affected.

Based on the current conditions, which ONE of the following correctly identifies the vital AC bus or buses that haslhave been de-energized?

A. Vital Bus I is de-energized. Vital Buses II, III, and IV are energized.

B. Vital Bus I and Vital Bus III are de-energized. Vital Buses II and IV are energized.

C. Vital Bus II is de-energized. Vital Buses I, III, and IV are energized.

D. Vital Bus II and Vital Bus IV are de-energized. Vital Buses I and III are energized.

47

48. Unit 1 plant conditions:

Reactor power = 100%

UPS 1A 1 Battery charger fails Based on the above conditions, which ONE of the following actions will automatically occur to power loads on DC bus 1A?

A. UPS 1B1 will power DC bus 1A B. Battery 1A will pickup loads for the next two hours C. DC bus 1A will cross connect to DC bus 1B D. UPS 1A2 will power DC bus 1A 48

49. A loss of 'A' DC Bus occurs followed by a Safety Injection. Which ONE of the following is correct regarding the operation of 1-SI-P-1A CA' Low Head Safety Injection Pump)?

A. 1-SI-P-1A is NOT running but can be started from the MCR.

B. 1-SI-P-1A is NOT running and can NOT be started from the MCR.

C. 1-SI-P-1A is running and can be stopped from the MCR.

D. 1-SI-P-1A is running but can NOT be stopped from the MCR.

49

50. Unit 1 Initial Conditions:

• A spurious safety injection from 100% power occurred four (4) minutes ago.

Current conditions:

• An electrical grid transient has JUST resulted in a Station Blackout.

Based on the current conditions, which ONE of the following correctly identifies the SEQUENCE that ALL equipment will automatically load onto the "H" bus after EDG #1 re-energizes the bus? (assume NO operator action)

A. (1) 1-VS-F-58A (Filtered Exhaust Fan), THEN (2) "EN group pressurizer heaters, THEN (3) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump)

B. (1) 1-VS-F-58A (Filtered Exhaust Fan), THEN (2) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump), THEN (3) "E" group pressurizer heaters C. (1) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump), THEN (2) 1-VS-F-58A (Filtered Exhaust Fan), THEN (3) "E" group pressurizer heaters D. (1) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump), THEN (2) "E" group pressurizer heaters, THEN (3) 1-VS-F-58A (Filtered Exhaust Fan) 50

51. Unit 1 Initial Conditions:

• Radiography operations are in progress on a section of main steam piping.

• The radiographers want to verify the correct position of the camera by using a main steam line radiation monitor located on the same elevation and close to the area where the radiography needs to take place.

• To obtain a baseline reading, the camera source was placed 3.21 feet away from the radiation monitor detector. The radiation monitor read 5.92 Rlhr.

Current conditions:

• The camera has been moved into position to image the piping section.

• Engineering calculations show that the camera should be placed 17.46 feet away from the radiation monitor detector.

The distances listed above include the difference in height from the camera to the radiation monitor detector. Consider the radiography camera as a radiation point source. Carry all calculations to three (3) decimal places.

Based on the current conditions, which ONE of the following correctly identifies the expected reading on the radiation monitor, if the camera was positioned correctly?

A. 0.037 Rlhr B. 0.200 Rlhr C. 1.088 Rlhr D. 2.538 R/hr 51

52. Initial conditions:

• Unit One at 100% power

• 1-SW-P-10A ("A" charging pump service water pump) in AUTO

• 1-SW-P-10B ("B" charging pump service water pump) in HAND Current conditions:

• "A" RSST was lost due to sudden pressure

• All equipment operated as designed Assume sufficient time has elapsed to allow for all automatic actions to occur.

Which ONE of the following states (1) the current status of the charging pump service water pumps and (2) why?

A. (1) Both pumps currently in service (2) 1-SW-P-10B remained on the bus and 1-SW-P-10A started on low discharge header pressure.

B. (1) Both pumps currently in service (2) 1-SW-P-10B remained on the bus and 1-SW-P-10A started due to opposite emergency bus undervoltage.

C. (1) Only 1-SW-P-10B in service.

(2) 1-SW-P-10B remained on the bus and 1-SW-P-10A started on low discharge header pressure but secured upon restoration of 1-SW-P-10B D. (1) Only 1-SW-P-10A in service.

(2) 1-SW-P-10B tripped on undervoltage and 1-SW-P-10A started due to opposite emergency bus undervoltage ..

52

53. Which ONE of the following states the power supplies DIRECTLY to both Unit 1 Rod Drive MG sets and what signal would DIRECTLY open the supply breakers to the MG Sets?

A. 'A' and 'C' 260 Volt Station Service Reactor Trip B. 'A' and 'C' 480 Volt Station Service AMSAC C. 'A' and 'C' 480 Volt Station Service Reactor Trip D. 'A' and 'C' 260 Volt Station Service AMSAC 53

54. A loss of coolant accident (LOCA), coincident with a failure of ALL containment spray pumps to start, causes containment pressure to INCREASE.

Which ONE of the following correctly describes the expected equipment status as containment pressure continues to rise?

A. All containment recirculation fans will operate at pressures up to 17.7 pSia. At 17.7 psia, containment recirculation fans 1A and 1B will automatically trip. Containment recirculation fan 1C will continue to run at pressures greater than 17.7 psia.

B. All containment recirculation fans will operate at pressures up to 23.0 psia. At 23.0 psia, containment recirculation fans 1A and 1B will automatically trip. Containment recirculation fan 1C will continue to run at pressures greater than 23.0 psia.

C. All containment recirculation fans will operate at pressures up to 17.7 psia. At 17.7 psia, all containment recirculation fans will automatically trip.

D. All containment recirculation fans will operate at pressures up to 23.0 psia. At 23.0 psia, all containment recirculation fans will automatically trip.

54

55. Unit 1 Initial Conditions:

• 100% Power.

Current conditions:

• Containment pressure transmitter 1-LM-PT-100B failed a calibration surveillance four (4) days ago.

• All Technical Specification 3.7, "Instrumentation Systems," required actions for 1-LM-PT-100B have been completed.

Based on the current conditions, which ONE of the following identifies (1) the MINIMUM containment pressure, AND (2) the MINIMUM number of OPERABLE containment pressure channels that must actuate in order to close 1-RM-TV-1 OOAlB/C (Containment Particulate and Gas Radiation Monitor 1-RM-RI-159/160 trip valves)?

A. (1) 23.0 psia (2) two (2)

B. (1) 23.0 psia (2) three (3)

C. (1) 17.7 psia (2) two (2)

D. (1) 17.7 psia (2) three (3) 55

56. Unit 1 Initial Conditions:

• 100% Power.

