Text

Initial Written Retake Examination, 2009-301 Final SRO Written Exam
	ML090690067
Person / Time
Site:	North Anna
Issue date:	02/10/2009
From:	; NRC/RGN-II
To:	; Virginia Electric & Power Co (VEPCO)
References
	50-338/09-301, 50-339/09-301, ES-401, ES-401-8 50-338/09-301, 50-339/09-301
	Download: ML090690067 (121)
	v • d • e

ES-401 Name:

Site-Specific 'SRO Written Examination Cover Sheet u.s. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Form ES...401...8 Date:

February 10, 2009 Region:

I D II ~ III D IV D Start Time:

Facility/Unit:

North Anna Units 1 &2 Reactor Type: W ~ CE D BW D GE D 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 theRO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete the combined examination, and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months if you are only taking the SRO portion.

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

Applicant's Signature Results RO/SRO-Only/Total Examination Values Applicant's Scores Applicant's Grade 75

/ ~ / ~

Points

__ / __ / __

Points

__ / __ / __

Percent

NAPS 2009 NRC SRO Re-Take Exam

1. Given the following conditions:

RCS TAVE is 572°F.

"0" bank rods are at 190 steps.

Rod control is in AUTOMATIC.

Median/Hi TAVE input to automatic rod control fails LOW.

Which ONE of the following identifies the Immediate Operator Action required, and the effect of the transient on Shutdown Margin?

A.

Verify redundant instrument channel indication NORMAL; Shutdown Margin has increased.

B.

Verify redundant instrument channel indication NORMAL; Shutdown Margin has not changed.

C.

Place Rods in MANUAL; Shutdown Margin has increased.

D.

Place Rods in MANUAL; Shutdown Margin has not changed.

NAPS 2009 NRC SRO Re-Take Exam

2. Given the following conditions:

Unit 1 is at 25% power and slowly ramping up following a refueling outage The "B" Condens~te Pump and "B" Main Feedwater Pump are both tagged out Power is lost to the "B" Station Service Bus As a result of this transient, the GATe should initially expect to see level in the "B" SG and level in the "A" and "C" SGs (Assume no operator action)

A.

increase; increase B.

increase; decrease C.

decrease; increase D.

decrease ; decrease

NAPS 2009 NRC SRO Re~Take Exam

3. With Unit 1 at 100% power, annunciator A-E5, RCP 1A VIBRATION ALERT/DANGER,. is received.

The Backboards Operator reports the following:

"A" RCP Seismic vibration = 7 mils "A" RCP Proximity vibration = 12 mils Which ONE of the following identifies the status of "A" Rep vibration, and the actions required by AR-A-E5, RCP 1A VIBRATION ALERT/DANGER?

A.

Seismic vibration is above the ALERT level; Increase frequency of monitoring RCP vibration and consider shutting down "A" RCP if increasing trend continues.

B.

Seismic vibration is above the DANGER level; Trip the reactor and perform the immediate operator actions of 1-E-0, Reactor Trip or Safety Injection, then stop "A" RCP.

C.

Proximity vibration is above the ALERT level; Increase frequency of monitoring RCP vibration and consider shutting down "A" RCP if increasing trend continues.

D.

Proximity vibration is above the DANGER level; Trip the reactor and perform the immediate operator actions of 1-E-0, Reactor Trip or Safety Injection, then stop "A" RCP.

NAPS 2009 NRC SRO Re-Take Exam

4. As the ReS is cooled down from Hot Standby to Cold Shutdown, RCP seal injection flows will

________, and running RCP motor current will A.

decrease; increase B.

increase; increase C.

increase; remain the same D.

decrease; remain the same

NAPS 2009 NRC SRO Re-Take Exam

5. Given the following conditions:

The OATC has, established a 65 gp~ dilution in accordance with 1-GOP-8.3.1, Placing the Blender in the Dilute Mode of Operation The operator has adjusted the flow rate to the desired value, and matched the setpoint on the controller Shortly thereafter, the demand on controller 1-CH-FC-1114A, Primary Water to Blender Flow Controller, fails to 100%

Which ONE of the following identifies the response of the Blender?

A.

Annunciator B-D7, PG WATER TO BLENDER FLOW DEV, alarms as soon as the failure occurs; 1-CH-FCV-1114A, PG to Blender Flow Control Valve, automatically closes.

B.

Annunciator B-D7, PG WATER TO BLENDER FLOW DEV, alarms as soon as the failure occurs; 1-CH-FCV-1114B, Blender Makeup to VCT, automatically closes.

C.

Annunciator B-D7, PG WATER' TO BLENDER FLOW DEV, alarms approximately 20 seconds after the failure occurs; 1-CH-FCV-1114A, PG to Blender Flow Control Valve, automatically closes.

D.

Annunciator B-D7, PG WATER TO BLENDER FLOW DEV, alarms approximately 20 seconds after the failure occurs; 1-CH-FCV-1114B, Blender Makeup to VCT, automatically closes.

NAPS 2009 NRC SRO Re-Take Exam

6. Given the following conditions:

RHR is in-service with 1-RH-P-1A running and 1-CC-P-1 8 running 1-RH-P-18 and 1-CC-P-1A are available with their respective control switches in AFTER-STOP The normal supply breaker to 1H 4160-Volt 8us is inadvertently opened.

Which ONE of the following identifies the pump configuration after EDG loading is complete?

A.

ONLY 1-RH-P-1 8 running and ONLY 1-CC-P-1 8 running.

B.

ONLY 1-RH-P-18 running and 80TH 1-CC-P-1A & 1-CC-P-18 running.

c.

NO RHR pumps running and ONLY 1-CC-P-1 8 running.

D.

NO RHR pumps running and 80TH 1-CC-P-1A & 1-CC-P-1 8 running.

NAPS 2009 NRC SRO Re-Take Exam

7. Given the following conditions:

Rod H-2, Control Bank.0 group 1 is mis-aligned LOW Operators are preparing to realign the rod in accordance with 1-AP-1.3, Control Rod Out of Alignment 1-AP-1.3 will direct the crew to

as a result of this action, the crew should expect a alarm when Rod H-2 is withdrawn.

A.

Open the lift coil disconnect switches for Control Bank 0, Gp 1 rods ONLY, EXCEPT for Rod H-2; Rod Control Non-Urgent Failure.

B.

Open the lift coil disconnect switches for Control Bank 0, Gp 1 rods ONLY, EXCEPT for Rod H-2; Rod Control Urgent Failure.

C.

Open the lift coil disconnect switches for ALL Control Bank 0 rods, EXCEPT for Rod H-2; Rod Control Non-Urgent Failure.

D.

Open the lift coil disconnect switches for ALL Control Bank 0 rods, EXCEPT for Rod H-2; Rod Control Urgent Failure.

