Latest revision as of 11:32, 12 March 2020

Text

Initial Exam 2009-301 Final SRO Written Examination
	ML091700317
Person / Time
Site:	McGuire, Mcguire
Issue date:	06/19/2009
From:	; NRC/RGN-II
To:	; Duke Energy Corp
References
	Download: ML091700317 (146)
	v • d • e

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

Date: Facility/Unit: McGuire Nuclear Station Region: I @/ III/IV Reactor Type: @/CE/BW/GE Start Time: Finish Time:

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

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

Applicant's Signature Results RO/SRO-Only/Total Examination Values / / Points Applicant's Score / / Points Applicant's Grade / / Percent NUREG-1021, Revision 9

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator

1. A B C 0 26. A B C 0

2. A B C 0 27. A B C 0

3. A B C 0 28. A B C 0

4. A B C 0 29. A B C 0

5. A B C 0 30. A B C 0

6. A B C 0 31. A B C 0

7. A B C 0 32. A B C 0

8. A B C 0 33. A B C 0

9. A B C 0 34. A B C 0

10. A B C 0 35. A B C 0

11. A B C 0 36. A B C 0

12. A B C 0 37. A B C 0

13. A B C 0 38. A B C 0

14. A B C 0 39. A B C 0

15. A B C 0 40. A B C 0

16. A B C 0 41. A B C 0

17. A B C 0 42. A B C 0

18. A B C 0 43. A B C 0

19. A B C 0 44. A B C 0

20. A B C 0 45. A B C 0

21. A B C 0 46. A B C 0

22. A B C 0 47. A B C 0

23. A B C 0 48. A B C 0

24. A B C 0 49. A B C 0

25. A B C 0 50. A B C 0 Page 2 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator

51. A B C D 76. A B C D

52. A B C D 77. A B C D

53. A B C D 78. A B C D

54. 79. A B C D A B C D

55. A B C D 80. A B C D

56. A B C D 81. A B C D

57. A B C D 82. A B C D

58. A B C D 83. A B C D

59. A B C D 84. A B C D

60. A B C D 85. A B C D

61. A B C D 86. A B C D

62. A B C D 87. A B C D

63. A B C D 88. A B C D

64. A B C D 89. A B C D

65. A B C D 90. A B C D

66. A B C D 91. A B C D

67. A B C D 92. A B C D

68. A B C D 93. A B C D

69. A B C D 94. A B C D

70. A B C D 95. A B C D

71. A B C D 96. A B C D

72. A B C D 97. A B C D

73. A B C D 98. A B C D

74. A B C D 99. A B C D

75. A B C D 100. A B C D Page 3 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 1 1 Pt Given the following:

Due to an under-frequency relaying problem, all NCPs tripped causing a reactor trip

The crew is performing the actions of ES-O.1 (Reactor Trip Response)

NC System T-Hots are 575°F and slowly rising

NC System T-Colds are 549°F and slowly lowering

All S/G levels indicate 8% NR and rising slowly Which ONE (1) of the following are action(s) required in accordance with ES-0.1 for this condition?

A. Throttle open steam dumps or raise CA flow until NC system T-Hots are stabilized.

B. Throttle open steam dumps or raise CA flow until NC system T-Ave is stable at or trending to 55rF.

C. Close steam dumps and throttle CA flow to minimize cooldown; maintain total feed flow greater than 450 GPM.

D. Close steam dumps and throttle CA flow as necessary to stabilize NC system T-Colds; no other restrictions apply.

Page 4 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 2 1 Pt Given the following conditions on Unit 1:

SI actuated due to a LOCA inside Containment

"A" and "B" NI pumps are TRIPPED

"A" and "B" NV pumps are TRIPPED

NC system pressure is 250 PSIG

All other equipment is running as designed

Containment pressure is 8 PSIG

E-O, Reactor Trip or Safety Injection has been implemented Which ONE (1) of the following describes the required action and reason for the action with respect to the NC pumps?

A. Leave all NC pumps running to provide forced cooling flow of the NC system.

B. Leave all NC pumps running to prevent phase separation of NC system liquid.

C. Stop all NC pumps to prevent mechanical damage to the pump and motor.

D. Stop all NC pumps to minimize fluid mass loss out of the break.

Page 5 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 3 1 Pt Given the following conditions:

Unit 1 is operating at 100% RTP

AP-008 (Malfunction of NC Pump) has been implemented due to a seal malfunction on 1B NC pump Which ONE (1) of the following indicates that the failure is the 1B NC pump #2 Seal?

A.. #1 Seal Leak off flow - GOING UP

NC Pump number 2 Seal Standpipe HIGH level alarm - LIT B.. #1 Seal Leak off flow - GOING UP

NC Pump number 2 Seal Standpipe LOW level alarm - LIT C.. #1 Seal Leak off flow - GOING DOWN

NC Pump number 2 Seal Standpipe LOW level alarm - LIT D.. #1 Seal Leak off flow - GOING DOWN

NC Pump number 2 Seal Standpipe HIGH level alarm - LIT Page 6 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 4 1 Pt Given the following:

A startup is in progress on Unit 2 following refueling

NC system is solid with one NC pump in service

SU-8 (Heatup To 200°F) is in progress

LTOP is in service

NC system pressure is 360 PSIG Which ONE (1) of the following describes how bulk non-condensable gases are removed from the NC system prior to forming a bubble in the Pressurizer?

A. Notify Primary Chemistry to add Hydrazine.

B. Cycle the Reactor Vessel Head vents.

C. Vent the Volume Control Tank.

D. Cycle Pressurizer PORVs.

Page 7 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 5 1 Pt Given the following:

Unit 2 is operating at 100% RTP

Rod control in manual Which ONE (1) of the following will cause the Channel 2 OTL1T trip setpoint to decrease?

A. N-42 upper detector fails high.

B. Auctioneered high T-ave fails high.

C. NC system wide range pressure channel fails low.

D. Power reduction to 50% with normal pressure and temperature.

Page 8 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 6 1 Pt Unit 1 operators are responding to a LOCA in E-O.

Given the following conditions:

The OATC is performing supplementary action steps in E-O.

The 2nd RO is busy manually controlling feedwater flow to the SGs.

An alarm annunciates on the electrical relay back panel behind the main control boards (1AD11-ES, Transfer Trip System B Trouble).

What is the correct response by the RO who is the OATC in accordance with OMP 2-27 A. The OATC can investigate the alarm because the surveillance area does not apply during EOPs.

B. The OATC can investigate the alarm because the electrical relay panel is within the OATC defined surveillance area.

C. The OATC must remain within the surveillance area. The alarm should not be investigated until another operator is available.

D. The OATC is allowed to momentarily leave the surveillance area to investigate the alarm in the event of an emergency affecting the safety of operations.

Page 9 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 7 1 Pt Given the following:

Unit 1 was operating at 100% RTP

S/G 1B steam line breaks inside Containment

Containment pressure is currently 4.0 PSIG Which ONE (1) of the following correctly describes CF system indications on the Main Control Board as a result of the above conditions?

A. A and B CF Pump Turbine Tripped annunciators - LIT 1CF-2 and 1CF-5 (CF Pump Discharge valves) indicate CLOSE B. A and B CF Pump Turbine Tripped annunciators - LIT 1CF-2 and 1CF-5 (CF Pump Discharge valves) indicate OPEN C. A and B CF Pump Turbine speed indications - 2800 RPM 1CF-2 and 1CF-5 (CF Pump Discharge valves) indicate CLOSE D. A and B CF Pump Turbine speed indications - 2800 RPM 1CF-2 and 1CF-5 (CF Pump Discharge valves) indicate OPEN Page 10 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 8 1 Pt A Safety Injection due to a stuck open Pressurizer Safety valve has occurred.

Which ONE (1) of the following describes an indication which will be adversely affected by this accident condition and the reason the indication is affected?

A. Subcooling.

Voiding in the reactor vessel head causes inaccurate CET input to the subcooling monitor.

B. Pzr level.

Voiding in the reactor vessel head forces water out of the reactor vessel into the Pressurizer.

C. Subcooling.

Rapid vapor space depressurization causes inaccurate pressure input to the subcooling monitor.

D. Pzr level.

Rapid vapor space depressurization results in Pzr level reference leg flashing.

Page 11 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 9 1 Pt Unit 2 is responding to a LOCA and the operators have reached the step of ES-1.2 (Post LOCA Cooldown and Depressurization) which requires isolating cold leg accumulators.

Which ONE (1) of the following describes why the Cold Leg Accumulators are isolated at this step?

A. To prevent injecting the CLA nitrogen bubble into the reactor and creating a hard bubble in the Pressurizer.

B. Ensures that NC system pressure can be reduced to below NO pump shutoff head, and further Cold Leg Accumulator injection will not occur.

C. Ensures that the inventory contained in the CLA's is available should the need arise to recover core cooling later in the procedure.

D. To prevent CLA injection when the NC system depressurizes, resulting in thermal stress and PTS concerns from the cold water on the reactor vessel.

Page 12 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 10 1 Pt Which of the following best describes the Priority Mode of Diesel Sequencer operation?

A. Actuated by a Blackout signal and begins sequencing if bus voltage ONLY is greater than minimum required.

B. Actuated by a Blackout signal and begins sequencing if bus voltage AND frequency are greater than minimum required.

C. Actuated by a Safety Injection signal and begins sequencing if bus voltage ONLY is greater than minimum required.

D. Actuated by a Safety Injection signal and begins sequencing if bus voltage AND frequency are greater than minimum required.

Page 13 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 11 1 Pt Given the following conditions:

The crew has entered FR-H.1 (Response to Loss of Secondary Heat Sink) due to a loss of inventory in the S/Gs and failure of the CA pumps to start Time 1400 1410 1420 1430 S/G 1A WR [%] 43 37 30 26 S/G 1B WR [%] 41 32 25 20 S/G 1C WR [%] 42 34 29 25 S/G 10 WR [%] 40 33 26 21 Total feed flow [GPM] 0 0 0 0 Cont press [PSIG] 0.75 2.1 3.2 2.8 Which ONE (1) of the following is the EARLIEST time (if any) that the crew would be required to initiate NC system Feed and Bleed based on plant conditions?

A. Feed and Bleed is not required.

B. 1410 C. 1420 O. 1430 Page 14 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 12 1 Pt Given the following plant conditions:

A reactor trip has occurred on Unit 1 as a result of a Loss of Offsite Power (LOOP)

1B DIG has started and loaded normally

1A DIG did not start and attempts to start it have been unsuccessful

The operating crew has entered ES-O.2 (Natural Circulation Cooldown)

The crew is preparing to start a cooldown and depressurization using natural circulation Based on the conditions above, the required SUBCOOLING MARGIN during the cooldown and depressurization is (1) based on (2)

A. (1) 50°F (2) preventing void formation in the reactor vessel head.

B. (1) 50°F (2) providing adequate thermal driving head for natural circulation.

C. (1) 100°F (2) preventing void formation in the reactor vessel head.

D. (1) 100°F (2) providing adequate thermal driving head for natural circulation.

Page 15 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 13 1 Pt Given the following:

A BLACKOUT has occurred on 2ETB

DIG 2B failed to start due to an S6N relay actuation

Annunciator PaneI2AD-11-F4 (Battery EVCD Undervoltage) is in alarm Per AP-15 (Loss of Vital or Aux Control Power) which ONE (1) of the following addresses the Battery EVCD undervoltage condition?

A. Cross tie EVDD to EVDB.

B. Align Battery Charger EVCS to Battery EVCD.

C. Swap Battery Charger Connection box to 1EMXB.

D. Swap Battery Charger Connection box to 1EMXH.

Page 16 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 14 1 Pt Given the following on Unit 1:

1KC-132 (LID HX Cooling Water Control Valve) has failed

LD HX outlet temperature is 115°F and increasing If LD Hx outlet temperature reaches (1) 1NV-127A (LD Hx Outlet 3-Way Temp Cntrl) will AUTO divert letdown flow to the (2)

A. (1) 138°F (2) VCT to protect the demineralizer resin B. (1) 13SOF (2) RHT to prevent a reactor power reduction C. (1) 120°F (2) VCT to protect the demineralizer resin D. (1) 120°F (2) RHT to prevent a reactor power reduction Page 17 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 15 1 Pt Given the following conditions on Unit 1:

A small break LOCA occurred in Containment

SI has actuated on both trains

1B NV pump tripped on overcurrent

Containment pressure is 3.5 PSIG Based on the above conditions, ALL cooling would be lost to the NC pump ...