Current conditions:

• A controller failure causes the Unit 1 Operator to place the Charging Flow Controller to MANUAL.

• The Unit 1 Operator attempts to reduce charging flow to 20 gpm to mitigate a high Pressurizer Level.

Based on the current conditions, which ONE of the following correctly describes the behavior of 1-CH-FCV-1122 when the Operator attempts to reduce charging flow to 20 gpm?

A. The Flow Limit Summator no longer limits flow, and 1-CH-FCV-1122 can be manually closed to allow 20 gpm flow.

B. The Flow Limit Summator no longer limits flow; however, 1-CH-FCV-1122 can only be manually closed to allow 25 gpm flow.

C. The Flow Limit Summator will limit charging flow to a minimum of 25 gpm.

D. The Flow Limit Summator will limit charging flow to a minimum of 30 gpm.

56

57. Initial Unit 1 Conditions:

- Unit 1 is at 100% power

- All control rods are fully withdrawn Current Unit 1 Conditions:

- A control rod in the 'D' Control Bank drops to the bottom of the core

- Unit 1 is at 70% power

- Delta flux is within band Which ONE of the following correctly states (1) the control rod select switch position to recover the rod in accordance with 0-AP-1.01, Control Rod Misalignment AND (2) when the step counters are required to be reset in accordance with 0-AP-1.01?

A. (1) Place the ROD CONT MODE SEL SWITCH to MANUAL for rod recovery.

(2) Step counters are required to be reset prior to rod recovery.

B. (1) Place the ROD CONT MODE SEL SWITCH to the affected bank for rod recovery.

(2) Step counters are required to be reset prior to rod recovery.

C. (1) Place the ROD CONT MODE SEL SWITCH to MANUAL for rod recovery.

(2) Step counters are NOT required to be reset until after the rod is recovered.

D. (1) Place the ROD CONT MODE SEL SWITCH to the affected bank for rod recovery.

(2) Step counters are NOT required to be reset until after the rod is recovered.

57

58. Unit 1 plant conditions; Unit Startup in progress following a reactor trip at Middle of Life Reactor power = 90%

1-GOP-1.5 UNIT STARTUP, 2% REACTOR POWER TO MAX ALLOWABLE POWER in progress Axial Flux Difference = 0 All Control Rods are Fully Withdrawn at 225 steps Based on the above conditions; (1) which ONE of the following states the maximum rate at which power can be increased to 100% lAW 1-GOP-1.5 (2) how will Axial Flux Difference change as power is increased?

A. (1) 3% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) become positive B. (1) 3% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) become negative C. (1) No rate limitation exists (2) become positive D. (1) No rate limitation exists (2) become negative 58

59. Initial Unit 2 conditions:

- Reactor power = 100% steady state Current Unit 2 conditions:

- A LOCA is in progress E-1, Loss of Reactor or Secondary Coolant, is being performed

- All RCPs have been stopped

- Containment pressure = 47 psia and slowly increasing

- Total AFWflow = 485 gpm

- SG WR levels are: "A" = 48%, "S" = 40%, "C" = 39%

- RCS pressure = 920 psig

- IR Nls = 2E-11 amps, with SUR = 0

- CETCs indicate 600°F

=

- RVLlS full range 40%

Which ONE of the following correctly states the procedure to which the control room crew is required to transition?

A. FR-C.1, Response to Inadequate Core Cooling S. FR-C.2, Response to Degraded Core Cooling C. FR-Z.1, Response to Containment High Pressure D. FR-H.5, Response to Steam Generator Low Level 59

60. Unit 1 initial conditions:

Reactor trip from 40% power SI actuated 1-E-0 REACTOR TRIP OR SAFETY INJECTION initiated Current plant conditions:

A NR SG level = 26% decreasing B NR SG level = 22% decreasing C NR SG level = 29% decreasing Based on the above plant conditions; which ONE of the following correctly states (1) the MINIMUM SG level at which the first signal to start an AFW pump occurs (assuming no operator action) and (2) the MINIMUM required Steam Generator Narrow Range Level that allows SI flow reduction lAW 1-E-0 REACTOR TRIP OR SAFETY INJECTION?

A. (1) 13%

(2) Greater than 12%

B. (1) 13%

(2) Greater than 22%

C. (1) 17%

(2) Greater than 12%

D. (1) 17%

(2) Greater than 22%

60

61. Unit 1 Initial Conditions:

• 100% power.

• Rod control is selected to P-446, Channel III turbine first stage impulse pressure.

• P-447, Channel IV turbine first stage impulse pressure, fails LOW.

• Annunciator 1H-D7, "STM DUMP PERM," is lit.

Current conditions:

• No operator actions have been performed to address the P-447 failure.

• Control rods are INSERTING in automatic.

• Steam flow on all channels is INCREASING.

Based on the current conditions, which ONE of the following correctly identifies the cause?

A. A main steam line safety valve has lifted and will not reseat.

B. Median Tave has failed HIGH.

C. P-446 has failed HIGH.

D. P-464, Steam header pressure, has failed HIGH.

61

62. Unit 1 Initial Conditions:

o Unit was at 28% power when condenser vacuum began to degrade.

Current conditions:

o Annunciator 1E-E3, "DELTA FLUX DEVIATION," is lit.

o Annunciator 1G-H8, "ROD BANK DEXTRA LO LIMIT," is lit.

o Annunciator 1F-B6, "TURB LO VAC," has been lit for five (5) minutes.

o Control rods are inserting in automatic.

o Condenser vacuum continues to DECREASE with no signs of recovery.

o Turbine is at 14% load and DECREASING.

Based on the current conditions, which ONE of the following identifies the required operator action, in accordance with 1-AP-14.00, "LOSS OF MAIN CONDENSER VACUUM?"

A. Commence an emergency boration using 1-AP-3.00, "EMERGENCY BORATION."

B. IMMEDIATELY trip the Reactor and enter 1-E-O, "REACTOR TRIP OR SAFETY INJECTION."

C. IMMEDIATELY trip the Turbine and stabilize the unit using the steam dumps.

D. IF condenser vacuum is less than 24.5 in-Hg for a five (5) minute period, THEN trip the Turbine and stabilize the unit using the steam dumps.

62

63. Unit 1 initial conditions:

Shut down for refueling Containment purge in progress Current plant conditions:

A High radiation signal on the containment particulate (1-RM-RI-159 CTMT PARTC) radiation monitor occurs Based on the above conditions which ONE of the following correctly states the status of the containment purge components?