NAPS 2009 NRC SRO Re-Take Exam

8. Unit 1 is in Mode 6.

Technical Specifications require ONE RHR loop operable and in operation whenever water level is

~~~~~~~~~;

co~pliancewnh this LCO ensures~~~~~~~~~~

A.

greater than or equal to 23 feet above the top of the reactor vessel flange; proper ~ixing of RCS coolant to ~ini~ize the possibility of localized dilution.

B.

greater than or equal to 23 feet above the top of the reactor vessel flange; acceptable li~its are.~aintained in the event of a fuel handling accident.

C.

less than 23 feet above the top of the reactor vessel flange; proper ~ixing of RCS coolant to ~ini~ize the possibility of localized dilution.

D.

less than 23 feet above the top of the reactor vessel flange; acceptable li~its are ~aintained in the event of a fuel handling accident.

NAPS 2,009 NRC SRO Re-Take Exam

9. Operators are responding to a LOCA and*are performing Step 12 of 1-E-1, Loss of Reactor or Secondary Coolant, which directs them to "Check if Low-Head SI Pumps Should be Stopped."

The following plant conditions exist:

One charging pump is running and no others are available to be started RCS pressure is 3S0 psig and stable Containment pressure is 17 psia and slowly decreasing RWST level 'is S8.°k and slowly decreasing Based on these plant conditions, which ONE of the following identifies the action required by 1-E-1, and the basis for the action?

A.

Leave Low-Head SI pumps running since they are providing core cooling.

B.

Leave Low-Head SI pumps running to prepare for the transfer to Cold Leg Recirc.

C.

Stop Low-Head SI pLimps to minimize the potential for pump and motor overheating.

D.

Stop Low-Head SI pumps to minimize the potential for contaminating the RWST.

NAPS 2009 NRC SRO Re-Take Exam

10. During 100% power operation, PRT in-leakage is identified as increasing.

Which ONE of the following identifies a possible source?

A.

RCP #2 sealleakoff.

B.

RCP seal return relief valve leakage.

C.

RCS loop stop valve ~tem leakoff.

D.

Reactor vessel head vent valve leakage.

NAPS 2009*NRC SRO Re-Take Exam

11. Unit 1 is at 100% power when the following events occur:

0830 An instrument malfunction results in a PORV opening. The OATC attempts to close the affected PORV, but it remains full open. The OATC attempts to close the associated PORV block valve, but the PORV block valve breaker trips as soon as the switch is taken to CLOSE. The OATC manually trips the Unit and the crew enters 1-E-O, Reactor trip or Safety Injection.

0845 The Shift Manager classifies the event in accordance with EPIP-1.01, Emergency Manager Controlling Procedure.

Based on reporting requirements, the latest time that state and local authorities must be notified is

____, and the latest time that the NRC must be notified is A.

0900 ; 0945 B.

0900 ; 1000 C.

0930 ; 0930 D.

0930 ; 1030

NAPS 2009 NRC SRO Re-Take Exam

12. One of the purposes of the CC System is to cool the unit from to 1400 F within assuming a Service Water Temperature of 950 F.

A.

350oF; 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months B.

3500 F ; 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C.

547oF; 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months D.

5470 F ; 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

NAPS 2009 NRC SRO Re-Take Exam

13. Unit 1 was operating at 100% power when a LOCA occurred inside containment.

The following plant conditions exist:

The crew is currently implementing 1-E-0, Reactor Trip or Safety Injection Containment pressure is 21 psia and stable Containment pressure peaked at 25 psia Core Exit TCs indicate 510°F RCS pressure is 1100 psig Total high-head safety injection flow to the core is 500 gpm Based on these plant conditions, the Reactor Coolant Pumps should A.

not be secured B.

be secured because RCS pressure is less than 1275 psig C.

be secured due to low subcooling D.

be secured due to loss of component cooling flow

NAPS 2009 NRC SRO Re-Take Exam

14. Operators are stabilizing the plant following a reactor trip due to a loss of offsite power.

The following plant conditions exist:

RCS temperature is 551°F and stable.

PRZR pressure is 2100 psig and stable with one group of heaters energized.

PRZR liquid temperature is 600°F and stable.

Charging flow is 40 gpm and constant with 1-CH-FCV-1122 in MANUAL.

Seal injection flows are 8 gpm each.

Sealleakoff flows are 3 gpm each.

PRZR level increased to 35%

the crew has just placed 1-CH-HCV-1200B in service and letdown flow indicates 73 gpm.

Which ONE of the following describes the plant response?

A.

PRZR level will remain stable; PRZR pressure will remain stable B.

PRZR level will decrease; PRZR pressure will remain stable C.

PRZR level will remain stable; PRZR pressure will decrease D.

PRZR level will decrease; PRZR pressure will decrease

NAPS 2009 NRC SRO Re-Take Exam

15. The ensures that the allowable heat generation rate (kw/ft) is not exceeded; the setpoint for this trip is A.

Overpower LlT trip; 107.9°Jb B.

Overpower Ll"T trip; 126.4°Jb C.

Power Range Neutron Flux trip - High SIP; 109%

D.

Power Range Neutron Flux trip - High SIP; 110%

NApoS 200goNRC SRO Re-Take Exam

16. Given the following conditions:

Unit 1 is at 100% power Containment Pressure Protection Channel II (1-LM-P-1 OOB) failed on the previous shift All bistables for the failed channel have been placed in TEST in accordance with 1-MOP-55.75, Containment Pressure Protection Instrument Subsequently, Containment Pressure Protection Channel III (1-LM-P-100C) fails HIGH.

Which ONE of the following identifies the response of the Engineered Safeguards Features Actuation System?

A.

No automatic actuations.

os.

Safety Injection actuation ONLY.

c.

Safety Injection actuation and Main Steam Line Isolation ONLY.

D.

Safety Injection actuation, Main Steam Line Isolation, and CDA.

NAPS 2009 NRC SRO Re-Take Exam

17. Given the following conditions:

A spurious turbine runback occurs and is terminated at 80% power Operators are stabilizing the unit, a'nd annunciator A-H7, A.F.D MONITOR, has just alarmed Which ONE of the following identifies the MINIMUM number of NI channels with an AFD that exceeds the allowable limits of the Reactor Data book to consider the alarm valid, and the Technical Specification required action?

A.

1 NI channel; reduce power to < 50°J'o within 30 minutes.

B.

1 NI channel; reduce power to < 75% within 30 minutes.

C.

2 NI channels; reduce power to < 50% within 30 minutes.

D.

2 NI channels; reduce power to < 75°J'o within 30 minutes.

NAPS 2009 NRC SRO Re-Take Exam

18. Which ONE of the following responses is NOT an indication that the #2 seal has failed on a Reactor Coolant Pump?

A.

Increasing seal injection flow.

B.

Increased level in the' Primary Drains Transfer Tank.

C.

1 sealleakoff is lower than normal.

D.

RCP Standpipe Hi level alarm lit.

NAPS 2009 N*RC SRO Re-Take Exam

19. Given the following conditions:

Unit 1 was initially at 100% power The crew started ramping the unit 10 minutes ago to perform a Turbine Valve Freedom Test TAVE is presently 3 degrees higher than TREF 1-MS-PT-1446, Channel III First Stage Pressure, is selected for control Based on these plant conditions, which ONE of the following identifies how the steam dumps will respond if 1-MS-PT-1447, Channel IV First Stage Pressure fails LOW?

A.

Steam dumps arm but remain closed because there is insufficient demand signal.

B.

Steam dumps arm and ONLY baf}ks 1 & 2 modulate open until TAVE matches TREF.

C.

Steam dumps DO NOT arm and are disabled since the TREF input is failed LOW.

D.

Steam dumps arm and all valves modulate open until TAVE matches TREF.

NAPS 2009 NRC SRO Re-Take Exam

20. 10 CFR 50.46, ECCS Acceptance Criteria, states that the ECCS is designed to ensure peak cladding temperature is maintained less than or equal to during a Large Break LOCA; The CETC temperature that is used to monitor the Core Cooling CSF Status tree is A.

1200°F; the average temperature of ALL GETCs in that tCCM train.

B.

1200°F; the average temperature of the five highest CETCs in that ICCM train.

c.

2200°F; the average temperature of ALL CETCs in that tCCM train.

D.

2200°F; the average temperature of the five highest CETCs in that tCCM train.

NAPS 2009 NRC SRO Re-Take Exam

21. 1-ECA-O.O, Loss of All AC Power, Attachment 4 (Attempting to Restore Power to 1H [1J] Emergency Bus) directs the crew to place all charging pumps in Pull-To-Lock if RCP temperature exceeds 235°F; this action is taken to A.

Seal Water Outlet; prevent damage to th.e RCP seal package B.

Seal Water Outlet; prevent damage to the thermal barrier heat exchanger C.

Pump Radial Bearing; prevent damage to the RCP seal package D.

Pump Radial "Bearing; prevent damage to th.e thermal barrier heat exchanger

NAPS 2009 NRC SRO Re-Take Exam

22. Given the following conditions:

The crew has tripped Unit 1 and initiated Safety Injection due to a loss of RCS inventory Conditions continue to degrade and COA automatically actuates Based on the above conditions, which ONE of the following identifies the automatic response of the Containment Cooling System?

A.

CARFs tripped when Safety Injection was initiated; Cooling water to CARFs isolated when Safety Injection was initiated.

B.

CARFs tripped when COA actuated; Cooling water to CARFs isolated when Safety Injection was initiated.

C.

CARFs tripped when Safety Injection was initiated; Cooling water to CARFs isolated when COA actuated.

D.

CARFs tripped when COA actuated; Cooling water to CAR.Fs isolated when COA actuated.

NAPS 2009 NRC SRO Re-Take Exam

23. Given the following conditions:

Following a reactor trip, Unit 1 was stabilized with all parameters on program and stable The crew initiated emergency boration in accordance with 1-ES-0.1, Reactor Trip Response With 1-CH-P-1 B initially in service, a loss of off-site power occurs.

Following the loss of offsite power. the CATC notes the following:

1~CH-P-1A breaker closed with 75 amps indicated 1-CH-P-1 B breaker closed with 10 amps indicated 1-CH-P-1C secured Annunciator C-G6, RCP 1A-B-C LABYTH SEAL LO FLOW, is lit Annunciator C-C5, CHG PP TO REGEN HX HI-LO FLOW, is lit Annunciator C-A7, CHG PP 1C 15H7 LOCKOUT, is lit PRZR level 20% and slowly decreasing VCT level 44% and slowly increasing

'Which ONE of the following identifies the cause of these conditions, and the procedure that should be implemented to correct these conditions?

A.

Discharge check valve failure on 1-CH-P-1 B; 1-AP-49, Loss of Normal Charging.

B.

Charging line rupture upstream of 1-CH-FCV-1122; 1-AP-16, Excessive RCS Leakage.

C.

Alarms are consistent with the loss of offsite power; 0-AP-10, Loss of Electrical Power.

D.

Rupture of RCP seal injection line upstream of 1-CH-HCV-1186; 1-AP-33.2, Loss of RCP Seal Cooling.

NAPS 2.009 NRC SRO Re-Take Exam

24. Following successful'completion of maintenance, Unit 1 is exiting an outage that began 4 days ago.

Given the following conditions:

The RCS is solid RCS temperature is 256°F and stable RCS pressure is 300 psig and stable RH R is in service A loss of ALL instrument air (inside and outside containment) occurs, and the crew entered 1-AP-28,

'Loss of Instrument Air, but was unable to restore a source of air.

Which ONE of the following identifies the ReS temperature response, and the action required by 1-AP-28 assuming the crew is unable to restore IA pressure?

A.

RCS temperature will increase; open RHR H/X CC Outlet Isolation Valves 1-CC-TV-103A & -1038 locally.

B.

RCS temperature will decrease; open RHR H/X CC Outlet Isolation Valves 1-CC-TV-103A & -1038 locally.

C.