A. seals ONLY.

B. motor bearings ONLY.

C. seals AND pump lower bearings.

D. motor bearings AND pump lower bearings.

Page 18 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 16 1 Pt Given the following initial conditions:

Unit 2 is operating at 100% RTP

PZR Pressure Channel Select is in 1-2 position

NC system pressure is 2235 PSIG The following occurs:

2NC-27 (Pressurizer Spray Control) RED lamp is lit and the GREEN lamp is extinguished

2NC-29C (Pressurizer Spray Control) GREEN lamp is lit and the RED lamp is extinguished

NC system pressure is decreasing

Attempts to close 2NC-27 using its controller have been unsuccessful Which ONE (1) of the following describes the actions required in accordance with AP-11 (Pressurizer Pressure Anomalies)?

A. Manually trip the reactor and stop 2A NC pump.

B. Ensure the PZR Channel Select is swapped to a non-affected channel.

C. Manually place the Pressurizer Spray Emergency Close switch to the close position.

D. Ensure the Pressurizer Backup heaters energize at 2210 PSIG to stop the pressure decrease.

Page 19 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 17 1 Pt Given the following:

Unit 1 is shutdown

NC system temperature is 140°F

'A' Train components are in operation

DIG 1B is tagged for maintenance

The NC system WR level is 55 inches Given the following events occur:

The 1A RN pump trips on overcurrent

The operating crew has completed the actions of AP-20 Case 1 (Loss of Operating RN Train)

What actions are required (if any) per Tech Spec 3.7.7 (Nuclear Service Water System)?

A. RN is not required to be Operable in Mode 5:

No action is required per this LCO.

B. Only ONE train of RN is required to be Operable:

No action is required per this LCO.

C. BOTH trains of RN are required to be Operable:

Enter the applicable action statement of TS 3.7.7.

D. Only 9NE train of RN is required to be Operable but must have Emergency Power available:

Enter the applicable Action Statement of TS 3.7.7.

Page 20 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 18 1 Pt Given the following conditions:

A LOCA has occurred on Unit 1

ECA 1.1 (Loss of Emergency Coolant Recirculation (ECR) has been implemented

All ECCS pumps are taking suction from the FWST

FWST level is 100 inches

Containment sump level is 1.5 feet Which ONE (1) of the following describes the mitigation strategies of ECA 1.1?

A. Attempts are made to locate and isolate the cause of the loss of inventory from containment and makeup is initiated to the FWST. NC system is cooled down and depressurized to allow the CLA's to inject.

B. Attempts are made to locate and isolate the cause of the loss of inventory from containment and makeup is initiated to the FWST. NC System is cooled down, depressurized and placed on RHR cooling.

C. Containment spray and Injection flow are minimized, makeup is initiated to the FWST, NC System conditions are maintained stable until FWST and containment sump inventories are recovered or ECR capability is restored.

D. Containment spray and Injection flow are minimized, attempts are made restore ECR, makeup is initiated to the FWST, and the NC System is cooled down, depressurized and placed on RHR cooling.

Page 21 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 19 1 Pt Given the following:

Unit 1 is shutdown in Mode 5

A purge of the Containment atmosphere is in progress

Personnel are preparing for refueling operations

1EMF-38(L) (Containment Particulate Low Range) experiences a loss of power Which ONE (1) of the following lists the automatic action(s) that will occur?

A. VQ secured, VP secured, Containment Sump pumps 1A 1, 1A2, 1B1, 1 B2 and Incore Sump pump secured, Containment evacuation alarm sounded.

B. VQ secured, VP secured, Containment Sump pumps 1A 1, 1A2, 1B1, 1 B2 and Incore Sump pump secured ONLY.

C. VQ and VP secured ONLY.

D. VQ secured ONLY.

Page 22 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 20 1 Pt Given the following:

1NV-241 (Seal Water Injection Flow Control) failed closed and cannot be reopened

Normal LID has been isolated

Excess LID has been placed in service at a flow rate of 26 GPM

Seal Injection is presently 7 GPM per NC pump

Total NC pump seal leak off flow is 12 GPM

PZR level is 44%

Assuming no further operator action is taken, how long will it take for the PZR Heater feeder breakers to trip open?

REFERENCE PROVIDED A. 164 Minutes B. 180 Minutes C. 360 Minutes D. 520 Minutes Page 23 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 21 1 Pt Which ONE (1) of the following describes the effect of an UNDER-COMPENSATED Intermediate Range Channel following a Reactor Trip?

A. Channel will indicate high, Source Ranges will correctly energize due 1/2 logic.

B. Channel will indicate low, prematurely energizing the Source Ranges due to 1/2 logic.

C. Channel will indicate high, preventing P-6 from energizing the Source Ranges due to not satisfying 2/2 logic.

D. Channel will indicate low, Source Ranges will not energize until P-6 is met from other Channel due to 2/2 logic.

Page 24 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 22 1 Pt lAW Tech Spec 3.7.1 (Main Steam Safety Valves), the FIRST safety valve is set to open at --11L PSIG and the LAST safety valve is set to open at ~ PSIG.

A. (1) 1170 (2) 1220 B. (1) 1190 (2) 1220 C. (1) 1170 (2) 1225 D. (1) 1190 (2) 1225 Page 25 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 23 1 Pt Which ONE (1) of the following is the basis for the minimum temperature of the FWST as it relates to a Containment Spray actuation?

A. Prevent soluble boron from coming out of solution which could result in clogging of the NS spray nozzles and reduced spray flow.

B. Prevent an excessive reduction in containment pressure, which could decrease the rate at which steam can be vented out the break and increases peak clad temperature.

C. Prevent exceeding the maximum external pressure load on the containment shell due to an internal vacuum caused by overcooling of the Post-Accident Containment atmosphere.

D. Prevent thermal shock of NS system piping specifically the NS spray nozzles which could be exposed to adverse containment conditions following a High Energy Line Break inside Containment.

Page 26 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 24 1 Pt Given the following:

Unit 1 is operating at 100% RTP

The 1A DIG has just been declared inoperable due to an oil leak on the Woodward governor Which ONE (1) of the following describes ALL checks which must be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure compliance with TS 3.8.1 (AC Sources - Operating)?

A. Flowpaths through which 1ETA AND 1ETB are being supplied by offsite power.

B. Flowpath through which 1ETA is being supplied by offsite power AND determine that 1ETA is being supplied independently from 1ETB.

C. Flowpath through which 1ETA OR 1ETB is being supplied by offsite power AND determine that 1ETA and 1ETB are being supplied independently.

D. Flowpaths through which 1ETA AND 1ETB are being supplied by offsite power AND determine that 1ETA and 1ETB are being supplied independently.

Page 27 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 25 1 Pt Given the following conditions on Unit 1:

Reactor power is 40%

1CM-843 (1A Main Condenser Vacuum Breaker) has failed in the intermediate position. Maintenance is attempting to close this valve

EXH HOOD TEMP HI alarm is lit

Condenser vacuum is 26 inches Hg and DEGRADING slowly

(1ZJP5000) CSAE steam pressure is reading 100 PSIG Which ONE (1) of the following actions is directed by AP-23 (Loss of Condenser Vacuum) and will be the MOST effective in mitigating this event?

A. Reduce turbine load B. Start additional RC pumps C. Open the Exhaust Hood Spray valves D. Dispatch operator to restore CSAE steam pressure Page 28 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 26 1 Pt In the event of a fire which activates the fire protection system, which ONE (1) of the following sequence of events is the normal starting sequence for the fire pumps?

A. LEAD Jockey Pump starts at 43" level in the pressurizer tank STBY Jockey Pump starts at 41" level in the pressurizer tank "A" Fire Pump starts at 81 psig "B" Fire Pump starts at 76 psig "c" Fire Pump starts at 71 psig B. LEAD Jockey Pump starts at 51" level in the pressurizer tank STBY Jockey Pump starts at 45" level in the pressurizer tank "A" Fire Pump starts at 81 psig "B" Fire Pump starts at 76 psig "c" Fire Pump starts at 71 psig C. LEAD Jockey Pump starts at 43" level in the pressurizer tank STBY Jockey Pump starts at 41" level in the pressurizer tank "A" Fire Pump starts at 83 psig "B" Fire Pump starts at 78 psig "c" Fire Pump starts at 73 psig D. LEAD Jockey Pump starts at 51" level in the pressurizer tank STBY Jockey Pump starts at 45" level in the pressurizer tank "A" Fire Pump starts at 83 psig "B" Fire Pump starts at 78 psig "c" Fire Pump starts at 73 psig Page 29 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 27 1 Pt Which ON E (1) of the following is supplied from DC Distribution Center SDSP?

A. FWPT 'A' Emergency Oil pump B. 1SM-1AB, 10 Main Steam Isol C. Turbine Emergency Bearing Oil pump D. 1CA161C, RN Supply to Turbine Driven CA pump Page 30 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 28 1 Pt Which ONE (1) of the following describes the correct reason for ensuring that NC system pressure, is less than 2335 PSIG during implementation of FR-S.1 (Response to Nuclear Power Generation/ATWS)?

A. Minimize the potential for reactor vessel pressurized thermal shock.

B. Verifies or ensures a sufficient boron flow rate into the NCS.

C. Minimize the potential of the PRT rupture disc bursting.

D. Verifies that the pressurizer PORVs are closed.

Page 31 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 29 1 Pt Unit 2 is currently operating at 60% power with a load increase in progress. Given the following events and conditions:

Pressurizer level control is in the "1-2" position (normal control)

Pressurizer level is at program level and in "automatic"

Pressurizer level Channel 1 fails at its current level If the channel failure remains undiscovered, which ONE (1) of the following describes the system response as plant load is increased from 60% to 90%?

A. Charging flow decreases Letdown isolates Pressurizer heaters turn off B. Charging flow increases Pressurizer backup heaters energize Pressurizer level stabilizes at ..., 52%

C. Charging flow increases Pzr Low Level Control annunciator alarms Pressurizer level increases to the trip setpoint D. Charging flow decreases Pzr Low Level Control annunciator alarms Letdown isolates Page 32 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 30 1 Pt Given the following:

A LOCA has occurred on Unit 1

ES-1.2 (Post LOCA Cooldown and Depressurization) has been implemented Which ONE (1) of the following describes major actions of ES-1.2?

A. (1) Establish a 100°F/hrcooldown rate (2) Sequentially stop all but one NV pump B. (1) Cooldown and depressurize at maximum rate (2) Sequentially stop all but one NV pump C. (1) Establish a 1OO°F/hr cooldown rate (2) Secure all NC pumps D. (1) Cooldown and depressurize at maximum rate (2) Secure all NC pumps Page 33 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 31 1 Pt Given the following:

Unit 2 is operating at 100% RTP

A Loss of Offsite Power occurs on both Units 1 and 2 Which ONE (1) of the following provides the assured source of cooling water to maintain containment temperatures within Tech Spec limits?

A. BOTH '2A' AND '2B' RN pumps B. RV pump selected in "Auto" C. '2A' RN pump ONLY D. '2B' RN pump ONLY Page 34 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 32 1 Pt Unit 1 was operating at 100% when the following sequence of events occurred:

A Loss-of-Offsite Power resulted in a Reactor Trip

When the 1A DIG attempted to load 1ETA, a fault on the bus resulted in an 87G (Generator Differential) relay actuation

The plant transient resulted in a LOCA inside containment and a safety injection

Containment pressure is 3.0 PSIG and increasing Which ONE (1) of the following describes the status of the VE (Annulus Ventilation) fans?

A. ONLY 1B VE fan is running.

B. 1A and 1B VE fans are running.

C. ONLY 1B VE fan will start after a 10 minute time delay.

D. 1A and 1B VE fans will start after a 10 minute time delay.

Page 35 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 33 1 Pt Unit 1 & 2 are operating at 100% RTP:

The TCC has notified the Control Room of impending grid frequency disturbances

The crew is performing actions of AP-05 (Generator Voltage and Electric Grid Disturbances)

Both Unit 1 & 2 have been operating with frequency less than 59.5 Hz for greater than 20 Minutes Which ONE (1) of the following describes why Turbine First Stage Impulse pressure is reduced to less than 340 PSIG prior to opening the Switchyard PCB's in accordance with AP-05?

A. Opening the Switchyard PCB's prior to Turbine First Stage Impulse pressure decreasing below 340 PSIG will result in a Turbine Trip.

B. Turbine First Stage Impulse pressure <340 PSIG ensures that a Turbine Trip will occur once separated from the Grid but that a Reactor Trip will not occur.

C. Opening the Switchyard PCB's above 340 PSIG Turbine First Stage Impulse pressure will initiate a Complete Loss of Load transient and the Main Generator load will reduce to In-House loads.

D. Turbine First Stage Impulse pressure <340 PSIG ensures that the AMSAC circuit will not trip the Main Turbine during the resulting feedwater transient as the Turbine runs back to In-House loads.