A. Containment purge supply fans (1-VS-F-4A and B) off Containment purge supply MOVs (1-VS-MOV-100A and B) closed, Containment purge discharge MOVs (1-VS-MOV-100C and D) closed B. Containment purge supply fans (1-VS-F-4A and B) off Containment purge supply MOVs (1-VS-MOV-100A and B) closed, Containment purge discharge MOVs (1-VS-MOV-100C and D) open C. Containment purge supply fans (1-VS-F-4A and B) on Containment purge supply MOVs (1-VS-MOV-100A and B) closed, Containment purge discharge MOVs (1-VS-MOV-100C and D) open D. Containment purge supply fans (1-VS-F-4A and B) on Containment purge supply MOVs (1-VS-MOV-100A and B) open, Containment purge discharge MOVs (1-VS-MOV-100C and D) open 63

64. Plant initial conditions:

Reactor Power = 100% both units Instrument Air Systems are in their NORMAL alignments (split out)

Operator reports an air leak on Unit 1 instrument air header Unit 1 instrument air pressure is currently 85 psig and DECREASING Based on the above conditions, which ONE of the following correctly states the expected status of (1) Unit 2 Instrument Air pressure and (2) the Unit 1 Instrument Air Compressor?

A. (1) Unit 2 Instrument Air pressure will decrease until air pressure equals 80 psig.

(2) operating.

B. (1) Unit 2 Instrument Air pressure will decrease until air pressure equals 80 psig.

(2) NOT operating.

C. (1) Unit 2 Instrument Air pressure will decrease until 2-IA-C-1 starts.

(2) operating.

D. (1) Unit 2 Instrument Air pressure will decrease untiI2-IA-C-1 starts.

(2) NOT operating.

64

65. Which ONE of the following states (1) the capacity of each Fire Water Tank and (2) if a domestic water leak occurred, the tank level at which the leak would stop?

A. (1) 250,000 gallons per tank (2) 200,000 gallons per tank B. (1) 300,000 gallons per tank (2) 250,000 gallons per tank C. (1) 300,000 gallons per tank (2) 200,000 gallons per tank D. (1) 250,000 gallons per tank (2) 50,000 gallons per tank 65

66. Which ONE of the following (1) correctly states the maximum allowable length of a valve wrench used lAW OP-AA-100, Conduct of Operations, AND (2) whether OP-AA-100 allows a valve wrench to be used on manual valves as well as motor operated valves (MOVs)?

A. (1) Valve wrench length is limited to approximately 1.5 times the handwheel diameter.

(2) Valve wrench is permitted to be used on manual valves but not MOVs.

B. (1) Valve wrench length is limited to approximately 2.0 times the handwheel diameter.

(2) Valve wrench is permitted to be used on manual valves but not MOVs.

C. (1) Valve wrench length is limited to approximately 1.5 times the handwheel diameter.

(2) Valve wrench is permitted to be used on both manual valves and MOVs.

D. (1) Valve wrench length is limited to approximately 2.0 times the handwheel diameter.

(2) Valve wrench is permitted to be used on both manual valves and MOVs.

66

67. Unit 1 Initial Conditions:

• Core re-fueling operations are in progress.

• Approximately 3/4 of the new core has been loaded without incident.

Current conditions:

• One Source Range count rate is double (2X) the initial reference value.

• The other Source Range count rate is (1.75X) (less than double) the initial reference value.

• The 11M plot is approaching 0.65.

Based on the current conditions, which ONE of the following identifies the MINIMUM conditions that would require stopping core alterations, in accordance with the Precautions and Limitations of 1-0P-FH-001, "CONTROLLING PROCEDURE FOR REFUELING?"

A. Core alterations are required to be stopped immediately and subcriticality reevaluated.

B. Core alterations may continue, but IF BOTH Source Range count rates reach one doubling from the reference value, then core alterations are required to be stopped immediately and subcriticality reevaluated.

C. Core alterations may continue, but IF the 11M plot approaches 0.5, then core alterations are required to be stopped immediately and subcriticality reevaluated.

D. Core alterations may continue, but IF BOTH Source Range count rates reach one doubling from the reference value, AND the 11M plot approaches 0.5, then core alterations are required to be stopped immediately and subcriticality reevaluated.

67

68. Unit 1 initial conditions:

Core re-Ioad in progress SR NI background count rate = 10 cps Current plant conditions:

1G-C1. NIS SOURCE RNG SHUTDN HI FLUX. alarms Based on the above conditions. which ONE of the following correctly states (1) the minimum count rate that would cause the alarm and (2) what actions are directed by ARP 1G-C1?

A. (1)42cps (2) Direct the refueling SRO to place fuel in a safe condition and evacuate containment.

B. (1) 42 cps (2) Emergency borate and direct the refueling SRO to stop all refueling activities.

C. (1) 60 cps (2) Direct the refueling SRO to place fuel in a safe condition and evacuate containment.

D. (1) 60 cps (2) Emergency borate and direct the refueling SRO to stop all refueling activities.

68

69. With the unit initially at 100% power, 1-MS-PT-1446 (Ch III Impulse pressure) fails to 0%. Unit conditions are as follows:

• Reactor power 95% - by delta-T

• "Dn bank control rods at 212 steps

• Delta Flux is currently -7

• Delta flux target is -1.5

• Tave is 568F

• Tref is 572F

• Steam generator levels are stable at 33% narrow range.

• RCS pressure is currently 2200 psig Which ONE of the following states the MOST LIMITING LCO (if any), and required actions?

A. No LCO actions are required.

B. 15 minute clock to restore delta-flux in band due to delta flux being outside target band.

C. 30 minute clock to restore pressurizer pressure as pressurizer pressure is outside the allowable band.

D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> clock to verify permissive status due to pimp failure to P-10 and P-7 interlocks 69

70. Unit One at 100% power and stable.

Chemistry morning report has been issued with the following parameters given:

- Unit One "A" safety injection accumulator (1-SI-TK-1A) - boron concentration- 2235 ppm.

- Unit One "8" safety injection accumulator (1-SI-TK-18) - boron concentration- 2500 ppm.

- Unit One "c" safety injection accumulator (1-SI-TK-1C) - boron concentration- 2300 ppm.

- "A" Waste Gas Decay Tank- 1.65% oxygen concentration.

- "A" Waste Gas Decay Tank- 25,720 curies.

Which ONE of the following states ALL the above parameters that require entry into a Technical Specification LCO?

A. "A" accumulator boron and "A" Waste Gas Decay Tank 02 content

8. "8" accumulator boron and "A" Waste Gas Decay Tank O2 content C. "A" accumulator boron and "A" Waste Gas Decay Tank curie content.