RCS temperature will increase; adjust RHR Heat Exchanger Return Valves 1-CC-MOV-100A &-1008.

D.

RCS temperature will decrease; adjust RHR Heat Exchanger-Return Valves 1-CC-MOV-100A & -1008.

NAPS 2009 NRC SRO Re-Take Exam

25. Unit 1 is at 100% power and a loss of ALL CC flow has occurred.

In accordance with 1-AP-15, Loss of Component Cooling, the reactor shall be tripped and affected RCPs stopped if RCP rises to 195°F.

A.

Motor Bearing temperature B.

Pump Radial Bearing temperature C.

Stator Winding temperature D.

Seal Water Outlet temperature

NAPS 2009 NRC SRO Re-Take Exam

26. Given the following conditions:

A large-break LOCA occurred with Unit 1 initially at 100% power A loss of offsite power occurs All equipment functioned as designed EXCEPT that 1J1, 480-Volt Emergency Bus in Rod Drive, deenergized due to an electrical fault Based on these plant conditions, which ONE of the following identifies the Quench Spray and Recirc Spray pumps that have power available?

A.

BOTH Quench Spray pumps and ONLY three Recirc Spray pumps.

B.

ONLY one Quench Spray pump and ONLY three Recirc Spray pumps.

C.

BOTH Quench Spray pumps and ONLY two Recirc Spray pumps.

D.

ONLY one Quench Spray pump and ONLY two Recirc Spray pumps.

NAPS 2009 NRC SRO Re-Take Exam

27. Given the following conditions:

A large-break LOCA occurred with Unit 1 initially at 100% power All equipment functions as designed EXCEPT that 1-QS-P-1A, "A" Quench Spray Pump, shaft shears on startup Based on this malfunction, which ONE of the following describes the impact on the Quench Spray System, and on the Recirculation Spray System?

A.

ONLY ONE QS Spray Ring will receive flow; ONLY ONE Inside Recirc Spray Pump sump will receive design flow from Quench Spray.

B.

ONLY ONE QS Spray Ring will receive flow; BOTH Inside Recirc Spray Pump sumps will receive design flow from Quench Spray.

C.

BOTH QS Spray Rings will receive flow; ONLY ONE Inside Recirc Spray Pump sump will receive design flow from Quench Spray.

D.

BOTH QS Spray Rings will receive flow; BOTH Inside Recirc Spray Pump sumps will receive design flow from Quench Spray.

NAPS 2009 NRC SRO Re-Take Exam

28. Which ONE of the following describes the automatic operation of 1-QS-MOV-102A, Chemical Addition Tank outlet valve?

A.

1-QS-MOV-102A will.open immediately provided BOTH the "A" Quench Spray Pump breaker is closed ANO COA is actuated.

B.

1-QS-MOV-102A will open immediately provided EITHER the "A" Quench Spray Pump breaker is closed OR COA is actuated.

C.

1-QS-MOV-102A will open after as-minute time delay provided BOTH the "A" Quench Spray Pump breaker is closed ANO COA is actuated.

D.

1-QS-MOV-102A will open after a S-minute time delay provided EITHER the "A" Quench Spray Pump breaker is closed OR COA is actuated.

NAPS 2009 NRC SRO Re-Take Exam

29. Unit 1 is at 100% power.

The OATC notes that all PRZR heaters are energized, and RCS pressure is 2270 psig and rising.

Based on these plant conditions, which ONE of the following identifies the failed channel and includes the plant response if no operator action is taken?

A.

1-RC-PT-1444 failed low; 1-RC-PCV-1455C will cycle open and closed B.

1-RC-PT-1444 failed low; 1-RC-PCV-1456 will cycle open and closed C.

1-RC-PT-1445 failed low; 1-RC-PCV-1456 will cycle open and closed D.

1-RC-PT-1445 failed low; 1-RC-PCV-1455C will cycle open and closed

NAPS 2009 NRC SRO Re-Take Exam

30. Given the following conditions:

Unit 1 is at 100% power The Pressurizer Level Channel Defeat Switch is selected to position 461/460 Annunciator B-G7, PRZ LO LEV HTRS OFF LETDOWN ISOL, alarms The OATC notes the following:

PRZR level is 66% and slowly increasing Demand on 1-RC-LC-1459G, PRZR Level Controller, is approximately 35%

and slowly decreasing Based on these plant conditions, which ONE of the following identifies the failed instrument, and the Immediate Operator Action required by 1-AP-3, Loss of Vital Instrumentation?

A.

1-RC-LI-1460 is failed low; Place 1-CH-FCV-1122, Charging Flow Control valve, in MANUAL and control level at program.

B.

1-RC-LI-1460 is failed low; Place Pressurizer Level Channel Defeat Switch in position 459/461, then verify Annunciator B-G7, PRZ LO LEV HTRS OFF LETDOWN ISOL, clears.

C.

1-RC-LI-1461 is failed low; Place 1-CH-FCV-1122, Charging Flow Control valve, in MANUAL and control level at program.

D.

1-RC-LI-1461 is failed low; Place Pressurizer Level Channel Defeat Switch in pos*ition 459/460, then verify Annunciator B-G7, PRZ LO LEV HTRS OFF LETDOWN ISOL, clears.

NAPS 2009 NRC SRO Re-Take Exam

31. Unit 1 is in Mode 6 with core off-load in progress.

A containment purge exhaust fan is started by the Backboards Operator. As a result of this action, reactor cavity level will and spent fuel pit level will A.

increase slightly; decrease slightly B.

increase slightly; not be affected C.

not be affected; not be affected D.

decrease slightly; increase slightly

NAPS 2009 NRC SRO Re~Take Exam

32. Which ONE of the following describes the actions required by 2-FR-S.1, Response to Nuclear Power Generation/ATWS, Attachment 4 (Remote Reactor Trip) if attempts to locally trip the reactor from the Rod Drive room are unsuccessful?

A.

De-energize BOTH 2A1 and 2C2 480-Volt Station Service Busses by opening their respective supply breakers.

B.

De-energize BOTH 2A1 and 2B2 480-Volt Station Service Busses by opening their respective supply breakers.

c.

Open BOTH Rod Drive M-G Set Motor Supply Breakers locally at the 2A1 and 2C2 480-Volt Station Service Busses.

D.

Open BOTH Rod Drive M-G Set Motor Supply Breakers locally at the 2A1 and 2B2 480-Volt Station Service Busses.

NAPS 2009 NRC SRO Re-Take Exam

33. Unit 1 tripped from 1000/0 power 22 minutes ago.

The following plant conditions exist:

Intermediate Range Channel N-35 indicates 8 x 10-10 amps and stable.

Intermediate Range Channel N-36 indicates 1 x 10-11 amps and stable.

High voltage is de-energized on BOTH Source Range Channels N-31 & N-32.

1-ES-0.1, Reactor Trip Response, is in effect and the crew is at the step, "Check If Source Range Detectors Should Be Energized.".

Based on these plant conditions, which ONE of the following identifies the malfunction that has occurred, and the action required by 1-ES-0.1?

A.

N-35 is under-compensated; Manually energize Source Range Detectors using the Source Range Block and Reset switches.

B.

N-35 is over-compensated; Manually energize Source Range Detectors using the Source Range Block and Reset switches.

C.

N-35 is over-compensated; Energize Source Range Detectors by bypassing N-35 and removing the instrument power fuses.

D.

N-35 is under-compensated; Energize Source Range Detectors by bypassing N-35 and removing the instrument power fuses.

NAPS 2009 NRC SRO Re-Take Exam

34. Core on-load has commenced and Containment Purge is in service on Unit 1.

1-RM-RMS-159, Containment Particulate Radiation Monitor, has just pegged high due to a malfunction of the monitor.

Which ONE of the following identifies the automatic actuations that will occur, and the impacts of this failure on fuel movement lAW 1-0P-4.1, Controlling Procedure for Refueling?

A.

ONLY Containment Purge Isolation; OPS Manager approval is required to resume fuel movement.

B.

Containment Purge Isolation AND Control Room Bottle Air Dump; OPS Manager approval is required to resume fuel movement.

C.

ONLY Containment Purge Isolation; Fuel movement may continue without any additional approvals.

D.

Containment Purge Isolation AND Control Room Bottle Air Dump; Fuel movement may continue without any "additional approvals.

NAPS 2009 NRC SRO Re-Take Exam

35. The SG PORV controllers (1-MS-PCV-1 01A, B, & C) are designed to control RCS TAVE at approximately

___ ; this is achieved by a controller pot setting of approximately A.

551°F; 5.5 B.

551°F; 7.0

NAPS 2009 NRC SRO Re-Take Exam

36. Given the following conditions:

Unit 1 generator output breaker G-12 has just been closed following a reactor startup All control systems are aligned per startup procedures "An train of steam dumps is isolated to repair 1-MS-TCV-1408A, which is de-energized and tagged out Main steam line pressure transmitter 1-MS-PT-1464 fails HIGH Which ONE of the following identifies the ReS temperature response, and the required action?

A.

RCS temperature increases; Manually insert rods to match TAVE and TREF' B.

RCS temperature increases; Verify rods automatically insert to match TAVE and TREF' C.

RCS temperature decreases; Close all MSTVs using Appendix R Switch.

D.

RCS temperature decreases; Place both Steam Dump Interlock Switches to OFF/RESET.

NAPS 2009.NRC SRO Re-Take Exam

37. Given the following conditions:

Unit 1 was operating at 100% power when a Main Steamline break occurred inside containment 1-E-2, Faulted Steam Generator isolation, has been completed The team transitioned to 1-ES-1.1, SI Termination, and has just completed isolating the BIT and establishing normal charging The following plant conditions exist:

Containment pressure is 19 psia and decreasing Intact SG narrow-range levels are 8% and increasing PRZR level is 16% and decreasing 1-CH-FCV-1122, Charging flow control valve is full open Based on these plant conditions, which ONE of the following identifies the action required by 1-ES-1.1, SI Termination?

A.

Manually start charging pumps and align the BIT; go to 1-E-1, Loss of Reactor or Secondary Coolant.

B.

When Containment pressure is less than 13 psig, then stop Quench Spray Pumps; Continue in 1-ES-1.1.

C.

Isolate letdown; if PRZR level continues to decrease, then manually actuate SI.

D.

Stop feed flow to SGs; if cooldown continues, then close MSTVs and bypass valves.

NAPS2009"NRC SRO Re-Take Exam

38. Because power is lost to the steam dump arming circuit, Attachment 24, (Unit 1 CRO Loss of Power Actions), of O-AP-10, Loss of Electrical Power, directs the operator to "Check SG PORVs controlling in AUTO or MANUAL due to loss of Steam Dumps" in the event of a loss of A.

120-VAC Vital Bus I-I B.

120-VAC Vital Bus I-II c.

120-VAC Vital Bus I-III D.

120-VAC Vital Bus I-IV

NAPS 2009 NRC SRO Re-Take Exam

39. Given the following conditions:

Unit 1 is at 100% power when the crew notes degrading condenser vacuum The crew has entered 1-AP-14, Low Condenser Vacuum The Turbine Building operator has been dispatched to perform Attachment 2, Turbine Building Corrective Actions The Turbine Building operator checks both condenser air ejector loop seal drain lines to condenser, and notes that one loop seal drain line is hot to the touch and the other is cool to the touch.

Which ONE of the following.identifies the action required in accordance with 1-AP-14?

A.

Isolate the hot loop seal drain line and secure the associated set of main condenser air ejectors.

B.

Isolate the hot loop seal drain line; when at least 15 minutes have elapsed, then slowly reopen the loop seal drain isolation valve.

C.

Isolate the cool loop seal drain line and secure the associated set of main condenser air ejectors.

D.

Isolate the cool loop seal drain line; when at least 15 minutes have elapsed, then slowly reopen the loop seal drain isolation valve.*

NAPS 2009 NRC SRO Re-Take Exam

40. Given the following conditions:

Operators have transitioned to 1-FR-H.1, Response to Loss of Secondary Heat Sink.

SG wide-range levels are approximately 50 % and slowly decreasing.

Operators are at Step 2 of 1-FR-H.1 and are unable to establish AFW flow to any of the SGs.

Which ONE of the following identifies the action required by 1-FR-H.1 with respect to the Reps?

A.

Stop all but ONE RCP.

B.

Stop all RCPs.

C.

Maintain RCPs operating unless support conditions degrade.