Page 36 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 34 1 Pt Given the following conditions:

Unit 1 is responding to a LOCA

All sources of feedwater have been lost, S/G NR levels are 17%

and decreasing

NC pumps are secured

FR-C.1 (Response to Inadequate Core Cooling) has been implemented

NI and NV pumps are unavailable

Peak Containment pressure reached 2.5 PSIG

S/G depressurization has failed to restore adequate core cooling

Core Exit Thermocouples are currently indicating 1210°F

NC pump support requirements can NOT be met What is the major action(s) required by FR-C.1 under these conditions?

A. Do NOT restart NC pumps, open all PZR PORVs and head vents to depressurize the NC system.

B. Restart all NC pumps and restore secondary heat sink in FR-H.1 before proceeding in FR-C.1.

C. Restart NC pumps one at a time until CETs are less than 1200°F to force two phase flow through the core for core cooling.

D. Do NOT restart NC pumps, continue efforts to initiate feed and bleed of the NC system to restore core cooling.

Page 37 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 35 1 Pt Unit 1 is shutdown with the following conditions:

1A and 1B ND trains are in service providing shutdown cooling lAW SO-9 (ND System Parallel Heat Exchanger Operation)

The flow rate on each ND train is currently 1500 GPM

NC system temperature is being maintained at 170°F

The air line for 1ND-34 (1A & 1B ND Hx Byp Isol) breaks off Which ONE (1) of the following describes the indications that will be observed by the Operators in the Main Control Room over the next 15 minutes?

1A ND Train Total 1B ND HX Inlet Flowrate Temperature A. Increase Increase B. Increase Decrease C. Decrease Increase D. Decrease Decrease Page 38 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 36 1 Pt Given the following events and conditions on Unit 2:

A large break LOCA occurred

SI and Sequencers were reset

Subsequently a Loss of Offsite Power occurs

2B DIG failed to start Which ONE (1) of the following correctly describes the restoration process for the Train A NV, NI, ND, and CA pumps?

A. 2A NV and 2A CA pumps automatically restart 2A NI and 2A ND pumps must be restarted by operator action B. 2A NV and 2A NI pumps automatically restart 2A ND and 2A CA pumps must be restarted by operator action C. 2A ND and 2A CA pumps automatically restart 2A NV and 2A NI pumps must be restarted by operator action D. 2A NI and 2A ND pumps automatically restart 2A NV and 2A CA pumps must be restarted by operator action Page 39 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 37 1 Pt Given the following conditions on Unit 1:

Rod M-4 has dropped due to a blown fuse

AP-14 (Rod Control Malfunction) has been implemented

The fuse has been replaced and the dropped rod has been realigned with the rest of the rods in Control Bank D Due to a miscommunication with I&E, the Pulse-to-Analog Converter was not reset to the correct bank step position.

The "Control Bank Lo-Lo Limit" annunciator will:

A. alarm SOONER than required because the RIL function of the OAC is calculating CB-D rods to be higher than actual.

B. alarm LATER than required because the RIL function of the OAC is calculating CB-D rods to be higher than actual.

C. alarm SOONER than required because the RIL function of the OAC is calculating CB-D rods to be lower than actual.

D. alarm LATER than required because the RIL function of the OAC is calculating CB-D rods to be lower than actual.

Page 40 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 38 1Pt Given the following sequence of events:

A Large Break LOCA occurs on Unit 1

All ECCS systems are injecting from the FWST

Safety Injection is reset

FWST level is currently 200 inches An Operator depresses the 'SS-RESET' pushbuttons on the 'CNTRL PERMISSIVE FOR RECIRC MODE 1NI-185A 1 184B' switches.

Concerning the following valves:

1NI-185A (RB Sump to Train AND & NS)

1NI-184B (RB Sump to Train B ND & NS)

1ND-19A (A ND Pump Suction from FWST or NC)

1ND-4B (B ND Pump Suction from FWST or NC)

Which ONE (1) of the following describes what the Operator observes with regards to the automatic operation of the ECCS valves listed above after the SS-RESET pushbuttons are depressed?

A. Immediately after depressing the SS-RESET pushbuttons, 1NI-185A1184B, OPEN AND 1ND-19A/4B CLOSE.

B. Immediately after depressing the SS-RESET pushbuttons, 1NI-185A/184B OPEN AND 1ND-19A/4B REMAIN OPEN.

C. When 2/3 FWST Lo Level Bistables are received, 1NI-185A/184B OPEN AND 1ND-19A/4B CLOSE.

D. When 2/3 FWST Lo Level Bistables are received, 1NI-185A/184B REMAIN CLOSED AND 1ND-19A/4B REMAIN OPEN.

Page 41 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 39 1 Pt Given the following:

AP-17 (Loss of Control Room) has been entered Which ONE (1) of the following describes the reason why the operator at the reactor trip breakers SHALL NOT TRIP the reactor until directed to do so by the SRO at the ASP?

A. Ensures control is established at the Auxiliary Shutdown Panel before placing the plant in a transient.

B. A "fire" may have caused the Control Room evacuation and a coordinated effort is required for the SRO to isolate controls from the normal power source(s).

C. Emergency Soration must be initiated prior to tripping the Reactor and because the control room indications are no longer available, action has to be verified by the SRO at the ASP.

D. Reactivity manipulations must be performed or directed by a licensed operator; therefore the operator at the Reactor Trip breakers must be directed by the SRO at the ASP.

Page 42 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 40 1 Pt Given the following:

Unit 1 is operating at 100% RTP with all systems in normal

Instrument Air (VI) is inadvertently lost to containment Which ONE (1) of the following Main Control Board indications will occur as a result?

A. 1NV-24B (NC Loop To Excess Letdown Heat Exchanger Isolation) in the CLOSE position.

B. 1NV-35A (Variable Letdown Orifice Outlet Flow Control Isolation) in the CLOSE position.

C. 1NV-459 (Variable Letdown Orifice Outlet Flow Control) in the OPEN position.

D. 1NV-1A (NC Letdown Isolation Valve) in the OPEN position.

Page 43 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 41 1 Pt Unit 1 is operating at 100% power with the PZR level control system in a 1-2 lineup.

The following events occur in the order listed:

1) Letdown isolates and all pressurizer heaters de-energize

2) Pressurizer level increases

3) Charging flow decreases What is the cause of these indications?

A Pressurizer level channell fails high.

B. Pressurizer level channel II fails low.

C. Pressurizer Level Master Controller output has failed low.

D. Auctioneered Hi-Tave input to the Pzr Control circuit fails low.

Page 44 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 42 1 Pt During recovery from a Loss of all AC power, operators are performing actions in ECA-O.O (Loss of All AC Power).

Which ONE (1) of the following describes the importance of locally ISOLATING the KC supply to the NC pump thermal barriers prior to restarting a KC pump?

A. To prevent damage to the NC pump thermal barriers due to thermal shock.

B. To prevent potential mechanical failure of the NC pump thermal barriers due to water hammer.

C. To prevent steam from forming and circulating in the KC system and ensures the KC system is available to cool equipment necessary for recovery.

D. To ensure elevated heat loads as a result of the loss of all AC power are within the design cooling capacity of the KC system prior to starting a KC pump.

Page 45 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 43 1 Pt Given the following conditions on Unit 1:

ES-1.2 (Post LOCA Cooldown and Depressurization) is being performed due to a LOCA inside Containment.

1A NV pump and 1A NI pump are running

Containment pressure is 4 PSIG

Pressurizer level is 33%

BOTH subcooling monitors are INOPERABLE

NC system CETs indicate 500°F

NC system WR pressure indicates 750 PSIG Which ONE (1) of the following describes the condition of the fluid in the NC system and the required actions (if any) based on these indications?

REFERENCE PROVIDED A. Saturated; Manually initiate a Phase A Isolation.

B. Subcooled; Manually initiate a Phase A Isolation.

C. Saturated; Start 1B NI and 1B NV pumps.

D. Subcooled; Start 1B NI and 1B NV pumps.

Page 46 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 44 1 Pt Unit 1 is responding to a steam break inside containment from 100%

RTP.

Given the following events and conditions:

Narrow range S/G level is 15% for each intact S/G

The NCPs were tripped

FR-P.1 (Response to Imminent Pressurized Thermal Shock Condition) has been implemented

NCS temperature is now stable

NCS pressure is stable with only the control group of pressurizer heaters energized

Letdown has been restored The crew has determined that a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months soak is required. Which ONE (1) of the following evolutions could be performed by the crew in the next hour while continuing on through the EP procedures?

A. Start 1 D NCP.

B. Place auxiliary spray in service.

C. Energize Backup heaters.

D. Increase CA flow to one intact S/G to raise NR level to 50%.

Page 47 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 45 1 Pt Given the following conditions on Unit 1:

Unit is operating at 70% RTP

The OATC determines that AUTO rod withdrawal is not functioning

Further investigation reveals that manual rod withdrawal is functioning normally

Control Bank '0' rods are currently at 190 steps Which ONE (1) of the following failures has caused this condition?

A. Turbine Impulse Pressure Channel II fails low.

B. Turbine Impulse Pressure Channell fails low.

C. Loop 2 ~ T Channel fails high.

O. PR Channel N-41 fails high.

Page 48 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 46 1 Pt Given the following conditions on Unit 2:

The core has been off-loaded to the Spent Fuel Pool

2A KF Pump is running

2B KF Pump is off

A Loss of Off-Site Power occurs

2A and 2B DIGs start and load normally

30 minutes after the loss of power, a Spent Fuel Pool Hi Temperature alarm is received Which ONE (1) of the following is the cause of this condition?

A. The 2A or 2B KF pump was not manually restarted (sequencer reset required).

B. The 2A or 2B KF pump was not manually restarted (sequencer reset NOT required).

C. The 2A KF pump ONLY did not automatically load on the Blackout sequence as designed.

D. The 2A and 2B KF pumps did not automatically load during the Blackout sequence as designed.

Page 49 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 47 1 Pt Given the following conditions on Unit 1:

NC system temperature is 250°F

2 NC pumps are in operation Which ONE (1) of the following lists the Tech Spec 3.5.3 (ECCS - Shutdown) operability requirements based on the above conditions?

A. ONE Charging subsystem and ONE RHR subsystem must be OPERABLE.

B. BOTH Charging subsystems and BOTH RHR subsystems must be OPERABLE.

C. ONE Charging subsystem and BOTH RHR subsystems must be OPERABLE.

D. BOTH Charging subsystems and ONE RHR subsystem must be OPERABLE.

Page 50 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 48 1 Pt A LOCA has occurred on Unit 2 inside Containment. The following conditions are observed:

Containment pressure is 2.5 PSIG

Containment sump level is 11 feet

1EMF51A (Reactor Building Activity) indicates 40 R/Hr

Containment hydrogen concentration is 0.4%

Which ONE (1) of the following procedures should be implemented?

A. FR-Z.1 (Response to High Containment Pressure)

B. FR-Z.2 (Response to Containment Flooding)

C. FR-Z.3 (Response to High Containment Radiation Level)

D. FR-ZA (Response to High Containment Hydroge,n)

Page 51 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 49 1 Pt Given the following:

Unit 1 is operating at 100% RTP

Unit 2 is refueling

Unit 1 is releasing a minimally decayed Waste Gas Decay Tank

A significant packing leak starts on isolation valve 1WG-160, 0NG Decay Tank Outlet to Unit Vent Control)

Which ONE (1) of the following correctly describes the automatic actions which will ensure that the leak is contained and filtered?

A. OEMF-50 (Waste Gas Disch Hi Rad) automatically closes 1WG-160 AND 1EMF-41 (Aux Bldg Vent Hi Rad) automatically stops the Auxiliary Building ventilation unfiltered exhaust fans.

B. 1EMF-35 (Unit Vent Part Hi Rad) automatically stops the Auxiliary Building ventilation unfiltered exhaust fans AND 1EMF-41 (Aux Bldg Vent Hi Rad) automatically aligns the Auxiliary Building ventilation filter trains.

C. 1EMF-36 (Unit Vent Gas Hi Rad) automatically closes 1WG-160 AND 1EMF-35 (Unit Vent Part Hi Rad) automatically aligns the Auxiliary Building ventilation filter trains.

D. 1EMF-41 (Aux Bldg Vent Hi Rad) automatically stops the Auxiliary Building ventilation unfiltered exhaust fans AND 1EMF-36 (Unit Vent Gas Hi Rad) automatically closes 1WG-160.

Page 52 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 50 1 Pt Which of the following plant evolutions does NOT require that a 11M plot be performed in accordance with NSD 304 (Reactivity Management)?

A. Initial reactor startup via boron dilution B. Primary plant cooldown prior to a refueling outage C. Fuel Loading D. Reactor startup following a trip from 50% power Page 53 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 51 1 Pt Which ONE (1) of the following lists the setpoints for the design feature of the NO system which functions to protect the system from overpressure?