D. "8" accumulator boron and "A" Waste Gas Decay Tank curie content.

70

71. You are assigned to oversee work being performed in a Radiation area.

Which ONE of the following describes: (1) the types of radiation that are measured by the "DAD" and (2) the requirements for DAD placement if work is to be performed in a contaminated area?

A. (1) Gamma & X-Ray ONLY (2) Inside protective clothing with TLD B. (1) Gamma & X-Ray ONLY (2) Outside protective clothing in a whirlpack

c. (1) Gamma, Beta and Neutron (2) Inside protective clothing with TLD D. (1) Gamma, Beta and Neutron (2) Outside protective clothing in a whirl pack 71

72. While taking LOGS in the auxiliary building, a mechanic, who is performing an overhaul on 1-CH-P-1A ("An charging pump), approaches you and asks for assistance in lifting the auxiliary oil pump. He states that he will only require your assistance for 20-30 minutes.

Which ONE of the following states the proper response to this request?

A. Provide assistance and when logs are complete, ask health physics to assign the dose received while helping the mechanic to the mechanic's RWP.

B. Render the requested assistance on Operations RWP as long as the dose received will not cause you to reach either your DOSE RATE LIMIT or DOSE LIMIT.

C. Inform the mechanic that you are unable to render the requested assistance while on the current Operations RWP.

D. Call health physics shift supervisor and request to be placed on the mechanic's RWP. Whe n complete, contact the health physics shift supervisor again, and get reassigned to the normal operations RWP.

72

73. Unit 1 initial conditions:

Reactor Trip Critical safety functions as follows.

SUBCRITICALITY - GREEN HEAT SINK - ORANGE CORE COOLING - ORANGE INVENTORY - YELLOW CONTAINMENT - RED INTEGRITY - ORANGE Based on the above conditions, when addressing Critical Safety Functions (CSFs) which ONE of the following CSFs has the highest priority and should therefore be addressed first?

A. Heat Sink B. Core Cooling C. Containment D. Integrity 73

74. Unit 1 initial conditions:

Loss of all feedwater has occurred from 100% power EOPs are progress Current plant conditions:

Transition to 1-FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK has just been made 1A Wide Range SG level = 4% decreasing 1B Wide Range SG level = 5% decreasing 1C Wide Range SG level = 6% decreasing RCS pressure = 2300 psig increasing

=

CETC 580 of increasing All RCPs are secured Based on the above conditions: (1) which ONE of the following actions are directed by 1-FR-H.1 and (2) why?

A. (1) Commence bleed and feed (2) At least Two SGs are considered dry so transition to another form of decay heat removal must be made before conditions degrade further B. (1) Commence bleed and feed (2) RCS pressure may reach pressurizer safety valve setpoints. so transition to another form of decay heat removal must be made to prevent water relief through the safety valves.

C. (1) Cross Connect with Unit 2 AFW and feed at the maximum available rate (2) To reduce RCS temperature to < 550 OF for establishing a heat sink D. (1) Cross Connect with Unit 2 AFW and feed at the maximum available rate (2) To increase SG level to > 7% in any SG for establishing a heat sink 74

75. Which ONE of the following describes that actions required on a failure of the reactor to trip (ATWS) in accordance with 1-FR-S.1 (Response to Nuclear Power Generation/ATWS)?

A. Place rod control in MANUAL and manually trip the turbine. If turbine will not trip, then close the main steam trip valves. Manually insert control rods.

B. Place rod control in MANUAL and manually trip the turbine. If turbine will not trip, then reduce limiter to zero. Manually insert control rods.

C. Place rod control in AUTOMATIC and manually trip the turbine. If turbine will not trip, then reduce limiter to zero. Verify automatic rod insertion.

D. Place rod control in AUTOMATIC and manually trip the turbine. If turbine will not trip, then close the main steam trip valves. Verify automatic rod insertion.

You have completed the testl 75

• EAL Tables

• Steam Tables

• Calculator

• This Document

NUMBER PROCEDURE TITLE REVISION 25 O-AP-12.01 LOSS OF INTAKE CANAL LEVEL PAGE 6 of 11 ACTION/EXPECTED RESPCNSE RESPONSE NOT OBTAINED NOTE: The NSS Point of contact is identified on the Plan of tha Day (POD).

17. CONSERVE INTAKE CANAL INVENTORY: o Notify NSS Point of Contact to Initiate the o . Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. start all available ESW INTAKE CANAL ALTERNATE MAKEUP pumps lAW O-OP-SW-002. GUIDELINE.

EMERGENCY SERVICE WATER PUMP OPERATION 1B. CHECK CW PUMPS - AT LEAST ONE o start CW pumps lAW OP-4S.1.1.

RUNNING STARTING ANY CW PUMP.

19. STOP ALL LIQUID RELEASES:

o LW o BR o CP BLDG sumps

20. CHECK SW TO RS HXS ON EITHER UNIT - o GO TO step 23.

IN SERVICE

NUMBER PROCEDURE IDLE REVISION 25 O-AP-12.01 LOSS OF INTAKE CANAL LEVEL PAGE 7 of 11 ACTION/EXPECTED RESPONSE RESPONSENOTOBT~NED NOTE: Based on heat load. UNIT AT POWER is defined as any Un~whlch is actually at powerQ!: any Un~

which has been shutdown for less than 35 days.

21. DETERMINE ALLOWABLE CC HX f'!W OUTLET VALVE POSITIONS FOR NON-ACCIDENT UNIT HXs:

Allowable CC HXs and SW outlet INITIAL UNIT CONDITIONS valve po~ns on non-accident Untt BOTH Units at Power 2 CC HXs with SW outlet valves open 19 turns for each HX 1 Unlt.t power and CONDITION 1:

1 Unit shutdown for greater than With 1 ESW pump operafing 35 days

• 1 CC HX whh SW outlet valve open 14 turns CONDITION 2:

WIth 2 ESW pumps operafing

• 1 CC HX whh SW outlet valve open 19tums

22. GO TO STEP 24

NUMBER PROCEDURE TITLE REVISION 25 O-AP-12.01 LOSS OF INTAKE CANAL LEVEL PAGE 8 of 11 ACTION! EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: Based on haatloed, UNIT AT POWER Is daflned asany Unllv.tllch Is acluaBy at powarQ!: any Un~

v.tllch has been shutdown for less than 35 days.