D.

Maintain RCPs operating until bleed and feed criteria are met.

NAPS 2009 NRC SRO Re-Take Exam

41. Given the following conditions:

80th units are initially at 100°/6 power 2-SW-P-18 is tagged out, and Service Water is throttled A loss of both Unit-1 4160-Volt Emergency buses occurs, and power is eventually restored to the 1H Emergency bus from the S80 diesel.

Which ONE of the following identifies how the Unit-1 Service Water pumps are operated in accordance with 1-ECA-0.O, Loss of All AC Power?

A.

1-SW-P-1A and 1-SW-P-1 8 control switches are maintained in AUTO during 1-ECA-0.0 recovery actions; 1-SW-P-1A will be started prior to exiting 1-ECA-0.0.

B.

1-SW-P-1A and 1-SW-P-1 8 control switches are maintained in AUTO during recovery actions; 1-SW-P-1A will be started after exiting 1-ECA-0.0.

C.

1-SW-P-1A and 1-SW-P-1 8 control switches are placed in PULL-TO-LOCK during recovery actions; 1-SW-P-1A will be started prior to exiting 1-ECA-0.0.

D.

1-SW-P-1A and 1-SW-P-1 8 control switches are placed in PULL-TO-LOCK during recovery actions; 1-SW-P-1A will be started after exiting 1-ECA-0.0.

NAPS 2009 NRC SRO Re-Take Exam

42. Given the following conditions:

Unit 1 is at 25 % power ramping up following a scheduled refueling Annunciator A-G1, CNDSR La VAC C-9 PERM NOT AVAIL is received The OATC notes that condenser vacuum is 4 in Hg abs and slowly degrading The crew enters 1-AP-14, Low Condenser Vacuum Which ONE of the following identifies the action required to mitigate these plant conditions?

A.

Go to 1-AP-2.2, Fast Load Reduction, reduce load and remove the main generator from service while continuing with 1-AP-14.

B.

Exit 1-AP-14, go to 1-AP-2.2, Fast Load Reduction, reduce load and remove the main generator from service.

C.

Trip the reactor and go to 1-E-O, Reactor Trip or Safety Injection, while continuing with 1-AP-14.

D.

Exit 1-AP-14, trip the reactor and go to 1-E-O, Reactor Trip or Safety Injection.

NAPS 2009*NRC SRO Re-Take Exam

43. Given the following conditions:

Unit 1 is at 100% power and stable Following a lightening strike the OATC has identified that several instruments powered from Vital AC Bus 1-111 have been lost Which ONE of the following failed instruments requires action within 1 Hour in accordance with Technical Specifications?

A.

Pressurizer pressure protection channel III, 1-RC-PT-1457.

B.

Pressurizer level channel III, 1-RC-LT-1461.

C.

"B" SG steam flow channel III, 1-MS-FT-1484 D.

"A" SG narrow-range' level channel III, 1-FW-LT-1476

NAPS 2009 NRC SRO Re-Take Exam

44. Which ONE of the following identifies the power level and sequence for securing a Main Feedwater (MFW) pump in accordance with 1-AP-2.2, Fast Load Reduction?

At approximately 55% power,

at approximately 50%

power, A.

open one MFW recirc valve; shutdown one MFW pump.

B.

shutdown one MFW pump; open one MFW recirc valve.

C.

close one MFW recirc valve; shutdown one MFW pump.

D.

shutdown one MFW pump; close one MFW recirc valve.

NAPS 2009 NRC SRO Re-Take Exam

45. Given the following conditions:

Unit 1 is at 70 0/b power "8" Main Feedwater (MFW) pump is tagged out The "G" MFW pump trips.

Which ONE of the following describes the plant response?

(Assume no operator action.)

A.

MFRVs modulate open, but SG levels continue to decrease.

B.

MFRVs modulate open and maintain SG levels.

C.

MFRVs and bypass valves modulate open and maintain SG levels.

D.

MFRVs and bypass valves fully open, but SG levels continue to decrease.

NAPS 2009 NRC SRO Re-Take Exam

46. With AFW in service to supply SG feedwater, ECST level suddenly begins to decrease rapidly due to a rupture in the tank.

Which ONE of the following identifies the preferred source of alternate makeup water used to supply AFW pump suction?

A.

Fire Protection System using motor-driven pump 1-FP-P-1.

B.

Standby 300,OOO-gallon condensate storage tank.

C.

In-service 300,OOO-gallon condensate storage tank.

D.

Fire Protection System using diesel-driven pump 1-FP-P-2.

NAPS 2009 NRC SRO Re-Take Exam

47. Given the following conditions:

Unit 1 is at 100% power AFW pump 1-FW-P-3A is tagged out for Maintenance A loss of offsite power occurs, and all equipment functions as designed EXCEPT the 1J EDG locked out.

Based on these plant conditions, which ONE of the following identifies the realignment that will provide flow to all 3 SGs AND also allows the BOP the ability to control AFW flow rate to each SG from the control board?

A.

Align 1-FW-P-2 to the MOV header.

B.

Align 1-FW-P-2 to the HCV header.

C.

Align 1-FW-P-3B to.the MOV header.

D.

Align 1-FW-P-3B to the HCV header.

NAPS 2009 NRC SRO Re-Take Exam

48. The NORMAL source of cooling water for charging pumps is the

in the event of a loss of the norm*al source, alternate cooling is aligned from the A.

Service Water System; Component Cooling System B.

Service Water System; Fire Protection System C.

Component Cooling System; PG Water System D.

Component Cooling System; Fire Protection System

NAPS 2009 NRC SRO Re-Take Exam

49. Given the following conditions:

Unit 1 is at 50% power 15F3, "F" Transfer Bus Supply to 1H Bus, is tagged out for breaker maintenance The 1H 4160-Volt Emergency bus is powered from the 1B Station Service Bus While checking neutral bus volts, the GATe inadvertently takes breaker 15B2, Station Service Transformer Supply to 1B Station Service Bus, to TRIP.

As a result of this action, final status of the 1B Station Service Bus will be

, and the final status of the 1H*4160-Volt Emergency bus will be A.

de-energized; energized B.

energized; energized C.

energized; de-energized D.

de-energized; de-energized

NAPS 2009 NRC SRO Re-Take Exam

50. A Loss of All AC has just occurred on Unit 1.

Five minutes have elapsed, and a source *of power has not yet been restored.

The operator should expect the voltage on all four (4) of the 125-Volt DC Busses to be their pre-event values. Over time, vital DC bus 1-1 voltage would be expected to drop as compared to vital DC bus 1-11 voltage.

A.

approximately the same as ; at approximately the same rate B.

approximately the same as ; at a slower rate c.

approximately 10 volts lower than; at approximately the same rate D.

approximately 10 volts lower than ; at a slower rate

NAPS 2009 NRC SRO Re-Take Exam

51. Which ONE of the following methods ensures that only one DC bus at a time is supplied from the swing charger?

A.

Each swing charger supply breaker to the DC bus has a Key-operated interlock mechanism.

B.

The swing charger output breakers are interlocked such that both breakers trip if you attempt to close one with the other one already closed.

C.

Each swing charger has a singfe dedicated plug-in type output breaker.

D.

The swing charger output breakers are electrically interlocked to prevent simultaneous closure.

NAPS 2009 NRC SRO Re-Take Exam

52. When the emergency diesel generator air compressors are swapped to the Lister diesels, the Lister diesel

, and air bank pressure is maintained by the loading and unloading of the air compressor.

A.

will run continuously; automatic B.

will run continuously; manual C.

is manually started each time air pressure drops to 150 psig; automatic D.

is manually started each time air pressure drops to 150 psig; manual

NAPS 2009 NRC SRO Re-Take Exam

53. Unit 1 is at 100% power.

After receiving annunciator J-E8, INSTRUMENT AIR LOW PRESS, operators entered 1-AP-28, Loss of Instrument Air.

The BOP notes Instrument Air pressure has decreased to

. 1-AP-28 directs the crew to trip the reactor and close Main Steam Trip Valves. These actions are taken primarily to A.

68 psig ; prevent an undesired automatic Safety Injection.

B.

68 psig ; prevent an uncontrolled RCS cooldown.

C.

88 psig ; prevent an undesired automatic Safety Injection.

D.

88 psig ; prevent an uncontrolled RCS cooldown.

NAPS 2009 NRC SRO Re-Take Exam

54. If an oil fire occurs in the North Fuel Oil Pump House, a will actuate to suppress the fire; the fire is classified as a A.

pre-action sprinkler system; Class B fire B.

pre-action sprinkler system; Class A fire C.

high-pressure CO2 system; Class B fire D.

high-pressure CO2 system; Class A fire

NAPS 2009 NRC SRO Re-Take Exam

55. Which ONE of the following components is automatically affected by a high alarm on 1-GW-RI-178-1, MGP Process Vent Radiation Monitor?

A.

Containment Vacuum Pump Discharge Valves B.

Low Level Liquid Waste Tank Vents C.

Boron Recovery Tank Vents D.

Containment Purge Exhaust Valves

NAPS 2009 NRC SRO Re-Take Exam

56. Both units are at 100% power.

Breaker 15G1 0, 1G/2G bus crosstie, will automatically close if power is lost to the A.

"A" RSST provided lake level is greater than 251 feet B.

"A" RSST provided lake level is less than 251 feet C.

"C" RSST provided lake level is greater than 251 feet D.

"c" RSST provided lake level is less than 251 feet

NAPS 2009 NRC SRO Re-Take Exam

57. Given the following conditions:

Both units are at 100% power 1-SW-P-1A and 2-SW-P-1A are running Unit 1 experiences a large-break LOCA Service Water flow to the CCHXs isolates A.

on Unit 1 ONLY, and 3 Service Water pumps will be running.

B.

on BOTH units, and 3 Service Water pumps will be running.

C.

on Unit 1 ONLY, and 4 Service Water pumps will be running.

D.

on BOTH units, and 4 Service Water pumps will be running.

NAPS 2009 NRC SRO Re-Take Exam

58. Unit 1 is at 1000/0 power with approximately 960 MW and 100 MVAR OUT.

The OATe notes the following conditions:

Sustained Generator output voltage increase MVAR sustained increase consistent with Generator output voltage MW stable Which ONE of the following identifies the required operator actions based on these plant conditions?

A.

Maintain Voltage Regulator in AUTO and adjust voltage using Base Adjust; Maintain less than 200 MVAR OUT.

B.

Maintain Voltage Regulator in AUTO and adjust voltage using Base Adjust; Maintain less than 200 MVAR IN.

C.

Place Voltage Regulator control sWitch in OFF and adjust voltage using Base Adjust; Maintain less than 200 MVAR OUT.

D.

Place Voltage Regulator control switch in OFF and adjust voltage using Base Adjust; Maintain less than 200 MVAR IN.

NAPS 2009 NRC SRO Re-Take Exam

59. The Service Air Compressors are powered from their respective unit's and the Instrument Air Compressors are powered from their repective unit's A.

"B" 480-Volt Station Service Bus; "H" 480-Volt Emergency Bus B.

"B" 480-Volt Station Service Bus ;. "'J" 480-Volt Emergency Bus c.

"G" 480v Station Service Bus; "H" 480-Volt Emergency Bus D.

"G" 480-Volt Station Service Bus; "J" 480-Volt Emergency Bus

NAPS 2009 N-RC SRO Re-Take Exam

60. If Chilled Water to Containment is lost, the indicated containment partial air pressure will and the digital containment partial air pressure indicators are A.

decrease; inoperable B.

decrease; operable C.

increase; inoperable D.

increase; operable

NAPS 2009 N*RC SRO Re-Take Exam

61. Unit 1 has experienced a DBA LOCA.

The following plant conditions exist:

Containment pressure is 25 psia Containment sump level is 11 feet 4 inches Which ONE of the following identifies the status of the Containment CSF Status Tree?

A.

An ORANGE path exists for both Containment Pressure and Containment Sump level.

B.

An ORANGE path exists for Containment Pressure ONLY.

C.

An ORANGE path ex.ists for Containment Sump level ONLY.

D.

An ORANGE path does NOT exist for Containment Pressure or Containment Sump level.

NAPS 2009 NRC SRO Re-Take Exam

62. Both units are at 1000/0 power with no equipment out of service (except as noted below).

Which ONE of the following events would initiate an automatic start of the SBO diesel?