A. Suction relief - 450 PSIG Discharge relief - 650 PSIG B. Suction relief - 400 PSIG Discharge relief - 600 PSIG C. Suction relief - 450 PSIG Discharge relief - 600 PSIG D. Suction relief - 400 PSIG Discharge relief - 650 PSIG Page 54 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 52 1 Pt Given the following:

A VI header rupture occurs causing the VI system to completely depressurize.

Which ONE (1) of the following describes how the VS system is prevented from depressurizing due to the break on the VI system?

A. The VI and VS headers are separated using a manual isolation valve. However, VS air compressor will start automatically to maintain VS header pressure.

B. The VI and VS headers are separated using a manual isolation valve. And, the VS air compressor must be manually started to maintain VS header pressure.

C. A valve will automatically close to separate the VI and VS headers. And, the VS air compressor will start automatically to maintain VS header pressure.

D. A valve will automatically close to separate the VI and VS headers. However, the VS air compressor must be manually started to maintain VS header pressure.

Page 55 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 53 1 Pt Given the following conditions:

Refueling is in progress on Unit 1

Reactor vessel head is removed and the refueling cavity is full Which ONE (1) of the following states the requirements for NO Loop operability in accordance with Tech Spec 3.9.5 (Residual Heat Removal (RHR) and Coolant Circulation - High Water Level) for the conditions above?

A. ONE NO Loop must be OPERABLE and in operation. An NO Loop consists of one NO pump, one heat exchanger, and a flow path from an NC system COLO leg to an NC system HOT leg.

B. TWO NO Loops must be OPERABLE with one NO Loop in operation. An NO Loop consists of one NO pump, one heat exchanger, and a flow path from an NC system COLO leg to an NC system HOT leg.

C. ONE NO Loop must be OPERABLE and in operation. An NO Loop consists of one NO pump, one heat exchanger, and a flow path from an NC system HOT leg to an NC system COLO leg.

O. TWO NO Loops must be OPERABLE with one NO Loop in operation. An NO Loop consists of one NO pump, one heat exchanger, and a flow path from an NC system HOT leg to an NC system COLO leg.

Page 56 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 54 1 Pt Given the following conditions on Unit 1:

A small break LOCA has occurred inside Containment

E-1 (Loss of Reactor or Secondary Coolant) has been implemented

1ETA is de-energized

The 1B NI pump has failed

The 1B Hydrogen Recombiner is out of service

Containment hydrogen concentration is 7%

The TSC has recommended purging containment to reduce hydrogen concentration to 3.5% before energizing the igniters Which ONE (1) of the following statements correctly describes the method for performing this evolution to control the off-site dose?

A. Containment air is exhausted to the Annulus where it is filtered prior to release to the unit vent stack.

B. Containment air is exhausted to the Auxiliary Building where it is filtered prior to release to the unit vent stack.

C. Containment air is exhausted through the Incore Instrument Ventilation system where it is filtered prior to release to the unit vent stack.

D. Containment air is exhausted through the Containment Air Release system where it is filtered prior to release to the unit vent stack.

Page 57 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 55 1 Pt Given the following:

Unit 2 is operating at 90% RTP after a start-up from a refueling outage

A Pressurizer PORV is found to be leaking

The leaking PORV's block valve has been shut The PRT has been cooled down to the following current conditions:

PRT level - 65 %

PRT pressure - 9 PSIG

PRT temperature - 100°F

Lower Containment temperature - 118°F What actions are required in accordance with OP/2/A/6150/004 (Pressurizer Relief Tank) to restore and maintain normal operating conditions in the PRT for the long term?

A. Vent the PRT to containment.

B. Vent the PRT to the waste gas system.

C. Continue to cool the PRT by initiating spray flow from the NCDT.

D. Continue to cool the PRT by initiating spray flow from the RMWST.

Page 58 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 56 1 Pt Given the following initial conditions on Unit 2:

A unit shutdown and cooldown was in progress

All actions associated with NC system pressure going below P-11 were completed

'8' Train components are in service The following sequence of events occurs:

1. 120VAC Vital panel board EKVD de-energizes due to an electrical fault

2. 'A' Main Steam line ruptures

3. Pressurizer pressure is 1840 PSIG and going down

4. Containment pressure is 1.2 PSIG and going up Which ONE (1) of the following would get an automatic START signal? (Assume no operator actions have occurred.)

A All ECCS equipment

8. None of the ECCS equipment C. Only 'A' Train ECCS equipment D. Only '8' Train ECCS equipment Page 59 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 57 1 Pt Given the following:

Unit 2 is operating at 100% RTP

It has been determined that eight (8) Ice Condenser doors will not open due to floor buckling Which ONE (1) of the following statements describes the Containment design pressure AND how peak Containment pressure during a Design Basis Accident is affected by the condition above?

A. Containment design pressure is 3 psig. Peak pressure will be reached LATER than normal.

B. Containment design pressure is 3 psig. Peak pressure will be reached SOONER than normal.

C. Containment design pressure is 15 psig. Peak pressure will be reached LATER than normal.

D. Containment design pressure is 15 psig. Peak pressure will be reached SOONER than normal.

Page 60 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 58 1 Pt Given the following conditions:

A power ascension to 100% RTP is in progress on Unit 1

The power increase is currently on hold at 80% RTP

Power Range Channel N-43 fails high Narrow Range level for _ _ _ increases to the 100% setpoint level.

Which ONE (1) of the following best completes the statement above?

A. all four S/Gs

8. S/G 'C' ONLY C. S/G 'A' and S/G '0' ONLY O. S/G '8' and S/G 'C' ONLY Page 61 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 59 1 Pt Given the following conditions on Unit 1:

A loss of voltage has occurred on 1ETA

Blackout loading is in progress

A Safety Injection signal is received before Blackout loading is completed on 1ETA Which of the following describes the events that occur from the time the Safety Injection signal is received?

A. The Blackout load sequence is completed, 1ETA is cleared of all non-SI loads, and the SI load sequence is actuated.

B. The Blackout load sequence stops, 1ETA is cleared of all non-Slloads, and the Slload sequence is actuated.

C. The Blackout load sequence is completed, 1ETA is cleared of all loads, and the Slload sequence is actuated.

D. The Blackout load sequence stops, 1ETA is cleared of all loads, and the Slload sequence is actuated.

Page 62 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 60 1 Pt Given the following conditions on Unit 1 :

The unit is at 100% power.

The following annunciator is received in the Control Room:

o BATT EVCA GROUND Which ONE (1) of the following describes the indication available if the ground is on the positive leg of battery EVCA, and the resulting operation of the DC Bus?

A. Ground indication light on the back of the Main Control Board will be extinguished. Equipment malfunctions are NOT expected with a ground on one side.

B. Ground indication light on the back of the Main Control Board will be brightly lit. Equipment malfunctions are expected with a ground on either the positive OR negative terminal.

C. Ground indication light on the back of the Main Control Board will be brightly lit. Equipment malfunctions are NOT expected with a ground on one side.

D. Ground indication light on the back of the Main Control Board will be extinguished. Equipment malfunctions are expected with a ground on either the positive OR negative terminal.

Page 63 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 61 Given the following conditions:

The unit is in Mode 5

A loss of NO has occurred

The crew is performing the actions of AP-19 (LOSS OF NO OR NO SYSTEM LEAKAGE)

BOTH NO Pumps were tripped

The crew is preparing to start "1A" NO Pump Which ONE (1) of the following describes the required position of the 1A NO Heat Exchanger outlet isolation (1 ND-29) and HX bypass isolation valve (1 N 0-34) in accordance with AP-19?

1 ND-29 1ND-34 A. OPEN OPEN B. CLOSED OPEN C. OPEN CLOSED D. CLOSED CLOSED Page 64 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 62 1 Pt Which of the following will occur simultaneously with an automatic Containment Spray Actuation?

A. Phase A Isolation B. Feedwater Isolation C. Main Steam Isolation D. Containment Ventilation Isolation Page 65 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 63 1 Pt Given the following:

Unit 1 @ 100% RTP

CETs currently indicate 615°F

A malfunction of the Reference Junction Box temperature control causes the junction box temperature to INCREASE by 20°F Assuming that ACTUAL core-exit temperature remains constant, which ONE (1) of the following lists the new indication for the non-safety related CETs?

Page 66 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 64 1 Pt Given the following conditions:

Unit 1 is in Mode 4

NO system is providing core cooling

AP-22 (Loss of VI) has been implemented

NO system flow rate is being controlled by throttling:

1NI-173A (1A NO to A & B Cold Legs Cont Outside Isol) 1NI-178B (1B NO to C & 0 Cold Legs Cont Outside Isol)

Which ONE (1) of the following describes the requirements in AP-22 for operating the manual loaders for 1NO-14 and 1NO-29 (1 B/1A NO Hx Outlet Isolation)?

A. Turn the manual loaders 5-6 times in the clockwise direction to ensure that the valves close when VI pressure is returned to normal.

B. Turn the manual loaders 5-6 times in the counter-clockwise direction to ensure that the valves close when VI pressure is returned to normal.

C. Turn the manual loaders 5-6 times in the clockwise direction to ensure that the valves remain open when VI pressure is returned to normal.

O. Turn the manual loaders 5-6 times in the counter-clockwise direction to ensure that the valves remain open when VI pressure is returned to normal.

Page 67 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 65 1 Pt Which ONE (1) of the exposures would exceed a 10CFR20 annual exposure limit?

A. 45 REM to the thyroid B. 17 REM to the lens of the eye C. 45 REM to the leg below the knee D. 17 REM to the skin Page 68 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 66 1 Pt Given the following:

In preparation for a containment entry RP requested that a U-1 RO to align 1EMF 38, 39 & 40 to sample Upper Containment only

Due to a miscommunication from RP the request is never made to realign this sample point to normal

1EMF 38, 39 & 40 sample point remains aligned to upper containment ONLY over the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

A VQ release in is in progress Which ONE (1) of the following describes the effect on the 1EMF 38,39 &

40 sample package's ability to monitor changing radiation levels associated with conditions inside U-1 Containment?

A. Due to the limited amount of communication between atmosphere in upper and lower containment, 1EMF-38 will not be able to detect a 1 GPM NC system leak within one hour.

B. Because air is released from upper containment, this will have no effect on the ability of the affected EMF's to auto terminate a VQ release should radiation levels increase.

C. Prolonged operation of this EMF sample package aligned to Upper Containment ONLY could result in damage to the associate sample pump resulting in a degraded sample flowrate.

D. Due to the normal mixing of containment atmosphere by the ventilation system, this does not significantly effect the ability of this sample package to sample the containment atmosphere.

Page 69 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 67 1 Pt Given the following:

Unit 2 is operating at 100% RTP

'B' Train is in service

2A RN train is in operation for testing

The RN trains are split with 2RN-41 B (Train B to Non-Ess Hdr Isol) closed

2B RN pump flow is reading approximately 2700 GPM Subsequently, 2RN-190B (RN to B KC HX Control) is positioned to a value of 100%.

Which ONE (1) of the following correctly describes the component that will stabilize at a LOWER value?

A. Containment temperature B. NC pump stator temperatures C. NC pump bearing temperatures D. Letdown Heat Exchanger temperature Page 70 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 68 1 Pt Unit 1 is operating at 100% RTP. Given the following events and conditions:

2100 Power is lost to all Control Room annunciators 2120 The OSM declares an Unusual Event 2125 The Emergency Notification form is completed, and the OSM directs an operator to make the notification to the state and counties 2130 A reactor trip occurs on Unit 1 and the OSM declares a Site Area Emergency 2135 The OSM completes a new Emergency Notification Form and directs the notification of state and counties Which ONE (1) of the following indicates the time when the state and counties must be first notified of the events described above?

A. 2135 B. 2140 C. 2145 D. 2150 Page 71 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 69 1 Pt A worker needs to repack a valve in an area that has the following radiological characteristics:

The worker's present exposure is 800 mREM for the year

=

General area dose rate 65 mREM/hr

Airborne contamination concentration =20 DAC The job will take 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months with a mechanic wearing a full-face respirator. It will only take 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months if the mechanic does NOT wear the respirator.

Which of the following choices for completing this job would maintain the workers exposure within the Station ALARA requirements?

A. The worker should wear the respirator; otherwise he will exceed 25% of the DAC limit.

B. The worker should NOT wear the respirator because the total TEDE dose received will be greater if he wears one.

C. The worker should NOT wear the respirator because the airborne contamination is low enough that a respirator is not required.

D. The worker should wear the respirator because the total TEDE dose received will be less than if he does not wear one.

Page 72 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 70 1 Pt The following conditions exist on Unit 1:

Reactor Power is 50%

A momentary loss of power occurs on EKVA causing the following instruments to fail low:

First Stage Turbine Impulse Pressure Channel 1

Power Range Channel N-41 Which ONE (1) of the following correctly completes the statement below?