23. DETERMINE ALLOWABLE CC HX SW OUTLET VALVE POSITION FOR HXs FOR BOTH UNITS Allowable CC HXs and SW outlet INITIAL UNIT CONDITIONS valve posIHons BOTH Un~ at Power Crosstle CC. 3 CC HXs aUowed with SW outlet valves 19 turns open for each HX.

• , UnH at power CONDITION 1: 1 ESW pump operatlng AND

• Unit at Power.

• 1 UnH shutdown for greater than 35 2 CC HXs with SW outlet valves days open 19 turns for each HX.

• UnH shutdown:

1 CC HX with SW oUtlet valve open 14 turns.

CONDITION 2: 2 ESW pumps oper.~ng

• UnH at Power:

2 CC HXs with SW oUtlet valves open 19 turns for each HX.

• UnH shutdown:

1 CC HX with SW oUtlet valva open 19 turns.

BOTH Units shutdown for greater 1 CC HX for ""ch UnH with SWouHet than 35 days valve open 19 turns for each HX.

DOMINION VPAP-2802 REVISION 31 PAGE 79 OF 205 6.3.3 One-hour Notlf1catlons NOTE: Some condilions. indicaled by "See EPIP-I.O I,.' may exceed an Emergency Action Level (EAL) as specified in EPIP-l.Ol. Emergency Manager Controlling Procedure.

If n condition exceeds no EAL. EPIPs control State and Federal agency notifiCiltions.

If an event or condition does not exceed an EAL. 1t may still be reportable in accordance with this procedure.

As soon as practical, but within one hour, the Shift Manager, Station Emergency Manager, or Sile Vice President shall notify the NRC Operations Ceoter of:

a. Deviation from Technical Specifications (permiued by 10 CPR 50.54(x>>) to protect the health and safety of the public, when no action consistent with license conditions and Technical Specifications cnn provide adequate or equivalent protection. [10 nOR 5O.71lb)(J)1

b. An automatic safety sY5lem that does not function as required during opemtion. See EPlP-l.O J. 110 eFR ;'IO~'7,6(d)(I)(Il)(A)1 NOTE: Notifications required by Sleps 6.3.3.c .. 6.3.3.d., and 6.3.3.e., are exempt from the requirement thot Safeguards Infonnation be transmitted only by protected telecommunications circuit" approved by NRC.

C. An accidental criticality or loss of SNM. See EPIP-1.0 1.

110 Cl'""K 70.52 (a). 10 Cl<1l72.74{a), 10 en 14.1111)

DOMINION VPAP-2802 REVISION 31 PAGE 80 OF 205 NOTE: Step 6.3.3.d notifications need not duplicate Step 6.3.3.e. notifications.

110 CFR 74.11(c)'10 CFR 11-14(c)1

d. A loss of any 110 eFR 73.71(a){I), 10 CFR 73.67Itl)(3)(vm, 10 CFR 73.67(g)t3)(Wl\:

• SNM shipment

• Spent fuel shipment or Availability of supplemental infonnation after initial notification. (10 Cf.'R 73.7J(1l)(SJl (See also Step 6.15.3.a.3.)

or Recovery of or accounting for such lost shipment See also Step 6.153.a.2. [10 Ct'R 7J.71{Q)(i). 10 CFR 73..67(eJ(3J(rtI},10 CFR 73.67(g)(3)(llIl1 NOTE: Steps 6.3.3.e., 6.33.f.. 6.3.3.g., 6.3.3.h.notifications need not duplicate Step 6.3.3.d.

or to CFR 50.72 notifications. [10 ern 72.74(c), 10 CFR 73.71(e). 10 CFR '-I.ll(d]

e. A reason to believe that a person has committed or caused. or attempted to commit or cause, or has made a credible threat to commit or cause (See also Step 6.15.3.b.2.).

[10 CFR 73.71(b)(1), 10 CIfR 73 App. G.t, 10 t.'FR 7052 (R), 10 CFR 72.74{1I), 10 O"R 74.11(a)):

• Theft, loss, or unlawful diversion of SNM

• Significant physical damage to the Station, nuclear fuel. or carrier of nuclear fuel

• Interruption of normal operation through unauthorized use of or tampering with its machinery. components, or controls, including the security system

f. Unanthorized entry into a protected area. material access area. controUed access area. vital area., or transport

g. Failure. degradation, or the discovered vulnerability in a safeguard system that could allow unauthorized or undetected access to a protected area, controlled access area, vital area. or transport for which compensatory measures have not been employed.

DOMINION VPAP-2802 REVISION 31 PAGE 81 OF 205 NOTE: FitneS$-for-duty events are reported in accordance with 10 CFR 26 instead of 10 CFR 73.71. See Steps 6.3.6.b. and 6.8.1. 110 CFR ".73<<)]

h. Actual or attempted introduction of contraband into a protected area, material access area. or transport.

i. Discovery that an undeclared or misclassified event or condition met all the following criteria:

• Exceeded an Emergency Action Level (EAL) as specitied in EPIP-1.0 I.

Emergency Manager Controlling Procedure

• The basis for the emergency class no longer exists at the time of discovery

• No other reasons exist for an emergency declaration In addition, the following shall be notitied:

• Department of Emergency Management (at approximalely the same time)

• Director Nuc1ear Protection Services and Emergency Preparedness

)OMINION VPAP-2802 REVISION 31 PAGE 82 OF 205 6.3.4 Four-hour Notifications NOTE: Some conditions. indicated by "See EPIP-I.OI." may exceed an Emergency Action Level (EAL) as specified in EPIP-I.O I, Emergency Manager Controlling Procedure.

If a condition exceeds an EAL, EPIPs control State and Federal agency notifications.

If an event or condition does not exceed an EAL. it may still be reportable in accordance with this procedure.

a. As soon as practical, but within foar hours. the Shift Manager shall notify the NRC Operations Center via the ENS of:

NOTE: If a unit enters a limiting condition for operation (LCO) and a unit shutdown is started due to the LCO, the event Is reportable even if shutdown is not completed. LCOs terminated by a unit shutdown for an unrelated reason are still reportable if tbe condition would not have been corrected within the LCO time limit for shutdown.

I. Initiation of plant shutdown (reduction of power or temperature) required by Technical Specifications. The initiation of plant shutdown does not include mode changes required by Technical Specifications if initiated after the plant is already in a shutdown condition. See EPIP- 1.01. [10 CFR,o.72(bJl2XI),

10 CFR SO.3ti(d)(1)(I)(A).10 CFR50.36 (d)(2)(I), NlfREG Hl21ltml.l.2.J]

2. Any event that results or should have resulted in ECCS discharge into the RCS as a result of a valid signal except wben the actuation results from and is part of a pre-planned sequence daring testing or reactor operatiolL [10 CFR SO.72(b)(2HIYXA)1

3. Any event or condition that results in actuation of tbe reactor protection system (RPS) when the reactor is critical except wben actuation results from and is part of a pre-planned sequence during testing or reactor operation.