A.

With "e" RSST out of service, a loss of 34.5 KV bus #4 occurs.

B.

With "8" RSST out of service, a loss of 34.5 KV bus #4 occurs.

C.

Spurious trip of breaker 15F1, "F" transfer bus supply.

D.

Spurious trip of breaker 1501, "0" transfer bus supply.

NAPS 2009 NRC SRO Re-Take Exam

63. In order to maintain an active license, a Reactor Operator must stand a MINIMUM of __ 12-hour shifts per calendar quarter.

A.

4 B.

7 C.

5 D.

8

NAPS 2009 NRC SRO Re-Take Exam

64. Given the following conditions:

Unit 1 is at 100% power.

Annunciator B-H2, PRZ POWER RELIEF LINE HI TEMP is received During the next 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , PRT level increases from 72°/b to 76°/b RCS Leakage prior to the event was 0.5 gpm Identified, and 0.3 gpm Unidentified Based on the above conditions, which ONE of the following describes the Technical Specification implications of the leaking PORV?

(Reference provided)

A.

The limit for RCS Operational Leakage has NOT been exceeded.

B.

The limit for Identified Leakage has been exceeded.

C.

The limit for Pressure Boundary Leakage has been exceeded.

D.

The limit for RCS Pressure Isolation Valve (PIV) Leakage has been exceeded.

NAPS 2009 NRC SRO Re-Take Exam

65. The status of annunciator patch cords is checked on a basis using O-GOP-2.7, Annunciator Patchcord Control, and the is used to track the total number of disabled annunciators.

A.

quarterly; Central Reporting System B.

quarterly; Operations Aggregate Impact Report C.

weekly; Central Reporting System D.

weekly; Operations Aggregate Impact Report

NAPS 2009 NRC SRO Re-Take Exam

66. The SG N-16 Radiation Monitors are required by Technical Requirements in

, and can not be declared functional until reactor power is at least __'

A.

Modes 1, 2, and 3 ; 5"OJ'c, B.

Modes 1, 2, and 3 ; 25%

C.

Modes 1 &2 ; 5°J'c, D.

Modes 1 & 2 ; 250/0

NAPS 2009 NRC SRO Re-Take Exam

67. In accordance with VPAP-2101, Radiation Protection Program, a Radiation Worker's Annual Administrative Dose Limit for Total Effective Dose Equivalent (TEDE) is rem/calendar year at the worker's home site.

A.

B.

2 C.

3 D.

4

NAPS 2.009 NRC SRO Re-Take Exam

68. Unit 1 is at 100°A, power. '

1-SV-RM-121, Condenser Air Ejector Radiation Monitor, spikes high resulting in a Hi alarm AND a Hi-Hi alarm on the drawer, and then the indication returns to normal.

Based on the above, which ONE of the following describes the response of 1-SV-RM-121, Condenser Air Ejector Radiation Monitor?

A.

The Hi alarm clears \\lYhen the indication returns to normal; the Hi-Hi alarm remains in and must be reset by the operator.

B.

The Hi alarm clears when the indication returns to normal; the Hi-Hi alarm remains in and must be reset by I&C.

C.

The Hi-Hi alarm clears when the indication returns to normal; the Hi alarm remains in and must be reset by the operator.

D.

The Hi-Hi alarm clears when the indication returns to normal; the Hi alarm remains in and must be reset by I&C.

NAPS 2*009 NRC SRO Re-Take Exam

69. 1-FR-H.1, Response to Loss of Secondary Heat Sink, is in effect.

When attempting to establish SG feed flow from the Condensate System, there is a NOTE prior to the step "Initiate depressurization of all SGs to less than 610 psig by dumping steam to condenser at maximum rate."

This NOTE states that each Main Steamline flow should be kept less than 1.0E6 LBM/HR.

The purpose of this NOTE is to remind the operator that exceeding 1.0E6 LBM/HR may result in A.

aRCS cooldown rate in excess of Technical Specification limits.

B.

a challenge to the PTS status tree.

C.

an inad~ertant Steamline DP Safety Injection.

.0.

an undesired Main Steamline Isolation.

NAPS 2009 NRC SRO Re-Take Exam

70. Once the TSC is fully operational, the relief Station Emergency Manager will assume all of the following responsibilities from the on-shift SEM EXCEPT A.

authorizing notification to state and local governments.

B.

classifying the emergency (escalation/de-escalation).

C.

making protective action recommendations (PAR).

°0.

authorizing deviations to procedures pursuant to 10 CFR 50.54x.

NAPS 2009 NRC SRO Re-Take Exam

71. Operators are responding to a large-break LOCA with a loss of offsite power.

The team is implementing 1-E-1, Loss of Reactor or Secondary Coolant.

The OATC reports that the following alarms have just been received:

F-E8, AFW SUPPLY 20 MIN WATER REMAINING J-A2, RWST LO LEVEL H-H1, EMER DIESEL GEN #1H DIFFERENTL J-F5, LHSI PP 1B LO OR OL TRIP Based on these alarms, thecrewwill~~~~~~~~~~~~~~

A.

verify LHSI pump suction has transferred to containment sump and continue in 1-E-1, Loss of Reactor or Secondary Coolant.

B.

perform 1-AP-22.5, Loss of Emergency Condensate Storage Tank 1-CN-TK-1, in order to preclude implementing 1-FR-H.1, Loss of Secondary Heat Sink.

C.

transition to 1-ES-1.3, Transfer to Cold Leg Recirculation, then return to 1-E-1, Loss of Reactor or Secondary Coolant.

D.

transition to 1-ES-1.3, Transfer to Cold Leg Recirculation, then transition to 1-ECA-1.1, Loss of Emergency Coolant Recirculation;

NAPS 2009 NRC SRO Re~Take Exam

72. Operators have entered 1-ECA-1.2, LOCA Outside Containment.

1-ECA-1.2 will have the operator isolate cold leg injection piping from

, and if RCS pressure is INCREASING AFTER the isolation, the crew will transition to A.

High-Head Sf pumps; 1-E-1, Loss of Reactor or Secondary Coolant.

B.

High-Head Sf pumps; 1-ECA-1.1, Loss of Emergency Coolant Recirculation.

C.

Low-Head SI pumps; 1-ECA-1.1, Loss of Emergency Coolant Recirculation.

D.

Low-Head SI pumps; 1-E.;.1, Loss of Reactor or Secondary Coolant.

NAPS 2009 NRC SRO Re-Take Exam

73. Unit 1 tripped from 100% power, and a loss of secondary heat sink occurred.

The following plant conditions exist:

The crew is currently in 1-FR-H.1, Response to Loss of Secondary Heat Sink Bleed and feed has been established A Main Steamline break occurs inside containment and CDA actuates, but only one Quench Spray pump started RCS cold-leg temperatures have decreased to 26SoF Based on these conditions, which ONE of the following identifies the correct operator response?

A.

Reset CDA, reset Phase 'B', and place instrument air back in service to PORVs.

B.

Transition to 1-FR-Z.1, Response to High Containment Pressure.

C.

Transition to 1-FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

D.

Verify proper alignment of equipment using 1-E-0, Reactor Trip or Safety Injection.

NAPS 2009 NRC SRO Re-Take Exam

74. Unit 1 was operating at 1000/0 powerwhen a LOCA occurred.

The crew has just transitioned from 1-E-1, Loss of Reactor or Secondary Coolant, to 1-FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

The following plant conditions exist:

Containment pressure is 15 psia and is slowly trending down All RCPs are stopped RVLIS Full Range is reading 100%

RCS subcooling based on Core Exit TCs is 105°F RCS pressure is 850 psig Based on these plant conditions, which ONE of the following identifies the actions the crew will perform in accordance with 1-FR-P.1, and the reason for these actions?

A.

Remain in 1-FR-P.1, terminate SI and depressurize to minimize subcooling in order to limit pressure stress on the vessel.

B.

Exit 1-FR-P.1 and return to procedur~ step in effect, pressure is too low for PTS to be a concern.

C.

Exit 1-FR-P.1 and go*to 1-ES-1.1, SI Termination, SI must be terminated promptly to mitigate the PTS condition.

D.

Remain in 1-FR-P.1, leave SI in service and determine soak requirements in order to minimize thermal stresses during subsequent recovery actions.

NAPS 2009 NRC SRO' Re-Take Exam

75. Given the following conditions:

Unit 1 has experienced a.LOCA After establishing Cold Leg Recirculation, operators observed indication of sump blockage The crew has just transitioned to 1-ECA-1.1, Loss of Emergency Recirculation Capability The STA reports an ORANGE path on Containment Pressure.

Based on these plant conditions, which. ONE of the following identifies the crew response, and the reason for their response?

A.

Remain in 1-ECA-1.1; FRPs are not implemented until specifically directed by 1-ECA-1.1.

B.

Remain in 1-ECA-1.1; 1-FR-Z.1 is not to be performed as long as 1-ECA-1.1 is in effect.

C.

Immediately transition to 1-FR-Z.1 and operate Containment Depressurization Systems as directed, then return to 1-ECA-1.1; precludes a Red Path on the Containment status tree.

D.

Immediately transition to 1-FR-Z.1; Verify Phase A Isolation and Steamline Isolation, then return to 1-ECA-1.1; precludes a Red Path on the Containment status tree.

NAPS 2009 NRC SRO Re-Take Exam

76. Given the following conditions:

Unit 1 is at 100% power "

VCT level is initially 52%

VCT pressure is initially 30 psig The GATC notes 1-CH-LCV-1115A, VCT Level Control valve, is in the DIVERT position.

The following conditions exist:

VCT automatic makeup initiates 1-CH-LC-1112C, VCT Level Controller, indicates 100% demand VCT level is stable at 23%

VCT pressure is stable at 17 psig Based on the change in plant conditions, leakoff from the Rep number 2 seal to the standpipe will To mitigate this event, the crew will enter 1-AP-16, Increasing Primary Plant Leakage, and "A.

increase; isolate letdown by closing 1-CH-HCV-1200A, B, & C, and 1-CH-LCV-1460A & B.

B.

increase; align 1-CH-LCV-1115A to VCT by locally opening breaker 7 on 1-EP-CB-26B.

C.

decrease; isolate letdown by closing 1-CH-HCV-1200A, B, & C, and 1-CH-LCV-1460A & B.

D.

decrease; align 1-CH-LCV-1115A to VCT by locally opening breaker 7 on 1-EP-CB-26B.

NAPS 2009 NRC SRO Re-Take Exam

77. Unit 1 is in Mode 4 with a heatup in progress following a scheduled refueling outage.

The following plant conditions exist:

RHR is in service RCS temperature is 270°F SI accumulator pressures and levels are all at normal operating level and pressure SI accumulator MOVs are all CLOSED The Safeguards watchstander reports the following:

The breaker for "B" SI accumulator MOV is ON The breaker for "C" SI accumulator MOV is OFF Based on these plant conditions, which. ONE of the following identifies the action to be taken, and the reason for the action?

A.

Place the breaker for "C" SI accumulator MOV in ON; SI accumulators are maintained available for Shutdown LOCA mitigation concerns.

B.

Place the breaker for "C" SI accumulator MOV in ON; SI accumulator MOVs are energized to prepare for the change to Mode 3.

C.

Place the breaker for "B" SI accumulator MOV in OFF; Power is removed from the MOVs to preclude an inadvertent dilution of the RCS.

D.

Place the breaker for "B" SI accumulator MOV in OFF; Power is removed from the MOVs for RCS overpressure protection.

NAPS 2009 NRC SRO Re-Take Exam

78. Given the following:

Unit 1 is at 75% power, ramping up at O.3%/minute to return to full power The following events occur:

Annunciator A-D1, Rod Control Urgent failure is received The OATC reports that AUTO rod motion has stopped The ramp was held to investigate the alarm Local observation confirms the alarm is in the 2BD Power Cabinet Based on these plant conditions, which ONE of the following identifies the Technical Specification implications and includes the actions required by 1-AR-A-D1, Rod Control Urgent Failure?

A.

Control Rods are INOPERABLE; Operate Rods in MANUAL to maintain Tave and Tref matched until the problem is corrected.

B.

Control Rods are OPERABLE; Operate Rods in MANUAL to maintain Tave and Tref matched until the problem is corrected.

C.

Control Rods are INOPERABLE; Adjust Turbine load or RCS boron concentration to maintain Tave and Tref matched until the problem is corrected.

D.