Tref on the Tave / Tref recorder indicates ...

A. 585°F but control rods do not move.

B. 55rF but the control rods do not move.

C. 585°F and the control rods step out at 72 steps per minute.

D. 55rF and the control rods step in at 72 steps per minute.

Page 73 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 71 1 Pt Plant conditions:

Both units are at 100% power

Maintenance has requested an opportunity to do troubleshooting on Battery EVCA

An Electronic Risk Assessment Tool (ORAM) assessment has resulted in a WHITE status on this activity Which ONE (1) of the following identifies how this maintenance activity should be handled?

A. Allow the maintenance to be performed; The risk assessment is minimal.

B. Allow the maintenance to be performed; The risk assessment is acceptable.

C. Do NOT allow the maintenance to be performed; The risk assessment is high.

D. Do NOT allow the maintenance to be performed; ORAM cannot determine the risk level.

Page 74 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 72 1 Pt Unit 1 is being returned to full power following a refueling outage when the following annunciators are received in the Control Room:

"S/G A FLOW MISMATCH LO STM FLOW" (1AD4-A1)

"LOOP OfT DEV" (1AD6-E10)

"A NC PUMP LO FLO ALERT" (1AD6-F1)

The Safety breaker AND the Feeder breaker for 1A NC pump indicate CLOSED

Reactor power remains at 55% RTP Based on the above conditions, which ONE (1) of the following actions is required to be performed?

A. Reduce reactor power to less than 48% and trip 1A NC pump.

B. Manually trip the reactor and go to E-O, Reactor Trip or Safety Injection.

C. Select Delta-T defeat for the affected loop per the Annunciator Response Procedure.

D. Place the A S/G Feed Reg Valve in Manual and reduce feed flow.

Page 75 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 73 1 Pt The following conditions exist:

Both units are operating at 100% RTP

The following alarms are received on both units:

o 1AD-11 C-5 (XFMR A URGENT ALARM) o 2AD-11 C-5 (XFMR A URGENT ALARM)

Each alarm is the result of a loss of BOTH Cooling Groups to the respective transformer Which ONE (1) of the following describes the difference in the Operator's response to these two annunciators?

A. The 1A Main Transformer cooling must be restored within 8 min 45 sec in order to prevent a turbine runback to <56%.

B. The 1A Main Transformer cooling must be restored within 28 min 45 sec in order to prevent a turbine runback to <56%.

C. The 2A Main Transformer cooling must be restored within 8 min 45 sec in order to prevent a turbine runback to <56%.

D The 2A Main Transformer cooling must be restored within 28 min 45 sec in order to prevent a turbine runback to <56%.

Page 76 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 74 1 Pt Given the following conditions on Unit 1:

A LOCA has occurred

Safety Injection has been initiated Which ONE (1) of the following describes the operation of the Containment Cooling system based on these conditions?

A. All VU units have started and RV containment isolation valves are open.

B. All VU units have started and RV containment isolation valves are closed.

C. All VU units have shunt tripped off and RV containment isolation valves are open.

D. All VU units have shunt tripped off and RV containment isolation valves are closed.

Page 77 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 75 1 Pt Unit 1 is operating at 100% RTP. Given the following events and conditions:

In preparation for a Unit 1 TO CA pump performance test the following valves are positioned as indicated:

o 1CA-64AB (TO CA Pump To 1A S/G) -- CLOSED o 1CA-52AB (TO CA Pump To 1B S/G) -- CLOSED o 1CA-48AB (TO CA Pump To 1C S/G) -- OPEN o 1CA-36AB (TO CA Pump To 10 S/G) -- OPEN

An air line supplying 1SA-48ABC (SM FRM S/G 1C TO TO CA PUMP ISOL) ruptures resulting in a loss of air to the valve Which ONE (1) of the following describes the impact to the Unit 1 CA system AND what actions are required by the operating crew, in accordance with OPS MANAGEMENT EXPECTATIONS, to mitigate the consequences of this event?

A. ALL flow control valves OPEN:

Crew should close all TO CA flow control valves as soon as practical.

B. ALL flow control valves OPEN:

Crew should delay closing all TO CA flow control valves until appropriate procedures have been implemented.

C. Flow control valves remain AS IS:

Crew should close all TO CA flow control valves as soon as practical.

D. Flow control valves remain AS IS:

Crew should delay closing all TO CA flow control valves until appropriate procedures have been implemented.

Page 78 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 76 1 Pt Given the following:

Unit 2 is operating at 45% RTP

A failure of the RC system has occurred resulting in a severe reduction in RC flow to the Main Condenser

Main Condenser vacuum is at 21" Hg and degrading and the crew has entered AP-23 (Loss of Condenser Vacuum)

Annunciator 2AD-1, F8 (TURB EXH HOOD HI-HI TEMP) alarms Which ONE (1) of the following actions would be required to mitigate the current plant conditions?

A. Reduce Turbine Generator load and start additional RC pumps per AP-23.

B. Reduce Turbine Generator load and dispatch operators to start Main Vacuum pumps per AP-23.

C. Trip the turbine, transition to AP-02 (Main Turbine Trip), stabilize reactor power at 12-15%.

D. Trip the reactor, transition to E-O (Reactor Trip or Safety Injection).

Page 79 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 77 1 Pt Given the following:

Unit 1 is operating at 100% RTP

Train 'B" SSPS testing is in progress In accordance with OMP 5-3 (TSAIL) entry into Tech Spec 3.3.1 (RTS Instrumentation) for "One Rx Trip Breaker Train Inoperable" must be made when the (1) and if the RTA Bypass (BYA) breaker is racked in and closed while RTB Bypass (BYB) breaker is racked in and closed, (2) would receive a trip open signal.

Which ONE (1) of the following correctly completes the statement above?

A. (1 ) BYB breaker is racked in ONLY (2) RTA and RTB breakers ONLY B. (1 ) BYB breaker is racked in and closed (2) RTA and RTB breakers ONLY C. (1 ) BYB breaker is racked in ONLY (2) RTA, RTB, BYA, and BYB breakers D. (1) BYB breaker is racked in and closed (2) RTA, RTB, BYA, and BYB breakers Page 80 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator*

Question: 78 1 pt Unit 1 & 2 are operating at 100% RTP.

The TCC has notified the Control Room that the "Real Time Contingency Analysis" (RTCA) indicates that switchyard voltage would not be adequate should a Unit Trip occur

AP-05, (Generator Voltage and Electric Grid Disturbances) has been implemented on BOTH units Concerning Unit 1, which ONE (1) of the following is the reason why Tech Spec 3.0.3 is entered, AND the reason why Tech Spec 3.8.1 (AC Sources - Operating) is entered?

A. Both DIGs Load Sequencers being Inoperable, AND Both DIGs being Inoperable.

B. Both DIGs Load Sequencers being Inoperable, AND Both Offsite Power Sources being Inoperable.

C. Both ECCS trains being Inoperable, AND Both Offsite Power Sources being Inoperable.

D. Both ECCS trains being Inoperable, AND Both DIGs being Inoperable.

Page 81 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 79 1 Pt Given the following conditions:

Unit 1 is operating at 75% RTP when a dropped rod occurs

The crew is performing the actions of AP-14 (Rod Control Malfunction) when it is noted that AFD has exceeded the limits of Tech Specs

The dropped rod has not yet been recovered Which ONE (1) of the following describes the required actions in accordance with AP-14 and the reason for those actions?

A. Transition to AP-4 (Rapid Downpower) and reduce power to less than 50% to ensure that the fuel is not damaged due to exceeding Radial peaking factors.

B. Initiate a reactor trip and transition to E-O (Reactor Trip or Safety Injection) to ensure that the fuel is not damaged due to exceeding Radial peaking factors.

C. Transition to AP-4 (Rapid Downpower) and reduce power to less than 50% to ensure that the fuel is not damaged due to exceeding Axial peaking factors.

D. Initiate a reactor trip and transition to E-O (Reactor Trip or Safety Injection) to ensure that the fuel is not damaged due to exceeding Axial peaking factors.

Page 82 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 80 1 Pt Unit 1 is operating at 100% RTP with 'B' train components in service with a normal RN system alignment. 1B RN pump amps are swinging.

The following annunciators are in alarm:

"B RN Pump Suction Lo Press"

"B RN Pump Discharge Lo Press" Which ONE (1) of the following is the correct response to the above conditions based on the implementation of AP-20 (Loss of RN)?

A. Implement Case 1, Loss of Operating RN Train Place the '1A' RN pump in service and, Remain on Low Level Intake B. Implement Case 1, Loss of Operating RN Train Swap alignment to the Nuclear Service Water Pond and, Place the '1A' RN pump in service C. Implement Case 2, Loss of Low level or RC Supply Crossover Place the '1A' RN pump in service and, Remain on Low Level Intake D. Implement Case 2, Loss of Low level or RC Supply Crossover Swap alignment to the Nuclear Service Water Pond and, Place the '1A' RN pump in service Page 83 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 81 1 Pt With regards to the analysis for a Design Basis Containment Flooding event, which of the following describes (1) the basis for the maximum Containment sump level AND (2) the sources of water which contribute to the maximum sump level?

A. 1) Sufficient volume to prevent exceeding containment design pressure in the event of a hydrogen ignition event.

2) The entire contents of the NC system, FWST, NI Accumulators and CA Storage Tank.

B. 1) Ensures that vital equipment needed for the safe shutdown of the plant remain available.

2) The entire contents of the NC system, FWST, NI Accumulators and CA Storage Tank.

C. 1) Sufficient volume to prevent exceeding containment design pressure in the event of a hydrogen ignition event.

2) The entire contents of the NC system, FWST, and NI Accumulators ONLY.

D. 1) Ensures that vital equipment needed for the safe shutdown of the plant remain available.

2) The entire contents of the NC system, FWST, and NI Accumulators ONLY.

Page 84 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 82 1 Pt Given the following conditions on Unit 1:

A Loss of Offsite Power has occurred resulting in a partially stuck open PZR Safety and subsequent Safety Injection

The 1B D/G failed to start

Actions in E-O (Reactor Trip or Safety Injection) are complete

NC pressure is 1585 PSIG

RVLlS Lower Range level is 45% and decreasing

1NI-9A (NC Cold Leg Inj from NV) is closed and cannot be opened

CETs are reading 705°F

Subcooling Monitor is indicating -35°F Which ONE (1) of the following indicates both the correct procedure and associated strategy which will result in RECOVERY of NC system inventory?

A. Go to FR-C.1 (Response to Inadequate Core Cooling) and establish high head injection via the Standby M/U Pump and the PD pump.

B. Go to FR-C.2 (Response to Degraded Core Cooling) and establish high head injection via the Standby M/U Pump and the PD pump.

C. Go to FR-C.1 (Response to Inadequate Core Cooling) and dump steam from intact S/Gs to allow CLAs and/or ND to recover core cooling.

D. Go to FR-C.2 (Response to Degraded Core Cooling) and dump steam from intact S/Gs to allow CLAs and/or ND to recover core cooling.

Page 85 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 83 1 Pt Unit 1 is operating at 100% RTP The following indications are observed:

NC PUMP SEAL INJ LO FLOW alarm

CHARGING LINE ABNORMAL FLOW alarm

REGEN HX LETON HI TEMP alarm

NC pump lower bearing temperatures are slowly increasing Based on the above indications, the crew will enter (1) and the mitigating strategy for this procedure will require the operators to (2)

Which ONE (1) of the following correctly completes the statement above?

A (1) AP-8 (Malfunction of NC Pump)

(2) start the PO pump B. (1) AP-12 (Loss of Letdown, Charging, or Seal Injection)

(2) start the PO pump C. (1) AP-8 (Malfunction of NC Pump)

(2) operate 1NV-243 (Regen HX Tube Inlet Control Bypass) as necessary O. (1) AP-12 (Loss of Letdown, Charging, or Seal Injection)

(2) operate 1NV-243 (Regen HX Tube Inlet Control Bypass) as necessary Page 86 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 84 1 Pt Given the following:

Unit 1 has experienced a Loss of All AC Power

ECA-O.O (Loss of All AC Power) is in effect

The attempted start of the Standby Make-up Pump was unsuccessful

In accordance with ECA-O.O the crew is preparing to depressurize the intact S/Gs to 290 PSIG Which ONE (1) of the following describes the procedural requirements for depressurizing the S/Gs?

A. S/Gs should be depressurized at the maximum rate. S/G depressurization should NOT be stopped if reactor vessel head voiding occurs.

B. S/Gs should be depressurize to establish a cooldown rate close to 100°F / hr. S/G depressurization should NOT be stopped if reactor vessel head voiding occurs.