[10 CFR 5O.7l{b)(2)1,h)tB)J

DOMINION VPAP-2802 REVISION 31 PAGE 83 OF 205 NOTE: "Notification to other government agencies has been or will be made" is not necessarily an automatic notification to the NRC. Refer to NUREG - 1022. Event Reporting Guidelines 10 CFR 50.72 and 50.73. for discussions and examples or contact Station Licensing if clarification is needed. INUREG-IOl2, section 3.2.121

4. Any event or situation, related to the health and safety of the public or onsile personneL or protection of the environment, for which a news release is planned, or notification to other government agencies has been or will be made.

Such an event may include an onsile fatality or inadvertent release of radioactively contaminated materials. [Commitment 3.2.16] 110 CFR SO.72~bX2)(xlll

5. ISFSI Non-emergency Four-Hour Notifications shall include, if available at time of notification: flU CJo'R T2.7S(eJl:JlI

• The caBer's name and call back telephone number

• A description of the event, including time and date

• The exact location of (he event

• The quantities. and chemical and physical forms of the spent fuel. HLW or reactor related Greater than Class C (GTCC) waste involved

• Any personnel radiation exposure data

6. An action taken in an emergency thntdepnrts from a license condition. technical specification. or certificate of compliance when the action is immediately needed to protect Ihe puhlic health and safety and no licensed action that provides adequate or equivalent protection is immediately apparent-see Step 6.14.7 J. [10 eFR 72.15(b):l)]

7. An event at the [SFSI lhllt requires unplanned medical treatment at an offgire medical facility of an individual with radioactive contamination on the individual's clothing or body which could cause further radioactive contamination. (10 (""FR 72.75(e)(.3)]

DOMINION VPAP-2802 REVISION 31 PAGE 84 OF 205

8. Groundwater Protection Voluntary Communication Notifications to other government agencies may be reportable under 10 CFR 50.72(b)(2)(xiJ requirement for a 4-hour notification to the NRC operations center based upon the following guidance:

• If a licensee is notifying a local, state, or other federal agency in accordance with an existing law, regulation, or ordinance, (hen the licensee should make its notification to the NRC under the SO.72 notification requiremenl

• If a licensee is informally communicating with a local, state, or other federal agency (i.e .. not under a specific law, regnlation or ordinance), lhen lhe licensee has discretion as to whether to informally communicate with NRC (e.g., through the site resident inspector aneVor regional NRC office) or formally through the SO.72 notification process. If due to the site-specific circumstances or heightened sensitivity to the issue at that site. the issue is likely to produce strong media interest, then the licensee should consider notifying NRC under the 50.72 requirement becaose this is actually the underlying intent of the regulation.

DOMINION VPAP-2802 REVISION 31 PAGE 85 OF 205

b. Any person at the Station who observes smoke originating from Station equipment being released into the outdoor atmosphere shnl1 notify the Shift Manager as soon as possible.

1. If the smoke is not from a fire and there are no certified visible emissions evaluators available to determine the opacity of the smoke being released to the outdoor atmosphere, the Shift Manager or other Station personnel shall take the appropriate steps to determine the source. cause, and duration of the smoke being released .

• Once all of the pertinent information regarding the relea"ie of smoke has been obtained, the Electric Environmental Services (ESS) must be notified immediately .

• The ESS will reporl the release of smoke into the outdoor atmosphere to the appropriate DEQ regional oft1ce as soon as practical, but no later than four daytime business hours of the occurrence. with all of the pertinent information. If the DEQ regional oft1ce determines that it is necessary to obtain smoke readings after receiving all of the pertinent information. the ESS will dispatch a certified visible emissions evaluator to the Station to determine the opacity of the smoke being released into the outdoor atmosphere.

2. The ESS will prepare and submit aoy written reports to the DEQ regional office regarding the release of smoke into the outdoor atmosphere.

DOMINION VPAP-2802 REVISION 31 PAGE 86 OF 205 6.3.5 Eight-hour Notifications

a. As won as practical, but within eight hours, the Shift Manager shall notify the NRC Operations Center via the ENS of:

I. Any condition that results in the condition of the Station, including its principal safety barriers, being seriously degraded 1'0 CFR "',72(h)(3)(U)(AIi 2, Any event or condition that results in the Station being in an unanalyzed condition that significantly degrades plant safety, ['0 CFR 5O,72,h)(,l)('HB)]

3. Any event or condition that results in valid actuation of any of the following systems, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation: uo arR SO.72(b)(3)(lv)jA))

• Reactor Protection System (RPS) - (RPS actuation with the reactor critical may be reportable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> under !O CFR 50.72(b )(2)(iv)(B), see Step 6.3.4.a.3.)

• General containment isolation signals affecting containment isolation valves in more than one system or multiple Main Steam Isolation Valves (MSIVs)

• Emergency Core Cooling Systems (BCCS) including HHSI and LHSI (Actual discharges are reportable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> under 10 CFR 50.72(b)(2)(iv)(A), see Step 6.3.4.a.2.)

• Auxiliary Feedwater System

• Containment heal removal and depressurization systems including Containment spray and fan cooler systems

• Emergency Diesel Generators (EDGs)

4. Any event or condition that at the time of discovery could have prevented the fuitillment of the safety function of structures or systems that are needed to:

• Shut down the reactor and maintain it in a safe shutdown condition

• Remove residual heat

• Control the release of radioactive material: or

• Mitigate the consequences of an accident. See EPIP-l.O 1. 110 CFR SG.72(b)t3)(v)]

5. Any event requiring the transport of a radioactively contaminated person to an off-site medical facility for treatment. See also Step 6.27.2. ['0 CFR 50.72 (bl(3I(x11I[

Could also be a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report in accordance with 10 CFR 72.75 (b)(5).