Control Rods are OPERABLE; Adjust Turbine load or RCS boron concentration to maintain Tave and Tref matched until the problem is corrected.

NAPS 2009 N.RC SRO Re-Take Exam

79. Both units are at 100%

power.

The Technical Specification LCO for the Component Cooling Water (CC) System requires to be operable. The basis for this requirement is to ensure that A.

3 CC subsystems; one unit can be cooled down rapidly while the other unit is maintained at full power B.

3 CC subsystems; both units can be cooled down rapidly c.

4 CC subsystems; one unit can be cooled down rapidly while the other unit is maintained at full power D.

4 CC subsystems; both units can be cooled down rapidly

NAPS 2009 NRC SRO Re-Take Exam

80. Given the following conditions:

Unit 1 tripped from 100 otic> power due to a small-break LOCA RCS temperature is stable at approximately 500°F All SG pressures are stable at approximately 670 psig Containment pressure is 21 psia and increasing 1 psia every 30 minutes CDA has not actuated Operators have transitioned from 1-E-0, Reactor Trip or Safety Injection, to 1-E-1, Loss of Reactor or Secondary Coolant.

Based on these plant conditions, which ONE* of the following identifies the action(s) required by 1-E-1?

A.

Start Outside Recirc Spray pumps and inject the Chemical Addition Tank.

B.

Start Outside Recirc Spray pumps, but DO NOT inject the Chemical Addition Tank.

C.

Start Quench Spray pumps and inject the Chemical Addition Tank.

D.

Start Quench* Spray pumps, but DO NOT inject the Chemical Addition Tank.

NAPS 2009 NRC SRO Re-Take Exam

81. The crew is performing 1-ES-O.1, Reactor Trip Response, and is establishing emergency boration due to four (4) control rods indicating greater than 10 steps.

The crew was unable to open 1-CH-MOV-1350, Emergency Boration valve from the control room or locally.

Based on the given conditons, which ONE of the following identifies the valve manipulations required to align the boration flowpath lAW the procedure in effect, and includes the indication used to determine when the required amount of boric acid has been added?

A.

Open 1-CH-MOV-1115B & O,*charging pump suction from RWST, then close 1-CH-MOV-1115C

& E, charging pump suction from VCT; Charging flow B.

Open 1-CH-MOV-1115B & 0, charging pump suction from RWST, then close 1-CH-MOV-1115C

& E, charging pump suction from VCT; RWST level C.

Open 1-CH-FCV-1113A, boric acid to blender, and 1-CH-241, manual emergency borate valve; Boric Acid and PG Controller Integrators D.

Open 1-CH-FCV-111'3A, boric acid to blender, and 1-CH-241, manual emergency borate valve; Boric Acid Storage Tank level

NAPS 2009 NRC SRO Re-Take Exam

82. Given the following conditions:

Unit 1 is in Mode 5 for drained-down RCS maintenance Reactor coolant loops are isolated RCS level is +12" above centerline A loss of RHR occurs The crew is performing actions of 1-AP-11, Loss of RHR RCS temperature is rising, and NEITHER RHR pump can be started Which ONE of the following describes the preferred method, and the actions that will be taken to restore core cooling in accordance with 1~AP-11 ?

A.

Hot Leg Injection Forced Feed and Spill; start ONE charging pump, and if necessary, ONE LHSI pump.

B.

Cold Leg Injection Forced Feed and Spill; start TWO charging pumps and BOTH LHSI pumps.

C.

Cold Leg Injection Forced Feed and Spill; start ONE charging pump, and if necessary, ONE LHSI pump.

D.

Hot Leg Injection Forced Feed and Spill; start TWO charging pumps and BOTH LHSI pumps.

NAPS 2009 NRC SRO Re-Take Exam

83. Given the following conditions:

Unit 1 was initially in Mode 2 starting up following a reactor trip Burnup is 14,000 MWD/MTU The "A" main steamline failed catastrophically inside containment Operators have completed 1-E-O, Reactor Trip or Safety Injection and are preparing to transition The OATC notes that "A" loop cold-leg temperature indicates 325°F Based on these plant conditions, which ONE of the following identifies the procedure operators will implement upon leaving 1-E-0, and the indication that will be used to monitor the Subcriticality CSF Status Tree?

A.

1-FR-P.2, Response to Anticipated Pressurized Thermal Shock Condition; Source Range SUR.

B.

1-FR-P.2, Response to Anticipated Pressurized Thermal Shock Condition; Gamma-Metrics Wide-Range.

C.

1-E-2, Faulted Steam Generator Isolation; Source Range SUR.

D.

1-E-2, Faulted Steam Generator Isolation; Gamma-Metrics Wide-Range.

NAPS 2009 NRC SRO Re-Take Exam

84. Given the following conditions:

Unit 1 is at 100% power RCS boron is 1120 ppm Engineering identified calculation errors during the previous Main Steam Safety Valve (MSSV) Testing The data was recalculated, and the following lists the actual setpoints for the affected SG MSSVs:

MS-SV-101A -- 1092 psig MS-SV-103A--1115 psig MS-SV-104A -- 1160 psig MS-SV-1028 -- 1103 psig MS-SV-1038 --1120 psig MS-SV-1058 --1170 psig Which ONE of the following identifies the Technical Specification "Required Action, and the Technical Specification Bases for the action?

(Reference provided)

A.

Reduce power to < 52% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; ensure primary-to-secondary pressure limitations are not exceeded.

B.

Reduce power to < 37% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and reduce PR HI Flux trip to a maximum of 37%

RTP within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ; ensure primary-to-secondary pressure limitations are not exceeded.

C.

Reduce power to < 52% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; provide protection against overpressurizing the Reactor Coolant System.

D.

Reduce power to < 37%

RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and reduce PR HI Flux trip to a maximum of 37%

RTP within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ; provide protection against overpressurizing the Reactor Coolant System.

NAPS 2009 N*RC SRO Re-Take Exam

85. Given the following conditions:

Unit 1 is at 25°A> power The shift is raising power at 0.3% per minute lAW 1-0P-2.1, Unit Startup from Mode 2 to Mode 1 The following alarms are received in succession:

EH FLUID RESERVOIR LOW LEVEL EH FLUID RESERVOIR LOW-LOW LEVEL HP FLUID PUMP LOCKOUT Which ONE of the following identifies the correct operator response?

A.

Go to 1-AP-2.1, Turbine Trip Without Reactor Trip Required; place the running EHC pump in PULL-TO-LOCK.

B.

Go to 1-AP-2.1, Turbine Trip Without Reactor Trip Required; verify the running EHC pump tripped.

C.

Go to 1-E-O, Reactor Trip or Safety Injection; verify the running EHC pump tripped.

D.

Go to 1-E-O, Reactor Trip or Safety Injection; place the running EHC pump in PULL-TO-LOCK.

NAPS 2009 NRC SRO Re-Take Exam

86. Given the following conditions:

Unit shutdown was in progress for scheduled refueling Power was held, and the unit has been stabilized at 40°,lc, because the "A" MFRV did not appear to be responding to decreasing controller demand Local inspection revealed that "A" MFRV is mechanically bound, and is NOT capable of closing Based on these plant conditions, which ONE of the following completes the statement to identify the safety-related function that is not met, and the associated Technical Specification required action?

The "A" MFRV is NOT capable of providing isolation

. TS-required action is to A.

following a reactor trip to prevent an uncontrolled RCS cooldown; enter information-only action for "A" MFRV.

B.

following a high-energy line break on the "A" SG secondary side; enter information-only action for "A" MFRV.

C.

following a reactor trip to prevent an uncontrolled RCS cooldown; close OR isolate "A" MFRV within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

D.

following a high-energy line break.on the "A" SG secondary side; close OR isolate "A" MFRV within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

NAPS 2009 NRC SRO Re-Take Exam

87. Given the following conditions:

Unit 1 was initially at 100°A> power with 1H EDG tagged out for voltage regulator replacement A loss of offsite power occurred, and 1J EDG tripped on overspeed and cannot be restarted Letdown automatically isolated due to decreasing PRZR level, and PRZR level is still slowly decreasing The crew is performing 1-ECA-0.0, Loss of All AC Power.

Based on these plant conditions, which ONE of the following identifies the procedure used to restore power, and the recovery actions once power is restored?

A.

Use 0-OP-6.4, Operation of the SBO Diesel to energize an Emergency Bus; Following power restoration, manually actuate Safety Injection and go to 1-ECA-0.2, Loss of All AC Power with SI Required.

B.

Use 0-AP-10, Loss of Electrical Power to energize an Emergency Bus; Following power restoration, manually actuate Safety Injection and go to 1-ECA-0.2, Loss of All AC Power with SI Required.

C.

Use 0-OP-6.4, Operation of the SBO Diesel to energize an Emergency Bus; Following power restoration, go to 1-ECA-0.2, Loss of All AC Power with SI Required and manually load required equipment.

D.

Use 0-AP-1 0, Loss of Electrical Power to energize an Emergency Bus; Following power restoration, go to 1-ECA-0.2, Loss of All AC Power with SI Required and manually load required equipment.

NAPS 2009 NRC SRO Re-Take Exam

88. Given the following conditions:

Unit 1 is at 90% power "A" & "C" MFW pumps are running, and "B" MFW pump is tagged out "A" & "C" Condensate pumps are running, and "B" Condensate pump is tagged out The following alarms and conditions occur:

G-G6, CONDENSATE PP 1A-1 B-1C AUTO TRIP, alarms and the OATC notes that "A" Condensate pump is tripped F-B6, MAIN FD PPS SUCT HDR LO PRESS, and F-B5, MAIN FD PPS LO DIFF PRESS, are subsequently received MFW pump suction pressure is 270 psig and slowly lowering MFW pump differential pressure is 680 psig and slowly lowering Based on these plant conditions, which ONE' of the following identifies the direction(s) the SRO should provide to the OATC?

A.

Reduce turbine load until the MAIN FD PPS LO DIFF PRESS alarm is clear.

B.

Reduce turbine load until steam flow is less than available feed flow.

C.

Trip the reactor and go to 1-E-0, Reactor Trip or Safety Injection.

D.

Reduce turbine load until MFW pump suction pressure is greater than 300 psig.

NAPS 2009 NRC SRO Re-Take Exam

89. Given the following conditions:

Unit 1 is at 100% power Annunciator H-B2, BATTERY CHGR 1-11 TROUBLE alarms Battery Charger 1-11 DC Output breaker has tripped and CANNOT be reset Based on these plant conditions, which ONE of the following identifies the Technical Specification required action, and the action required by AR-H-B2?

A.

Restore DC electrical power subsystem to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; Place battery charger 1C-I in service.

B.

Restore DC electrical power subsystem to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; Place battery charger 1C-I in service.

C.

Restore DC electrical power subsystem to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; Place battery charger 1C-II in service.

D.

Restore DC electrical power subsystem to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; Place battery charger 1C-II in service.

NAPS 2009 NRC SRO Re-Take Exam

90. Given the following conditions:

Unit 1 has experienced a LOCA with multiple equipment failures The crew is implementing 1-FR-C.1, Response to Inadequate Core Cooling Attempts to establish safety injection flow have been unsuccessful The crew is at step 20 of 1-FR-C.1, which directs the operator to "Check if RCPs Should be Started" Core-exit TCs (CETC) are 1250°F and rising at approximately 3°F/min All RCPs are stopped SG levels are as follows:

"A" -- 25% narrow-range and slowly increasing "B" -- 30% narrow-range and stable "C" -- 50 % wide-range and slowly decreasing Containment pressure peaked at 38 psia, and is now 24 psia and slowly decreasing.

Based on these plant conditions, which ONE of the following identifies the actions and sequence required by 1-FR-C.1?

A.

Start "A" RCP; if CETCs remain the same after "A" RCP is running, then start "B" RCP; if CETCs remain the same after "B" RCP is running, then open PRZR PORVs and block valves.

B.

Start "A" RCP; if CETCs remain the same after "A" RCP is running, then start "B" RCP; if CETCs remain the same after "B" RCP is running, then open all Reactor vent valves and PRZR vent valves.

C.

Start ALL RCPs; if CETCs remain the same after all RCPs are running, then open PRZR PORVs and Block Valves, and open all available RCS vent paths to containment.

D.

Start ALL RCPs; if CETCs remain the same after all RCPs are running, then go to 1-SACRG-1, Severe Accident Control Room Guideline Initial Response.

NAPS 2009 N'RC SRO Re-Take Exam

91. Given the following conditions:

Both units are at 100% power.

1-SW-P-1A and 1-SW-P-1B are running Service Water is throttled in accordance with 0-OP-49.6, Service Water System Throttling Alignment The OATC observes 1-SW-P-1A discharge pressure is 0 psig, and motor current is 10 amps.

Which ONE of the following identifies the procedurally required action, and describes the status of the SW system AFTER the applicable pump is started?

A.

Start 2-SW-P-1A; The Service Water System can perform its required safety function in the event of a DBA only if NO additional failures occur.

B.

Start 2-SW-P-1A; The Service Water System can perform its required safety function in the event of a DBA assuming ONE additional failure.

C.

Start 2-SW-P-1 B; The Service Water System can perform its required safety function in the event of a DBA only if NO additional failures occur.

D.

Start 2-SW-P-1 B; The Service Water System can perform its required safety function in the event of a DBA assuming ONE additional failure.

NAPS 2009 NRC SRO Re-Take Exam

92. The following initial conditions existed:

Both Units at 100% power The 1H Diesel Generator is tagged out for pre-planned minor maintenance The following alarms and indications occur:

Annunciator 1T-03, FREQUENCY 59.8 HERTZ Annunciator 2T-03, FREQUENCY 59.8 HERTZ Operators have verified the alarms are valid 500kV switchyard voltage is 503kV and stable System Operator reports frequency is 59.6 Hertz and stable Based on these plant conditions, which ONE' of the following identifies the most limiting Technical Specification required action?

(REFERENCE PROVIDED),

A.

Restore one offsite circuit to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B.

Restore 1H EOG to operable status within 14 days.

C.

Restore either the offsite circuit or the 1H EOG to operable within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D.

Enter LCO 3.0.3 immediately.

NAPS 2009 NRC SRO Re~Take Exam

93. For the Emergency Condensate Storage Tank to be OPERABLE, it must contain a minimum volume of

_____; the Basis for this requirement is to A.

100,000 gallons; maintain the unit in MODE 3 for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

B.

100,000 gallons; maintain the unit in MODE 3 for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

C.

110,000 gallons; maintain the unit in MODE 3 for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

D.

110,000 gallons; maintain the unit in MODE 3 for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

NAPS 2009 NRC SRO Re-Take Exam

94. Operators are performing 1-ES-1.1, SI Termination, and are preparing to isolate the BIT.

The OATC is not able to open 1-CH-MOV-1373, Charging Pump Recirc Header Isolation Valve.

Based on these plant conditions, which ONE of the following identifies the sequence of actions required in accordance with 1-ES-1.1?

A.

Establish normal charging and maintain 25 gpm charging flow using 1-CH-FCV-1122 in MANUAL, then isolate the BIT.

B.

Establish normal charging and maintain 60 gpm charging flow using 1-CH-FCV-1122 in MANUAL, then isolate the BIT.

C.

Isolate the BIT, then establish normal charging and maintain 25 gpm charging flow using 1-CH-FCV-1122 in MANUAL.

D.

Isolate the BIT, then establish normal charging and maintain 60 gpm charging flow using 1-CH-FCV-1122 in MANUAL.

NAPS 2009 NRC SRO Re-Take Exam

95. In accordance with OP-AA-102, Operability Determination, the is responsible for approving prompt Operability Determinations (ODs).

A.

Shift Manager B.

Operations Manager C.

Engineering Manager D.

Licensing Manager

NAPS 2009 NRC SRO Re-Take Exam

96. 'An activity has been proposed that constitutes a test not described in the UFSAR. Additionally, performance of the activity conflicts with the requirements of the Technical Specifications.

Which ONE of the following identifies the requirements to conduct the activity?

A.

The activity will NOT require a 10 CFR 50.59 Evaluation; prior approval from the NRC is NOT required.

B.

The activity will NOT require a 10 CFR 50.59 Evaluation; prior approval from the NRC l§ required.

C.

The activity will require a 10 CFR 50.59 Evaluation; prior approval from the NRC is NOT required.

D.

The activity will require a 10 CFR 50.59 Evaluation; prior approva'i from the NRC l§ required.

NAPS 2009 NRC SRO Re-Take Exam

97. Given the following conditions:

A General Emergency has been declared and an off-site release is in progress All Emergency Response Organization facilities are manned and operational The TSC has determined that the release point can be isolated by sending an operator into the Auxiliary Building to isolate the flowpath Which ONE of the following identifies the exposure limit for this activity per EPIP-4.04, Emergency Exposure Limits, and the person (by title) who has the FINAL approval for authorizing emergency exposure limits?

A.

5 rem ; Station Emergency Manager.

B.

10 rem ; Station Emergency Manager.

C.

5 rem; Radiological Assessment Director.

D.

10 rem; Radiological Assessment Director.

NAPS 2009 NRC SRO Re-Take Exam

98. Both Units are at 1000/0 power.

A report is received from the FAA that credible insider information has identified that an air threat exists from a Boeing 747 inbound from Miami, Florida with an estimated time of arrival at the station of 30 minutes.

Which ONE of the following identifies the.agency that is contacted to confirm the threat, and the action required by 0-AP-9.01, Station Security Air Threat Operations Response, once the threat is verified to be authentic?

A.

Department of Homeland Security (DHS);

maintain both units stable; shutdown the units only if the threat becomes imminent.

B.

Nuclear Regulatory Commission (NRC);

maintain both units stable; shutdown the units only if the threat becomes imminent.

c.

Department of Homeland Security (DHS);

commence a rapid shutdown of both units using AP-2.2, Fast Load Reduction.

D.

Nuclear Regulatory Commission (NRC);

commence a rapid shutdown of both units using AP-2.2, Fast Load Reduction.

NAPS 2009 NRC SRO Re-Take Exam

99. Which ONE of the following describes a condition required to be reported to the NRC under 10CFR50.72, and the correct time limit for reporting?

A.

Deviation from the plant Technical Specifications authorized pursuant to 10CFR50.54(x);

report within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

B.

Condition that could have prevented fulfillment of a safety function needed to mitigate consequences of an accident; report within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

C.

Failure to perform required surveillance test within technical specification allowable time limits; report within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

D.

The nuclear power plant in an unanalyzed condition that significantly degrades plant safety; report within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

NAPS 2009 NRC SRO Re-Take Exam 100. Given the following conditions:

Unit 1 was at 100°,fc, power The reactor tripped due to a loss of,offsite power A loss of all SG feedwater occurred, and operators transitioned from 1-E-O, Reactor Trip or Safety Injection to 1-FR-H.1, Response to Loss of Secondary Heat Sink Operators could not restore a source of feedwater, and are establishing RCS bleed and feed in accordance with 1-FR-H.1 Operators were able to open ONLY ONE PRZR PORV.

Based on these plant conditions, the RCS bleed path is

, and the crew should A.

adequate; depressurize SGs to less than 610 psig while continuing efforts to re-establish a high pressure source to feed SGs.

B.

adequate; open Reactor and PRZR vents and align a low-pressure water source to feed SGs.

C.

inadequate; depressurize SGs to less than 610 psig while continuing efforts to re-establish a high pressure source to feed SGs.

D.

inadequate; open Reactor and PRZR vents and align a low-pressure water source to feed SGs.

VIRGINIA POWER NORTH ANNA POWER STATION Pressurizer Relief Tank l-RC-TK-2 Volume Versus Indicated Level 1-SC-S.l1 REVISION 1 PAGE 1 OF7 120 100 80

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11223 GalloDl.at 0% Indication J

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1000 2000 3000 4000 SOOO 6000 7000 8000 9000 10000 Gallons APPROVED BY:

CHAIRMAN, STAnON NUCLEAR SAFETY AND OPERATING COMMITI'EE

J::........ ~

. VIRGINIA POWER NORTH ANNA POWER STATION

.1-SC-S.l1

REVISION 1 PAGE 2 OF 7 PRESSURIZER RELIEF TANK l-RC-TK-2 VOLUME VERSUS INDICATED LEVEL

Level transmitter reference leg CaJ.ibrati~ntemperature is 700 P with'density 62.305 Ibm/fe.

Actual tank water temperature is 8SoP With a density of 62.172Ibmlff.

The transmitter will no longer increase with increasing level after 99.S percent indication.

EQUIVALENT INDICATED 1100P VOLUME LEVEL (%)

(GALLONS) 0.0 223.19 O~5 247.91 1.0 273.50 1.5 299.93 2.0 327.17 2.5 355.20 3.0 383.99 3.5 413.51 4.0 443.76 4.5 474.71 5.0 506.34 5.5 538.64 6.0 571.58 6.5 605.16 7.0 639.36 7.5 674.17 8.0 709.56 8.5 745.53 9.0 782.06 9.5 819.15 10.0 856.78 10.5 894.93 11.0 933.60

. 11.5' 972.78 12.0 1012.45 APPROVED BY:

CHAIRMAN, STATION NUCLEAR SAFETY AND OPERATING COMMITIEE DATE:~~/$t

VIRGINIA POWER NORTH ANNA POWER STATION

.. ~"'.:::s....".

" 'l-SC-5.11 REVISION 1 PAGE 3 OF 7 PRESSURIZER RELIEF'TANK l-RC-TK-2 VOLUME*VERSUS INDICATED LEVEL EQUIVALENT INDICATED.

1100P VOLUME LEVEL (%)

(GALLONS) 12.5 1052.61

,13.0 1093.25 13.5 1134.35 14.0 1175.91

14.5 1217.91 15.0 1260.35 15.5 1303.22 16.0 1346.51 16.5 1390.22 17.0 1434.32.

17.5

1478.82 18.0 1523.71 18.5 156.8.98

, 19.0 1614.61 19.5 1660.62 20.0 1706.97 20..5 1753.68

'21.0 1800.72 21.5 1848.11 22.0 1895.81 22.5 1943.84, 23.0 1992.18 23.5 2040."82 24.0 2089.77 24.5

2139.00 25.0 2188.53 25.5 2238.33 26.0 2288.40 26.5 2338.74 27.0 2389.34 27.5

2440.20 28.0 2491.30 28.5 2542.64 29.0 2594.22 29.5 2646.02.

.. VIRGINIA POwER

. NORTH ANNA POWER STATION

1-SC-5.11 REVISION 1 PAGE 4 OF 7 PRESSURIZER REI,IEF TANK l-RC-TK-2 VOLUME VERSUS INDICATED LEVEL EQUIVALENT.

INDICATED 1100P VOLUME LEVEL (%)

(GALLONS) 30.0 2698.05 30.5 2750.30 31.0 2802.75

. 31.5 2855.42 32.0

' 2908.28*

32.5 2961.33 33.0 3014.57 33.5 3068.00

.34.0 3121.60 34.5

3175.37 35.0 3229.30 35.5 3283.40 36.0 3337.64 36.5 3392.04 37.0 3446.58 37.5 3501.2*5 38.0 3556.06 38.5 3610.99 39.0 3666.04 39.5 3721.21 40.0 3776.49 40.5 3831.87 41.0 3887.35 41.5 3942.92 42.0 3998.58 42.5 4054.33 43.0 4110.15 43.5 4166.05 44~O 4222.00 44.5 4278.03 45.0 4334.11 45.5 4390.24 46.0 4446.41 46.5 4502..63 47.0 4558.89