C. S/Gs should be depressurized at the maximum rate. S/G depressurization should ONLY be stopped if reactor vessel head voiding occurs.

D. S/Gs should be depressurize to establish a cooldown rate close to 100°F / hr. S/G depressurization should ONLY be stopped if reactor vessel head voiding occurs.

Page 87 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 85 1 Pt Given the following conditions:

Unit 1 is operating at 100% power

AP-12 (Loss of Letdown, Charging, or Seal Injection) has been implemented due to fluctuating charging flow indication and 1B NV pump (the running pump) has been stopped

It is discovered that the VCT N2 pressure regulator has failed resulting in a high VCT pressure

The regulator has been replaced and VCT pressure returned to normal

Pressurizer level is currently at 25% and decreasing at a rate of 1% / min

Operators in the field are performing AP-12 Enclosure 1 (NV Pump Venting) and report that it will take at least 15 minutes to complete Which ONE (1) of the following is the required action per AP-12 for the conditions listed above?

A. Start the Standby Makeup pump per Enclosure 3 (Standby Makeup Pump Startup).

B. WHEN pressurizer level decreases to less than 17%, THEN start the Standby Makeup pump per Enclosure 3 (Standby Makeup Pump Startup).

C. Complete the actions of Enclosure 1 and then start an NV pump.

D. WHEN pressurizer level decreases to less than 17%, THEN stop performing Enclosure 1 and IMMEDIATELY start 1A NV pump.

Page 88 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 86 1 Pt Unit 1 is operating at 100% RTP.

Under which ONE (1) of the following conditions mayan active licensed STA assume the duties of the Control Room Supervisor?

The CRS or relief SRO is available to return to the control room within (1) AND the periods during which the STA assumes SRO duties do not exceed (2) in duration.

A. (1) 15 minutes (2) 15 minutes B. (1) 15 minutes (2) 10 minutes C. (1) 10 minutes (2) 15 minutes D. (1) 10 minutes (2) 10 minutes Page 89 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 87 1 Pt Given the following:

Unit 2 is responding to a large LOCA with 10 % failed fuel

2NI-1848 (18 NO Pump Suction From Cont Sump Isol) would not open during FWST swapover

2NI-1848 breaker has tripped and will not reset

The OSM has determined that manual alignment of 2NI-1848 is required to protect the health and safety of the general public

RP projects that expected dose rates in the area of the valve may exceed 150 REM/hr Which ONE (1) of the following TEOE exposure limits would apply to a worker who volunteered to manually open 2N 1-1848 in accordance with RP/004 General Emergency?

A. The worker( s) may exceed 25 REM

8. Do not exceed 25 REM C. Do not exceed 10 REM O. Do not exceed 5 REM Page 90 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 88 1 Pt Per RP-1 0 (NRC Immediate Notification Requirements) which ONE (1) of the following is an event requiring IMMEDIATE notification to the NRC?

A. Any event that may have caused an exposure of greater than or equal to 70 REM EDE.

B. Any event that threatens to cause an exposure of greater than or equal to 50 REM CEDE.

C. Any event that may have caused or threatens to cause an exposure of greater than or equal to 25 REM TEDE.

D. Any event that may have caused or threatens to cause an exposure of greater than or equal to 200 RAD SDE to the skin or extremities.

Page 91 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 89 1 Pt Given the following conditions:

45 minutes ago a large break LOCA occurred on Unit 1

Containment pressure indicates 9 PSIG

Containment hydrogen concentration is 1%

CETs indicate temperatures of 1100°F

RVLlS lower range level indicates 40%

EMF 51A indicates 185 R/hr

Subcooling margin indicates -35°F

All S/G NR level are off scale low with no CA flow indicated Which ONE (1) of the following is the correct classification and associated EAL Number for this event?

REFERENCES PROVIDED A. Site Area Emergency based on EAL # 4.1.S.2 B. General Emergency based on EAL # 4.1.G.2 C. Site Area Emergency based on EAL # 4.1.S.1 D. General Emergency based on EAL # 4.1.G.1 Page 92 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 90 1 Pt Which ONE (1) of the following describes the process for notating a Temporary Modification on a flow diagram or electrical one-line drawing in accordance with OMP 10-2, Temporary Modifications?

A. Affected drawings are notated with a pink cloud to refer the operator to the TM package.

B. Flow diagrams and electrical one line drawings SHALL always be red-marked to reflect the temporary modification.

C. Flow diagrams SHALL be red-marked to reflect the temporary modification; electrical one line drawings shall NOT be red marked to reflect the temporary modification.

D. Drawings in WCC office may be red-marked at the SRO's discretion, but the drawings in the Control Room and Tagout office must be changed by the responsible engineer.

Page 93 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 91 1 Pt Given the following conditions:

Unit 1 is in a refueling outage

Core unload is in progress

The Spent Fuel Pool Level Low computer alarm has actuated

Actual Spent Fuel Pool level is minus 2 feet and slowly going down

1EMF-17 (Spent Fuel Bldg. Refueling Bridge) and 1EMF-42 (Fuel Building Radiation Monitor) are in the trip 1 condition

Unit 1 "INCORE SUMP INST ROOM SUMP HI LEVEL" alarm is received Which ONE (1) of the following statements describes the mitigating actions for this event?

A. Implement AP-40, Loss of Refueling Cavity Level, and stop the KF pumps and ensure that VF has been placed in filter mode.

B. Implement AP-41, Loss of Spent Fuel Pool Cooling and Level, and stop the KF pumps and ensure that VF has been placed in filter mode.

C. Implement AP-40, Loss of Refueling Cavity Level, move the fuel transfer cart to the Refueling Canal side and have maintenance install the weir gate.

D. Implement AP-41, Loss of Spent Fuel Pool Cooling and Level, move the fuel transfer cart to the Refueling Canal side and have maintenance install the weir gate.

Page 94 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 92 1 Pt Given the following:

Unit 1 was operating at 100% RTP

While performing maintenance activities in the 1A D/G Room, one of the heat detectors associated with the Halon system was inadvertently actuated

The Halon actuation was successfully aborted by the fire watch depressing the ABORT/OFF pushbutton prior to actual Halon discharge Which ONE (1) of the following describes how the 1A D/G Halon system is affected AND what actions are required per SLC 16.9.3 (Halon Systems)?

A. Auto actuation ONLY is blocked Establish an Hourly fire watch within one hour B. BOTH Manual and Auto actuation of the system is blocked.

Establish a Hourly fire watch within one hour.

C. Auto actuation ONLY is blocked Establish a Continuous Fire Watch within one hour.

D. BOTH Manual and Auto actuation of the system is blocked.

Establish a Continuous Fire Watch within one hour.

Page 95 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 93 1 Pt Given the following conditions:

A large break LOCA has occurred on Unit 1

Containment pressure is 11 psig

FWST level is below the swap over setpoint

ECA-1.1 (Loss of Emergency Coolant Recirculation) is implemented Which ONE (1) of the following describes the priority for NS pump operation in this condition and the basis for that operation?

A. ONLY one NS pump is required to be running during ECA-1.1 even if containment pressure red path criteria is exceeded.

B. Both NS pumps are required to be running because FR-Z.1 was implemented in response to an orange path and FRPs have priority over ECA procedures.

C. ONLY one NS pump is required to be running because this conserves FWST water level while providing sufficient NS flow to reduce containment pressure.

O. Both NS pumps are required to be running because a total loss of NO causes the NS system to become relatively more important in reducing containment pressure.

Page 96 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 94 1 Pt Given the following conditions:

Unit 1 is operating at 25% RTP

Reactor Power input for 1EMF-71 thru 74 (Unit 1 Steam Line N-16 Monitors) has failed to 100% power.

Based on the conditions above, the alarm setpoints for 1EMF-71 thru 74 are (1) for current plant conditions. Per SLC 16.7.6 (Radiation Monitoring for Plant Operations) basis, the MINIMUM sensitivity required to ensure that the monitors remain OPERABLE is (2)

A. (1 ) lower than normal (2) 30 GPO B. (1 ) lower than normal (2) 5 GPO C. (1 ) higher than normal (2) 30 GPO O. (1 ) higher than normal (2) 5GPO Page 97 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 95 1 Pt Given the following conditions:

Unit 1 is at 100% RTP.

Battery EVCB is aligned for an equalize charge.

Battery Room Exhaust fans have become inoperable due to low fan capacity.

It was determined the measured air flows through the battery rooms are below required values.

Which ONE (1) of the following correctly describes the action(s) according to SLC 16.9.23 (Control Room Area Ventilation System) AND the SLC basis for those actions?

Immediately suspend the EVCB equalize charge ...

A. ONLY.

This is done to prevent H2 concentration from exceeding the 4%

flammability limit.

B. ONLY.

This is done to prevent H2 concentration from exceeding the 2%

volume limit in the battery room.

C. AND Lock open all BR-XF Check Dampers.

This is done to prevent H2 concentration from exceeding the 4%

flammability limit.

D. AND Lock open all BR-XF Check Dampers.

This is done to prevent H2 concentration from exceeding the 2%

volume limit in the battery room.

Page 98 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 96 1 Pt Given the following initial conditions:

Unit 1 is in Mode 6

NC system is mid-loop at +10"

Both NO trains are in operation

NC system temperature is 190°F

NO system flow rate is 2000 GPM Current (amps) indication for both NO pumps begins to fluctuate.

Which ONE (1) of the following describes the condition that is causing the abnormal indications and the actions taken in accordance with AP-19 (Loss of NO or NO System Leakage) to mitigate the problem?

A. NO pumps are cavitating.

Reduce the NO system flow rate.

B. NO pumps are cavitating.

Stop both NO pumps.

C. NO pumps are operating at a runout condition.

Reduce NO system flow rate.

O. NO pumps are operating at a runout condition.

Stop both NO pumps.

Page 99 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 97 1 Pt Given the following:

The VI system on Unit 1 has become heavily contaminated with oil due to a maintenance problem

The VI Air Dryer packages rapidly clog Which ONE (1) of the following describes the impact on VI system operation AND the actions directed by AP-22 (Loss if VI) to mitigate the consequences of the malfunction?

VI header pressure will decrease untiL ..

A. 1VI-1812 (VI Air Dryer Bypass Filter Isol) automatically opens at 85 PSIG. If the valve fails to open, Enclosure 5 (VI Dryer and VI to VS System Isolation) directs the operators to bypass the air dryers locally at 82 PSIG.

B. 1VI-1812 (VI Air Dryer Bypass Filter Isol) automatically opens at 90 PSIG. If the valve fails to open, Enclosure 5 (VI Dryer and VI to VS System Isolation) directs the operators to bypass the air dryers locally at 85 PSIG.

C. 1VI-820 (VI to VS Supply) automatically closes at 85 PSIG. As a backup to the automatic action, AP-22 directs the operators to manually close 1VI-820 at 82 PSIG.

D. 1VI-820 (VI to VS Supply) automatically closes at 90 PSIG. As a backup to the automatic action, AP-22 directs the operators to manually close 1VI-820 at 85 PSIG.

Page 100 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 98 1 Pt Given the following initial conditions:

Unit 2 was operating at 100% RTP

Pressurizer Pressure Control switch is in the '1-2' position Subsequently, the following conditions are observed:

NC pressure increases

Pressurizer PORVs 2NC-32B and 2NC-36B open at 2335 PSIG

Pressure modulates between 2315 PSIG and 2335 PSIG Which ONE (1) the following instrument failures would cause the conditions observed above AND if the Pressurizer Pressure Control switch remains in the '1-2' position what is current status of 2NC-34A?

A. PZR pressure channel I fails LOW.

2NC-34A is INOPERABLE.

B. PZR pressure channell fails HIGH.

2NC-34A is INOPERABLE.

C. PZR pressure channel I fails LOW.

2NC-34A is OPERABLE.

D. PZR pressure channell fails HIGH.

2NC-34A is OPERABLE.

Page 101 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 99 1 Pt In accordance with OP/1/A/6400/006 (Nuclear Service Water System),

Enc. 4.4 (Filling the SNSWP), which ONE (1) of the following describes the preferred source of makeup to the SNSWP during normal operations and the basis for it being the preferred source?

A. 'A' Train RN because 'B' Train aligned to the SNSWP makes the Unit 1 SSF inoperable.

B. 'B' Train RN because 'A' Train aligned to the SNSWP makes the Unit 1 SSF inoperable.

C. 'A' Train RN because it maintains train separation.

D. 'B' Train RN because it maintains train separation.

Page 102 of 103

U.S.N.R.C. Site-Specific Written Examination McGuire Senior Reactor Operator Question: 100 1 Pt Both Units are operating at 100% RTP.