DOMINION VPAP-2802 REVISION 31 PAGE 87 OF 205

6. An event that results in a major loss of emergency assessment capabilityl.

off-site response capability, or off-site communications capability, e.g .*

WlavailabiJity of any of the following (see Attachment 3, Emergency Response Unavailability, for unavailability criteria)2:

• Safety Parameter Display System3 (SPDS)

• Emergency response facilities4 (see Subsection 4.15)

• Emergency communications facilities and equipmenti

• Prompt Notit1cation System, including sirens

• Plant monitors necessary for accident monitoring See EPIP-l.O 1. t10 OR 50."(',(,_,)

7. Any instance of:

• A defecl in any spent fuel storage cask structure, system. or component thal is important to safety [10 eFR 72.7S(c)lJ or

• A significant reduction in the effectiveness of any spent fuel storage cask confinement system during use of the storage cask 110 CFR 7l.75(c}2 See EPIP-1.0 1.

b. If an Alert, Site Area Emergency. or General Emergency is declared:

1. The Station Coordinator Emergency Preparedness shall prepare a Summary Report from information in completed Emergency Plan Implementing Procedures, Control Room logs. and interviews with persons involved with the declaration and response, as appropriate. See Attachment 8, Example DEM Summary Report.

1. A major Joss of emergency ussesb1l1ent cnpability include~ eV~Dts thllt Significantly impair fulfiIbnent of the Emergency P1an. including safety assessment capability (e.g.. loss ofa significant portion ofContro] Room indicatioos). Loss of on*site meteorological information does not constitute a major loss of assessment capability and should not be reported under tbis part.

2. Ensineeringjtldgment may be needed. to asses~ the significance of losing certain equipment.

3. Unavailability of only the SPDS (one function of the Plant Computer System (PeS>> for less than eight hours is Dot reportnble. but unavailability of the SPDS and other assessment capability at the same time may be reportable. Scheduled pes outages or operution of PCS in the Simulator mode are not reportable if the SPDS can be made available in less thnn one hour.

4. EOF 1055 is reportable only jf both the LEOF and the CEOF are unavailable.

5. A momentary loss of off-site response cnpability or emergency communications (e.g., the bncl.:up power supply fails while security computer and emergency !..'ommunicatlons ore tempornriiy connected to perform a surveillance test) is not reportable.

DOMINION VPAP*2802 REVISION 31 PAGE 88 OF 205

2. The Site Vice Presiden~ Director Nuclear Station Safety and Licensing, or Plant Manager (Nuclear) shall approve the reporL

3. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after temtination of the event. Nuclear Emergency Preparedness shall ensure the report is delivered to the State Coordinator of the Virginia Department of Emergency Management. INAEP **" SEP .....]

c. If, on Dominion property or at Lake Anna Dam, there is a Dominion employee or contractor fatality (regardless ohhe time between the injury and death, or the length of an illness) or an event in which three or more Dominion employees or contractors are hospitalized:

1. The Shift Manager shall notify Supervisor Nuclear Site Safety (Station) with the following information:

• Number offatalities

• The employer of those killed

• The circumstances of the event

• The extent of injuries

2. Nuclear Site Safety (Station) sball notify OSHA as specified in Step 6.3.5.c.3.

See also Step 6.3.4.a.4.

3. Within eight hours after the occurrence, the Supervisor Nuclear Site Safety (Station) (as specified in Step 6.3.5.h.2.) shall notify See Step 6.3.l.a.) the Area Director of OSHA by telephone or facsimile. See Step 6.1.1.a. See also Step 6.3.4.0.4. [19 CFR 1904.8]

d. Whenever fire protection systems, portions of a system. or equipment are impaired or reduced in status for other than scheduled maintenance or scheduled testing activities (meaning an unplanned failure or state of degradation), the Shift Manager shall notify the Supervisor Nuclear Site Safety (Station). [Commitment 3.2.21]

(Surry)

North Anna notification to the Supervisor Nuclear Site Safety (Station) is within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per TRM requirements.

DOMINION VPAP-2802 REVISION 31 PAGE 89 OF 205 6.3.6 Twenty-four Hour NotifiClltlons

a. As soon as practical, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. the Shift Manager shall notify the NRC Operations Center with the ENS OfIIUCF1t20.2201tb)l:

NOTE: The requirements of Step 6.3.6.a.l. do not apply to doses that result from planned special exposures, that are within the limits for planned special exposures. and that are reported in accordance with Step 6.10.1 Le. [10 O"R lo.2lO2(e)]

I. An event that involves licensed material p05sessed by Dominion that may have caused or threatens to cause:

• An individual to receive. in a period of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

• • A total effective dose equivalent exceeding 5 ferns

- An eye dose equivalent exceeding 15 rems

• • A shallow-dose equivalent to the skin or extremities exceeding 50 rerns

• Release of radioactive material inside or outside a restricted area, so that, if an individual had been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, they could have received an intake in excess of one occupational annuallirnit on intake.

If an event involves radiological m'erexposure. DEM must be notified as specified in Step 6.27.2. See also Step 6.6.3.c.

2. ISFSI Twenty-Four Hour Notifications shall include. if available at time of notification: 110 CFR 72.73(t!)(.3)1

• The caller's name and call back telephone number

• A description of the event. including time and date

• The exact location of the event

• The quantities. and chemical and physical fonn of the spent fuel or HLW involved

• Any personnel radiation exposure data

3. An unplanned contamination event that requires access to the contaminated area by workers or the public to be restricted for more tban 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by imposing additional radiological controls or by prohibiting entry into the area [IOCfR 12.75fc)(1>>)

DOMINION VPAP-2802 REVISION 31 PAGE 90 OF 205

4. An event in whicb safety equipment is disabled or fails to function as designed when: (10 c.,lt 72.7~(d)(nl

• The equipment is required by regulation. license condition, or certit1eate of compliance to be available and operable to prevent releases that could exceed regulatory limits, to prevent exposure to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an acciden~

and

• No redundant equipment was available and operable to perform the required safety function

5. An event that prevents immediate actions necessary to avoid exposures to mdiation or radioactive material that could exceed regulatory limits or releases of radioactive materials that could exceed regulatory limits (e.g .. events such as fires., explosions. and toxic gas re]eases)-see Step 6. 14.7.f. nOCli'R 72..75(d)(1)(I)J

b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after discovery of a significant fitness for duty event, a Director shall notify the NRC Operations Center by telephone. See Step 6.1.1. iIOCFR26.73(bJl I. The notifier shaH document the notification in Section B of Attachment 4.

Significant Fitness for Duty Event NRC 24 Hour Notification.

2. The notifier shaH return tbe completed original of Attachment 4 to the Fitness for Duty Administmtor (Station) for further processing. See Step 6.8.1.

c. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Shift Manager shall notify NRC by telephone, telegraph, or facsimile, of any occurrence of an unusual or important event-causally related to Station operation-iliat indicates or could result in significant environmenta1 impact See also Step 6.26.2.b. (North Anna) n<APS EP.4.1 ......~i

d. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after discovery, Licensing (Station) shall notify (see Step 6.3.1.0.)

the NRC Regional Office by telepbone of failure to notify NRC of planned removal or significant change in the normal operation of equipment that controls the amount of radioactivity in Station effluents (North Anna).