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VIRGINIA POWER NORTH ANNA POWER STATION I-SC-5.11 REVISION 1 PAGES OF 7 PRESSURIZER REI,ffiP TANK l-RC-TK-2 VOLUME ~RSUS INDICATED LEVEL EQUIVALENT INDICATED 1100P VOLUME LEVEL el)

(GALLONS) 47.5 4615.17 48.,0 4671.48 48.5 4727.80

.' 49.0 4784.15 49.5*.

4840.50 50.0 4896.85 50.5 4953.20 51.0 5009.55

, 51.5 5065.89 52.0 5122.21 52.5 5178.51 53.0 5234.78 53.5 5291.02 54.0 5347.22 54;5 5403.37

. 55.0 5459.49 55.5 5515.54 56.0 5571.54 56.5 5627.48 57.0 5683.33 57.5 5739.13 58.0 5794.84 58.5 5850.46 59.0 5906.00 59.5 5961.43 60.0 6016.78 60.5 6072.00 61.0 6127.12 61.5 6182.13 62.0 6237.00 62.5 6291.76 63.0 6346.37 63.5 6400.85 64.0 6455.18 64~'5 6509.36

. VIRGINIA POWER NORTH.ANNA POWER STATION 1-SC-5.11 REVISION 1 PAGE 6 OF 7

. PRESSURIZER RELIEF TANK l-RC-TK-2 VOLUME VERSUS INDICATED LEVEL EQUIVALENT INDICATED

. llOoP VOLUME LEVEL (%)

(GALLONS) 65.0 6563.39 65.5 6617.26 66.0 6670.95 66.5 6724.48.

67.0 6777.82 67.5 6830.99 68.0 6883.96 68.5 6936.74 69.0 6989.31 69.5 7041.68 70.0 7093.83 70.5 7145.77 71.0

. 7197.48 71.5 7248.96 72.0 7300.20 72.5 7351.19 73.0 7401.94 73.5 7452.43 74.0 7502.66 74.5 7552.62 75.0

. 7602.30 75.5 7651.70 76.0 7700.82 76.5 7749.63 77.0 7798.15 77.5 7846.36 78..0 7894.25 78.5 7941.81 79.0 7989.05 79.5 8035.95 80.0 8082.51 80.5 8128.71 81.0 8174.56 81.'5 8220.05 82.0 8265.15

VIRGINIA POWER NORTH ANNA POWER 'STATION 1-SC-5.11 REVISION 1 PAGE 7 OF,7 PRESsURIZER RELIEF TANK l-RC-TK-2 VOLUME VERSUS INDICATED LEVEL EQUIVALENT INDICATED 110°F VOLUME LEVEL ($)

(GALLONS) 82.5 8309.87 83.0 8354.21 83.5 8398.14.

84.0 8441.66 84.5 8484.78 85.0 8527.47 85.5 8569.72 86.0

. 8611.55 86.5.

8652.90 87.0 8693.80 87.5 8734.24 88.0 8774.20 88.5 8813.66 89.'0 8852.62 89.5 8891.07 90.0 8929.00 90.5 8966.40 91.0 9003.26 91.5 9039.55 92.0 9075.27 92.5 9110.42 93.0 9144.96 93.5 9178.90 94.0 9212.21 94.5 9244.88 95.0 9276.90 95.5

' 9308.24 96.0 9338.89 96.5 9368.84 97.0 9398.05 97.5 9426.52 98.0 9454.21 98.5 9481.11

'99.0 9507.18 99.5 9532.40

NUCLEAR DESIGr\\l 3.7 PLANT SYSTEMS POf~TAL --

MSSVs 3.7.1 3.7.1 Main Steam Safety Valves (MSSVs)

LCO 3.7.1 Fi've MSSVs per steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS

~ - - - - - - - NOTE - - - - - - - - - - - - - - - -

Separate Condition entry is allowed for each MSSV.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more steam A.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months generators with one to less than or equal MSSV inoperable and to 52% RTP.

the Moderator Temperature Coefficient (MTC) zero or negative at all power levels.

B.

One or more steam B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months generators with one to less than or equal MSSV inoperable and to the Maximum the MTC positive at Allowable %RTP any power levels.

specified in Table 3.7.1-1 for the OR number of OPERABLE MSSVs.

One or more steam generators with two or AND more MSSVs inoperable.

(continued)

North Anna Units 1 and 2 3.7.1-1 Amendments 231/212

-_. NUCLEAR DES*iGN IN*FORMATION POr{TAl-MSSVs 3.7.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

(continued)

B.2

NOTE-----~---

Only required in MODE 1.

Reduce the Power Range 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Neutron Flux-High reactor trip setpoint to less than or equal to the Maximum Allowable %RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND OR C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more steam generators with greater than or equal to 4 MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SR 3.7.1.1 SURVEILLANCE

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

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per Table 3.7.1-2 in accordance with the Inservice Testing Program. Following testing, lift setting *shall be within +/-1%.

FREQUENCY In accordance with the Inservice Testing Program North Anna Units 1 and 2 3.7.1-2 Amendments 231/212

-- NUCLEAR L1ESIGN INFORMATION PORTAL MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power NUMBER OF OPERABLE MSSVs PER STEAM GENERATOR 4

3 2

North Anna Units 1 and 2 3.7.1-3 MAXIMUM ALLOWABLE POWER

% RTP 52 37 21 Amendments 231/212

-- NUCLEAR DESIGN INFORMJ\\TION PORTAL -,,-

MSSVs 3.7.1 Table 3.7.1-2 (page 1 of 1)

Main Steam Safety Valve Lift Settings

1 MS-SV-101A MS-SV-102A MS-SV-103A MS-SV-104A MS-SV-105A MS-SV-201A MS-SV-202A MS-SV-203A MS-SV-204A MS-SV-205A STEAM GENERATOR

2 Unit 1 VALVE NUMBER MS-SV-101B MS-SV-102B MS-SV-103B MS-SV-104B MS-SV-105B Unit 2 VALVE NUMBER MS-SV-201B MS-SV-202B MS-SV-203B MS-SV-2048 MS-SV-205B

3 MS-SV-101C MS-SV-102C MS-SV-103C MS-SV-104C MS-SV-105C MS-SV-201C MS-SV-202C MS-SV-203C MS-SV-204C MS-SV-205C LIFT SETTING (psig +/- 3%)

1085 1095 1110 1120 1135 1085 1095 1110 1120 1135 North Anna Units 1 and 2 3.7.1-4 Amendments 231/212

NUCLEAR DESIGN INFORMATION PORTAl--

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

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

b. Two emergency diesel generators (EDGs) capable of supplying the onsite Class IE power distribution subsystem(s);

c; One qualified circuit between the offsite transmission network and the onsite Class IE AC Electrical Power Distribution System and one EDG capable of supplying the onsite Class IE AC power distribution subsystem on the other unit for each required shared component; and

d. Required sequencing timing relays.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One LCO 3.8.1..a offsite circuit inoperable.

A.1 AND Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required OPERABLE offsite circuit(s).

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter (continued)

North Anna Units 1 and 2 3.8.1-1 Amendments 231/212

ACTIONS

--- NUCLEAR DESIGN INFORMATION PORTAl.-

AC Sources-Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2 Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no discovery of no offsite power offsite power to available inoperable one train when its redundant concurrent with required feature(s) is inoperabi 1ity of inoperable.

redundant required feature(s)

AND A.3 Restore offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months circuit to OPERABLE status.

AND 17 days from discovery of failure to meet LCO B.

One LCO 3.8.1.b EDG B.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable.

the required offsite circuits.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s) supported discovery of by the inoperable EDG Condition B inoperable when its concurrent with req~ired redundant inoperabi 1ity of feature(s) is redundant inoperable.

required feature(s)

AND (continued)

North Anna Units 1 and 2 3.8.1-2 Amendments 231/212

-- NUCLEAR DESIGN INFORMATION PORTAL,-

ACTIONS AC Sources-Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME B.

(continued)

B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO 3.8.1.b EDG is not inoperable due to common cause failure.

OR B.3.2 Perform SR 3.8.1.2 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE LCO*3.8.1.b EDG.

AND B.4 Restore EDG to 14 days OPERABLE status.

AND 17 days from discovery of failure to meet LCO C.

NOTE----------

C.1.1 Restore inoperable AAC 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Only applicable if DG to OPERABLE status.

Alternate AC (AAC) diesel generator (DG)

AND or one or more EDG on the other unit is C.1.2 Restore inoperable 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable.

EDG(s) on other unit to OPERABLE status.

One LCO 3.8.1.b EDG OR inoperable.

C.2 Restore EDG to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

North Anna Units 1 and 2 3.8.1-3 Amendments 231/212

.- NUCLEAR DESIGN INFORMATION PORTAl--

ACTIONS AC Sources-Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME E.

One required E.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months LCO 3.8.1.c EDG required offsite inoperable.

circuit(s).

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND E.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s) supported discovery of by the inoperable EDG Condition E inoperable when its concurrent with redundant required inoperabi 1ity of feature(s) is redundant inoperable.

required feature(s)

AND E.3 Declare associated 14 days shared component inoperable.

F.

NOTE----------

F.1.1 Restore inoperable AAC 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Only applicable if one DG to OPERABLE status.

or more LCO 3.8.1.b EDG(s) or AAC DG is AND inoperable.

F.1.2 Restore inoperable 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO 3.8.1.b EDG (s) to One required OPERABLE status.

LCO 3.8.1.c EDG inoperable.

OR F.2 Declare associated 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months shared component inoperable.

North Anna Units 1.and 2 3.8.1-5 Amendments 231/212

ACTIONS

,,- r~UCLEAR DESIGN INFORMATION PORTAl---

AC Sources-Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME G.

Two LCO 3.8.1.a G.1 Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from offsite circuits feature(s) inoperable discovery of inoperable.

when its redundant Condition G required feature(s) is concurrent with inoperable.

inoperabi 1ity of redundant required features AND G.2 Restore one offsite 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months circuit to OPERABLE status.

H.

One LCO 3.8.1.a

NOTE--~----------

offsite circuit Enter applicable Conditions inoperable.

and Required Actions of LCO 3.8.9, IIDistribution AND Systems-Operating,1I when Condition H ;s entered with One LCO 3.8.1.b EDG no AC power source to any.

inoperable.

train.

H.1 Restore offsite 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months circuit to OPERABLE status.

OR H.2 Restore EDG to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE status.

I.

Two LCO 3.8.1.b EDGs 1.1 Restore one EDG to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable.

OPERABLE status.

J.

Two required J.1 Declare associated' Immediately LCO 3.8.1.c EDGs shared components inoperable.

inoperable.

North Anna Units 1 and 2 3.8.1-6 Amendments 231/212

NUCLEAR DESIGN INFORIVIATION PORTAl*--

ACTIONS AC Sources-Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME K.

NOTE----------

K.1 Enter appropriate Immediately Separate Condition Conditions and entry is allowed for Required Actions for each sequencing timi~g any component made relay.

inoperable by inoperable sequencing timing relay(s).

One or more required sequencing timing AND relay(s) inoperable.

K.2.1 Place the component(s)

Immediately with the inoperable sequencing timing relay in a condition where it cannot be automatically loaded to associated emergency electrical bus.

OR K.2.2 Declare the associated Immediately EDG inoperable.

L.

Required Action and L.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, AND B, C, G, H, or I not met.

L.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months M.

Three or more of M.1 Enter LCO 3.0.3.

Immediately LCO 3.8.1.a and LCO 3.8.1.b AC sources inoperable.

North Anna Units 1 and 2" 3.8.1-7 Amendments 231/212

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 Points 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 D

C B

B D

D D

A C

B A

A C

D A

C C

A A

D A

A A

B A

D B

A A

C A

A A

D A

C B

B C

C A

A A

A B

B

49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 Points 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 ANSWER KEY REPORT NAPS 2009 NRC ~)ie~lake Exam sSRo

~

Amwe~~~~~~~~~~~~~~-

o A

D A

A A

C A

D C

o A

A C

A B

o B

A D

o D

o A

A A

ANSWER KEY REPORT NAPS 2009 NRC SRO Re-Take Exam 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 D

D D

A C

D A

D C

A D

C C

A A

D D

C B

A D

B D

A D

ML090690067

Text

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