Loss of Offsite Power occurred on Unit 1

Both DGs started and loaded as designed

1A DIG subsequently trips on overspeed

At Step 17 of ES 0.1 (Reactor Trip Response), the decision is made to implement AP-07 (Loss of Electrical Power)

Which ONE (1) of the following correctly describes the Time Critical local operator actions associated with AP-07?

A. Implement Enc. 7 (DC Bus Alignment) to realign Battery Charger EVCA to Unit 2 within one hour.

B. Implement Enc. 7 (DC Bus Alignment) to realign Battery EVCA to Battery Charger EVCS within one hour.

C. Implement Generic Enc. 13 (VC and VA System Operation) to restart the Train A VCNC Chiller within 37.5 minutes.

D. Implement Generic Enc. 13 (VC and VA System Operation) to swap Train A VCIYC Chiller power and water to Unit 2 and restart chiller within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 15 minutes.

Page 103 of 103

McGuire Nuclear Station 2009 Written Examination Answer Key Question Answer Question Answer Question Answer Question Answer 1 c 26 c 51 c 76 o 2 c 27 o 52 o 77 o 3 o 28 B 53 C 78 c 4 B 29 C 54 A 79 o 5 A 30 A 55 B 80 o 6 o 31 C 56 C 81 B 7 B 32 A 57 o 82 o 8 B 33 A 58 C 83 B 9 A 34 C 59 B 84 B 10 C 35 A 60 C 85 A 11 C 36 A 61 o 86 C 12 C 37 B 62 C 87 A 13 C 38 o 63 o 88 C 14 A 39 A 64 C 89 B 15 B 40 B 65 B 90 B 16 C 41 B 66 A 91 A 17 A 42 C 67 C 92 C 18 o 43 C 68 A 93 C 19 B 44 B 69 B 94 C 20 C 45 B 70 o 95 o 21 C 46 B 71 o 96 B 22 C 47 A 72 B 97 A 23 B 48 C 73 C 98 C 24 o 49 B 74 C 99 B 25 o 50 B 75 C 100 A

2009 SRO License Exam References Table of Contents The following references are provided:

Copy of the Steam Tables

RP/OIA/57001000

U-I Data Book Curve 7.38

U-I Data Book Curve 1. lOB

OP/1/A161 00/22 UNIT 1 ENCLOSURE 4.3 CURVE 7.38 PRESSURIZER (VOLUME VS. TANK LEVEL) 16000 I I I I I

I TOTAL TANK VOLUME =13777 GALLONS I 14000

, i..-"

~

, JIll'

~

12000 ~

~

V Ui 10000 z

o..J

..J t,....- I"

~ 8000 , ",. I-"

w

E .., ,. ",.

J o

..J

> 6000 ~ "

,. ~

~ "

4000 ,. V

, i..'

JIll'

, l/

..... ~

2000 .., ~

..... ~

a a 10 20 30 40 50 60 70 80 90 100 LEVEL (PERCENT)

This data is also available on the OAC.

UNIT 1

OP/1/A/61 00/22 ENCLOSURE 4.3 CURVE 1.10B SATURATION CURVE ADJUSTED FOR INSTRUMENT ERROR (WIDE RANGE NC PRESSURE SENSOR)

SATURATION TEMP.

ADJUSTED FOR NC PRESSURE (psig) INSTRUMENT ERROR (OF) 0.0 193.90 35.3 207.84 85.3 228.15 135.3 299.73 185.3 336.80 235.3 363.07 285.3 383.81 335.3 401.13 385.3 416.09 435.3 429.35 485.3 441.28 535.3 452.17 585.3 462.20 635.3 471.53 685.3 480.25 735.3 488.45 785.3 496.21 885.3 510.56 985.3 523.66 1085.3 535.73 1185.3 546.93 1285.3 557.41

-1385.3 567.26 1485.3 576.56 1585.3 585.37 1685.3 593.76 1785.3 601.75 1885.3 609.39 1985.3 616.71 2085.3 623.74 2185.3 630.50 2285.3 637.00 2385.3 643.27 2485.3 649.32

References:

1) Calculation File MCC-1552.08-00-0160, Instrument Uncertainties for ICCM and OAC Subcooling Margin, Rev. 1.

2) MCEI-0400-08, ICCM and OAC Error Adjusted Saturation Tables, Rev. O.

3) IP/O/A/3000/18, ICCM-86 Programming and Operation, approved 12/3/91.

UNITl OP/1/A161 00/22 ENCLOSURE 4.3 - CURVE 1.1 OB SATURATION CURVE ADJUSTED FOR INSTRUMENT ERROR (WIDE RANGE NC PRESSURE SENSOR) 3000 2500 I I

.. I 2000

-0)

'iii II

-c.

~

SUBCOOLED I I

~ 1500

~ I D-O Z / SATURATED I'

1000

/

500 o .... "

- i""'"

0.00 100.00 200.00 300.00 400.00 500.00 600.00 700.00 Temperature (degrees F)

UNITl

Duke Energy Procedure No.

McGuire Nuclear Station 10 RP 1A/5700/000 Classification Of Emergency Revision No.

014 Electronic Reference No.

MCOO48M3 Reference Use PERFORMANCE I

* * * * * * * *

UNCONTROLLED FOR PRINT * * * * * * * * * *

(ISSUED) - PDF Format

RP/O/A/57001000 Page 2 of3 Classification of Emergency

1. Symptoms 1.1 Notification of Unusual Event 1.1.1 Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

1.1.2 No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

1.2 Alert 1.2.1 Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.

1.2.2 Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

1.3 Site Area Emergency 1.3.1 Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public.

1.3.2 Any releases are not expected to exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

1.4 General Emergency 1.4.1 Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.

1.4.2 Releases can be reasonably expected to exceed EPA Protective Action Guidelines exposure levels offsite for more than the immediate site area.

2. Immediate Actions 2.1 Determine operating mode that existed at the time the event occurred prior to any protection system or operator action initiated in response of the event.

RP/O/Al5700/000 Page 3 of3 2.2 IF the plant was in Mode 1-4 and a valid condition affects fission product barriers, THEN proceed to Enclosure 4.1 (Fission Product Barrier Matrix).

2.3 IF a General Emergency is NOT declared in Step 2.2, OR the condition does not affect fission product barriers, THEN review the listing of enclosures to determine if the event is applicable to one of the categories shown.

2.4 Compare actual plant conditions to the Emergency Action Levels evaluated in 2.2 andlor 2.3, then declare the appropriate Emergency Class as indicated.

2.4.1 Event Declaration time --------------------

2.5 Implement the applicable Emergency Response Procedure (RP) for that classification and continue with subsequent steps of this procedure.

Notification of Unusual Event RP/O/Al5700/001 Alert RP/O/Al5700/002 Site Area Emergency RP/O/Al5700/003 General Emergency RP/O/Al5700/004.

3. Subsequent Actions 3.1 To escalate, de-escalate, or terminate the Emergency, compare plant conditions to the Initiating Conditions of Enclosures 4.1 through 4.7.

3.2 Refer to enclosure 4.9, Emergency Declaration Guidelines, as needed.

3.3 Refer to section D of the McGuire EPLAN as the basis document for classification of emergencies as needed.

4.0 Enclosures 4.1 Fission Product Barrier Matrix 4.2 System Malfunctions 4.3 Abnormal Rad LevelslRadiological Effluent 4.4 Loss of Shutdown Functions 4.5 Loss of Power 4.6 FirelExplosion and Security Events 4.7 Natural Disasters, Hazards and Other Conditions Affecting Plant Safety 4.8 Definitions/Acronyms 4.9 Emergency Declaration Guidelines 4.10 Radiation Monitor Readings for Enclosure 4.3 EALs 4.11 Commitment Reference for Emergency Action Levels

Enclosure 4.1 RP/O/A/57001000 Fission Product Barrier Matrix Page 1 of5 Use EALs to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all 3 barriers. ClassifY according to the table on page 2 of 5 of this enclosure.

Note 1: This table is only applicable in Modes 1-4.

Note 2: Also, an event (or multiple events) could occur which results in the conclusion that exceeding the Loss or Potential Loss thresholds is imminent (i.e., within 1-3 hours). In this imminent loss situation, use judgement and classify as if the thresholds are exceeded.

Note 3: When determining Fission Product Barrier status, the Fuel Clad Barrier should be considered to be lost or potentially lost if the conditions for the Fuel Clad Barrier loss or potential loss EALs were met previously (validated and sustained) during the event, even if the conditions do not currently exist.

Note 4: Critical Safety Function (CSF) indications are not meant to include transient alarm conditions which may appear during the start-up of engineered safeguards equipment. A CSF condition is satisfied when the alarmed state is valid and sustained. The STA should be consulted to affirm if any CSF has been validated prior to that CSF being used as the basis to classify an emergency. {I} Example: If ECA-O.O, Loss of All AC Power, is implemented with an appropriate CSF alarm condition valid and sustained, that CSF should be used as the basis to classify an emergency function restoration orocedure being- implemented within the confines ofECA-O.O.

EAL# Unusual Event EAL# Alert EAL# Site Area Emergency EAL# General Emergency 4.l.U.l Potential Loss of 4. 1.A. 1 Loss OR Potential Loss 4.l.S.l Loss OR Potential Loss 4.l.G.l Loss of All Three Barriers Containment of of Both Nuclear Coolant System Nuclear Coolant System AND Fuel Clad 4.1.U.2 Loss of Containment 4.l.A.2 Loss OR Potential Loss 4.l.S.2 Loss 4.l.G.2 Loss of Any Two Barriers of AND AND Fuel Clad Potential Loss Potential Loss of the Third Combinations of Both Nuclear Coolant System AND Fuel Clad 4.l.A.3 Potential Loss of 4.l.S.3 Loss of Containment Containment AND AND Loss OR Potential Loss Loss OR Potential Loss of Any Other Barrier of Any Other Barrier I

Enclosure 4.1 RP/O/A/5700/o00 Fission Product Barrier Matrix Page 2 of5 NOTE: If a barrier is affected, it has a single point value based on a "potential loss" or a "loss". "Not Applicable" is included in the matrix as a place holder only, and has no point value assigned.

Barrier Points (1-5) Potential Loss (X) Loss (X) Total Points Classification Containment 1 3 1-3 Unusual Event NCS 4 5 4-6 Alert Fuel Clad 4 5 7 -10 Site Area Emergency Total Points 11 - 13 General Emergency

1. Compare plant conditions against the Fission Product Barrier Matrix on pages 3 through 5 of 5.

2. Determine the "potential loss" or "loss" status for each barrier (Containment, NCS and Fuel Clad) based on the EAL symptom description.

3. For each barrier, write the highest single point value applicable for the barrier in the "Points" column and mark the appropriate "potential loss" OR "loss" column.

4. Add the points in the "Points" column and record the sum as "Total Points".

5. Determine the classification level based on the number of "Total Points".

6. In the table on page 1 of this enclosure, under one of the four "classification" columns, select the event (e.g. 4.1.A.l for Loss of Nuclear Coolant System) that best fits the loss of barrier description.

7. Using that EAL number (e.g. 4. 1.A. 1) select the preprinted notification form OR a blank form and complete the required information for Emergency Coordinator/EOF Director approval and transmittal.

Enclosure 4.1 RP/OIAl5700/o00 Fission Product Barrier Matrix Page 3 of5 4.1.C CONTAINMENT BARRIER 4.1.N NCSBARRIER 4.1.F FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS- POTENTIAL LOSS - LOSS -

(1 Point) (3 Points) (4 Points) (5 Points) (4 Points) (5 Points)

1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status

Containment-RED.

Not applicable. NCS Integrity-

Not applicable.

Core Cooling-

Core Cooling-RED.

RED. ORANGE.

Core Cooling -

RED Path is

Heat Sink-RED.

Heat Sink-RED.

indicated for > 15 minutes.

2. Containment Conditions 2. NCS Leak Rate 2. Primary Coolant Activity Level

Containment

Rapid unexplained

Unisolable leak

GREATER THAN II. Not applicable.

Coolant Activity Pressure> 15 decrease in exceeding the available makeup GREATER THAN PSIG. containment capacity of one capacity as 300 IlCi/cc Dose pressure following charging pump in indicated by a loss Equivalent Iodine

H2 concentration initial increase. the normal ofNCS subcooling. (DEI) 1-131.

>9%. charging mode

Containment with letdown

Containment pressure or sump isolated.

pressure greater than level response not 3 psig with less than consistent with one full train ofNS LOCA conditions.

and a VX-CARF operating.

CONTINUED CONTINUED CONTINUED

Enclosure 4.1 RP/OIAl57001000 Fission Product Barrier Matrix Page 4 of5 4.1.C CONTAINMENT BARRIER 4.1.N NCSBARRIER 4.1.F FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

(1 Point) (3 Points) (4 Points) (5 Points) (4 Points) (5 Points)

3. Containment Isolation Valves Status After 3. SG Tube Rupture 3. Containment Radiation Monitoring Containment Isolation Actuation

Not applicable.