INAPS Unllll.Icm~1.C(3)(b); unU 2 Urease. 2.C(3)(1I).1 By the fIrst business day after discovery, Licensing (Station) shall confirm the telephone notification by telegram. mailgram. or facsimile to the NRC Regional Office. See also Step 6.23.6.

DOMINION VPAP-2802 REVISION 31 PAGE 91 OF 205

e. If any unpermitted. unusual. or extraordinary discharge t enters or could be expected to enter State waters. as soon as possible, but not later than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> afterdisCQvery.

Electric Environmentnl Services shall notify (see Step 6.3.1.a.) the State Department of Environmental Quality (Water). See also Steps 6.3.4.a4., 6.3.2.f..

and 6.27.3.0. [VPDES Pl!rmHI

f. If an unplanned bypass (i.e.* intentional diversion of waste streams) occurs from any portion of a treatment worles, as soon as possible. but not later than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the bypass occurs. Electric Environmental Services shall notify (see Step 6.3. La.) the State Department of Environmental Quality (Water). I VI'DES JWrnICl 6.3.7 Seventy-two Hour Notifications If a Notification of Unusual Event is declared:

n. The Station Coordinator Emergency Preparedness shall prepare a Summary Report from information in completed Emergency Plan Implementing Procedures. Cnntrol Room logs. and interviews with persons involved with the declaration and response, as appropriate. See Attachment 8, Example DEM Summary Report

b. The Site Vice President, Director Nuclear Station Safety and Licensing, or Plant Manager (Nuclear) shall approve the report.

c. Nuclear Emergency Preparedness shall ensure tlle report is delivered to the State Coordinator of DEM within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the dec\arntion. INA>:P"4; SEP4.41 I. Unusual orextrnordinary discharge includes, but is not limited to: a) wlplnnned bypasses, b) upsets. c) spillage of materials resulting directly or indirectly from proces:o;ing opemtions or pollutant management activities.

d) bre<1kdown of proce~'ing or accessory equipment, e) failure of 0(' taking out of service. sewage or industrial waste treatment facilities. auxiliary facilities. or pollutant manD,gement activities, orf) flooding orotber acts of nature. [vroES Pennlt]

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,j OOOWIE 13 EK2.2 I 1.00 MCS A 2 0012 A4.07 I 1.00 MCS B 3 0013 K2.01 2 1.00 MCS A 4 0022 KI.04 2 1.00 MCS B 5 000008AAZ.20 2 1.00 MCS D 6 000009EK2.03 3 1.00 MCS A 7 000011EK3.125 1.00 MCS A 8 000015AAZ.092 1.00 MCS C 9 oooo22AG2.4.11 2 1.00 MCS C 10 oooo25AK1.01 3 1.00 MCS A 11 000027AK3.03 2 1.00 MCS C 12 oooo29EA1.02 2 1.00 MCS D 13 oooo38EA1.21 2 1.00 MCS B 14 000040AA2.05 I 1.00 MCS D 15 000055EK1.02 I 1.00 MCS D 16 000056AA1.293 1.00 MCS A 17 000057AAZ.20 2 1.00 MCS B 18 000058AG2.1.23 3 1.00 MCS D 19 00005AAZ.OI 2 1.00 MCS A 20 0OOO65AA1.033 1.00 MCS A 21 000077AAZ.07 I 1.00 MCS C 22 OOOOWIE04 EA2.1 2 1.00 MCS B 23 OOOOW1E05 EAl.l 2 1.00 MCS A 24 0003A3.04 3 1.00 MCS A 25 0003A4.083 1.00 MCS B 26 0004A4.083 1.00 MCS B 27 0005 G2.4.9 3 1.00 MCS D 28 00068 AG2.4.42 2 1.00 MCS D 29 0006K1.052 1.00 MCS C 30 0007A1.03 I 1.00 MCS B 31 0008 A2.04 2 1.00 MCS C 32 0008 G2.1.7 2 1.00 MCS A 33 OOOWlE07 EK2.12 1.00 MCS A 34 OOOWlE08 EK1.3 3 1.00 MCS A 35 OOOWlE16 EAl.l 2 1.00 MCS D 36 0010 A3.02 4 1.00 MCS A 37 0012 K1.05 2 1.00 MCS A 38 0026 K3.02 3 1.00 MCS A 39 0036AAZ.OI 2 1.00 MCS C 40 0037AG2.4.4 2 1.00 MCS A 41 0039 A1.05 I 1.00 MCS C 42 0039 K4.04 2 1.00 MCS A 43 0059 K4.16 3 1.00 MCS B 44 0060AK3.022 1.00 MCS A 45 0061 KS.02 2 1.00 MCS D 46 0061 K6.01 2 1.00 MCS A Tuesday, July 28, 2009 12:55:45 PM 1

ANSWER KEY REPORT for RO Portion of Exam Test Form: 0 Answers It', ,~ ':; "',- ;"__ /"n,

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47 0062 K2.01 I 1.00 MCS A 48 0062 K3.03 I 1.00 MCS D 49 0063 K3.02 2 1.00 MCS B 50 0064 K4.1O 2 1.00 MCS C 51 0073 K5.02 I 1.00 MCS B 52 0076 K4.02 3 1.00 MCS A 53 01 K2.05 2 1.00 MCS B 54 0103 AI.OI I 1.00 MCS B 55 0103 K4.061 1.00 MCS C 56 011 K6.06 2 1.00 MCS A 57 014A4.012 1.00 MCS B 58 015K5.1O 3 1.00 MCS D 59 017K5.022 1.00 MCS B 60 035A2.012 1.00 MCS C 61 041 A3.03 2 1.00 MCS B 62 055 G2.4.45 2 1.00 MCS B 63 072 K4.011 1.00 MCS A 64 078 K3.03 2 1.00 MCS C 65 086 AI.05 2 1.00 MCS B 66 G2.1.29 I 1.00 MCS A 67 02.1.40 I 1.00 MCS A 68 G2.1.44 2 1.00 MCS A 69 G2.2.40 3 1.00 MCS B 70 G2.2.422 1.00 MCS C 71 G2.3.5 2 1.00 MCS B 72 02.3.72 1.00 MCS C 73 G2.4.142 1.00 MCS C 74 G2.4.182 1.00 MCS A 75 02.4.493 1.00 MCS C SECTION 1 ( 75 items) 75.00 Tuesday, July 28, 2009 12:55:45 PM 2