Containment

Primary-to-

Indication that a ... Not applicable.

Containment isolation is Secondary leak SG is Ruptured and radiation monitor incomplete and a rate exceeds the has a Non-Isolable EMF 51 Aor 51 B release path from capacity of one secondary line Reading at time containment exists. charging pump in fault. since shutdown the normal charging mode

Indication that a 0-0.5 hrs > 99 RIhr with letdown SG is ruptured and 0.5-2 hrs > 43 RIhr isolated. a prolonged release 2-4 hrs > 31 RIhr of contaminated 4-8 hrs > 22 RIhr secondary coolant >8 hrs > 13 Rlhr is occurring from the affected SG to the environment.

4. SG Secondary Side Release With Primary-to- 114. Containment Radiation Monitoring 4. Emergency CoordinatorlEOF Director Secondary Leakage Judgement

Not applicable.

Release of II. Not applicable.

Not applicable.

Any condition, including inability to monitor secondary side to the barrier, that in the opinion of the the environment Emergency CoordinatorlEOF Director with primary-to- indicates LOSS or POTENTIAL LOSS of secondary leakage the fuel clad barrier.

GREATER THAN Tech Spec allowable. END CONTINUED CONTINUED

Enclosure 4.1 RP/O/A/57001000 Fission Product Barrier Matrix Page 5 of5 4.1.C CONTAINMENT BARRIER 4.1.N NCSBARRIER 4.1.F FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

(1 Point) (3 Points) (4 Points) (5 Points) (4 Points) (5 Points)

5. Significant Radioactive Inventory In 5. Emergency CoordinatorlEOF Director Containment Judgement

Containment Rad.

Not applicable.

Any condition, including inability to monitor Monitor EMF51A the barrier, that in the opinion of the or 5IB Emergency Coordinator/EOF Director Reading @ time indicates LOSS or POTENTIAL LOSS of since shutdown: the NCS barrier.

> 390Rlhr@

0- 0.5 hr

> 170Rlhr@

0.5 - 2 hr END

> 125 Rlhr@

2 - 4 hr

> 90 RIhr @

4 - 8 hr

> 53 RIhr @

> 8 hr.

6. Emergency Coordinator IEOF Director Judgement Any condition, including inability to monitor the barrier, that in the opinion of the Emergency Coordinator/EOF Director indicates LOSS or POTENTIAL LOSS of the containment barrier.

END

Enclosure 4.2 RP/OIAl57001000 System Malfunctions Page 1 of2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.2.U.l Inability to Reach Required 4.2.A.l Unplanned Loss of Most or All 4.2.S.1 Inability to Monitor a END Shutdown Within Technical Safety System Annunciation or Significant Transient in Specification Limits. Indication in Control Room Progress.

With Either (1) a Significant OPERATING MODE: 1,2,3,4 Transient in Progress, or (2)

Compensatory Non-Alarming OPERATING MODE: 1,2,3,4 4.2.U.l-l Plant is not brought to required Indicators Unavailable.

operating mode within Technical 4.2.S.1-1 The following conditions Specifications LCO Action Statement OPERATING MODE: 1,2,3,4 exist:

Time.

4.2.A.l-l The following conditions exist: Loss of most (>50%)

4.2.U.2 Unplanned Loss of Most or All Safety annunciators associated with System Annunciation or Indication in Unplanned loss of most (>50%) safety systems.

the Control Room for Greater Than annunciators associated with safety 15 Minutes. systems for greater than 15 minutes. AND OPERATING MODE: 1,2, 3, 4 AND A significant plant transient is in progress.

4.2.U.2-1 The following conditions exist: In the opinion of the Operations Shift ManagerlEmergency AND Unplanned loss of most (>50%) CoordinatorlEOF Director, the annunciators associated with safety loss of the annunciators or Loss of the OAC.

systems for greater than 15 minutes. indicators requires additional personnel (beyond normal shift AND AND compliment) to safely operate the unit. Inability to provide manual AND monitoring of any of the In the opinion ofthe Operations Shift following Critical Safety ManagerlEmergency CoordinatorlEOF EITHER of the following: Functions:

Director, the loss of the annunciators

A significant plant transient is or indicators requires additional in progress.

subcriticality personnel (beyond normal shift

core cooling compliment) to safely operate the unit. OR

heat sink

containment.

CONTINUED

Loss of the OAC.

END END

Enclosure 4.2 RP/O/A/57001000 System Malfunctions Page 2 of2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.2.U.3 Fuel Clad Degradation.

OPERATING MODE: 1,2,3*

4.2.U.3-1 Dose Equivalent 1-131 greater than the Technical Specification allowable limit. (*mode 3 with Tavg =:> SOOOP) 4.2.UA Reactor Coolant System (NCS)

Leakage.

OPERATING MODE: 1, 2, 3, 4 4.2.UA-1 Unidentified leakage 2: 10 gpm.

4.2.UA-2 Pressure boundary leakage 2: 10 gpm.

4.2.UA-3 Identified leakage 2: 25 gpm.

4.2.U.5 Unplanned Loss of All Onsite or Offsite Communications.

OPERATING MODE: ALL 4.2. U.5-1 Loss of all onsite communications capability (internal phone system, PA system, onsite radio system) affecting the ability to perform routine operations.

4.2.U.5-2 Loss of all offsite communications capability (Selective Signaling, NRC ETS lines, offsite radio system, commercial phone system) affecting the ability to communicate with offsite authorities.

END

Enclosure 4.3 RP/O/A/5700/000 Abnormal Rad Levels/Radiological Effluent Page 1 of5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.U.l Any Unplanned Release of Gaseous or 4.3.A.l Any Unplanned Release of 4.3.S.1 Boundary Dose 4.3.G.l Boundary Dose Liquid Radioactivity to the Gaseous or Liquid Resulting from an Resulting from an Environment that Exceeds Two Times Radioactivity to the Actual or Imminent Actual or Imminent the SLC Limits for 60 Minutes or Environment that Exceeds Release of Release of Longer. 200 Times the SLC limits Radioactivity that Radioactivity that for 15 Minutes or Longer. Exceeds 100 mRem Exceeds 1000 OPERATING MODE: ALL TEDE or 500 mRem mRemTEDEor OPERATING MODE: ALL CDE Adult Thyroid 5000 mRem CDE Note: (This applies to all EALs in the 4.3.D.l for the Actual or Adult Thyroid for IC). If the monitor reading is sustained Note: (This applies to all EALs in the Projected Duration the Actual or for the time period indicated in the EAL 4.3.A.l IC). If the monitor of the Release. Projected Duration AND the required assessments reading is sustained for the time of the Release.

(procedure calculations) cannot be period indicated in the EAL OPERATING MODE: ALL completed within this time period, AND the required assessments OPERATING MODE: ALL declaration must be made based on the (procedure calculations) cannot Note 1: These EMF readings are valid radiation monitor reading. be completed within this time calculated based on Note 1: These EMF readings are period, declaration must be made average annual calculated based on 4.3.U.l-1 A valid indication on radiation monitor based on the valid radiation meteorology, site average annual EMF- 49L, EMF-44L or EMF-31 monitor reading. boundary dose rate, and meteorology, site (when aligned to RC) of design unit vent flow rate. boundary dose rate, and

5.45E+06 cpm for;

:: 60 minutes or will 4.3.A.l-l A valid indication on Calculations by the dose design unit vent flow likely continue for;::: 60 minutes, which radiation monitor EMF- 49H assessment team use rate. Calculations by the indicates that the release may have of;::: 1. 56 E + 03 cpm for actual meteorology, dose assessment team use exceeded the initiating condition and ;::: 15 minutes or willlike1y release duration, and unit actual meteorology, indicates the need to assess the release continue for ;:::15 minutes, which vent flow rate. Therefore, release duration, and unit with procedure HP/O/BJ1009101O, indicates that the release may these EMF readings vent flow rate.

HP/OIBJ10091029, or SH/01B12005/001. have exceeded the initiating should not be used if dose Therefore, these EMF condition and indicates the need assessment team readings should not be to assess the release with calculations are available. used if dose assessment (Continued) procedure HP/O/BJ1009101O, team calculations are HP/O/BJ10091029, or (Continued) available.

SH/O/B12005/00 1.

(Continued)

Enclosure 4.3 RP/0/Al5700/000 Abnormal Rad Levels/Radiological Effluent Page 2 of5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.U.I-2 A valid indication on radiation monitor 4.3.A.I-2 A valid indication on Note 2: If dose assessment team Note 2: If dose assessment team EMF- 36L of~ 2.05E+04 cpm for radiation monitor EMF- 36L calculations cannot be calculations cannot be 2: 60 minutes or will likely continue for of~ 2.05E+06 cpm for completed in 15 minutes, completed in 15 minutes,

~ 60 minutes, which indicates that the 2: 15 minutes or will likely then valid monitor reading then valid monitor release may have exceeded the initiating continue for ~ 15 minutes, should be used for reading should be used condition and indicates the need to assess which indicates that the release emergency classification. for emergency the release with procedure may have exceeded the initiating classification.

HP/O/B/l009101O, HP/O/B/I009/029, or condition and indicates the need 4.3.S.1-1 A valid indication SHiO/B/2005/00 1. to assess the release with on radiation monitor 4.3.G.l-l A valid indication procedure HP10/B/l 009/0 10, EMF-36Hof on radiation monitor 4.3.U.I-3 A valid indication on radiation monitor HP/O/B/l009/029, or 2:3.4 E + 03 cpm EMF-36Hof EMF-31 (when aligned to WC or SHiO/B/2005/00 1. sustained for 2:3.4 E + 04 cpm WWCB) of~ 9.174 E+03 cpm for 2: 15 minutes. sustained for

~ 60 minutes or will likely continue for 4.3.A.I-3 Gaseous effluent being released 2:15 minutes.

~ 60 minutes which indicates that the exceeds 200 times the level of 4.3.S.1-2 Dose assessment team release may have exceeded the initiating SLC 16.11-6 for 2: 15 minutes as calculations indicate 4.3.G.I-2 Dose assessment condition and indicates the need to assess determined by Radiation dose consequences team calculations the release with procedure Protection (RP) procedure. greater than 100 indicate dose HP/O/B/l009/010, HP/O/B/l009/029, or mRem TEDE or 500 consequences SHiO/B/2005/001. 4.3.A.I-4 Liquid effluent being released mRem CDE Adult greater than 1000 exceeds 200 times the level of Thyroid at the site mRemTEDEor 4.3.U.I-4 Gaseous effluent being released exceeds SLC 16.11-1 for 2: 15 minutes as boundary. 5000 mRem CDE two times SLC 16.l1-6 for determined by Radiation Adult Thyroid at the 2: 60 minutes as determined by Radiation Protection (RP) procedure. 4.3.S.1-3 Analysis of field site boundary.

Protection (RP) procedure. survey results or field (Continued) survey samples 4.3.G.I-3 Analysis offield 4.3.U.I-5 Liquid effluent being released exceeds indicates dose survey results or two times SLC 16.11-1 for consequences greater field survey samples 2: 60 minutes as determined by Radiation than 100 mRem indicates dose Protection (RP) procedure. TEDE or 500 mRem consequences (Continued) CDE Adult Thyroid greater than 1000 at the site boundary. mRemTEDEor END 5000 mRem CDE Adult Thyroid at the site boundary.

END

Enclosure 4.3 RP/OIAl5700/000 Abnormal Rad Levels/Radiological Effluent Page 3 of5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.U.2 Unexpected Increase in Plant 4.3.A.2 Major Damage to Radiation or Airborne Concentration. Irradiated Fuel or Loss of Water Level that Has or OPERATING MODE: ALL Will Result in the Uncovering of Irradiated 4.3.U.2-1 Indication of uncontrolled water level Fuel Outside the Reactor decrease of greater than 6 inches in the Vessel.

reactor refueling cavity with all Does not apply to spent fuel in dry irradiated fuel assemblies remaining cask storage. Refer to EPLAN covered by water. section D basis document.

4.3.U.2-2 Uncontrolled water level decrease of OPERATING MODE: ALL greater than 6 inches in the spent fuel pool and fuel transfer canal with all 4.3.A.2-1 An unplanned valid trip II irradiated fuel assemblies remaining alarm on any of the covered by water. following radiation monitors:

4.3.U.2-3 Unplanned valid area EMF reading exceeds the levels shown in Enclosure Spent Fuel Building 4.10. Refueling Bridge lEMF-17 2EMF-4 END Spent Fuel Pool Ventilation lEMF-42 2EMF-42 Reactor Building Refueling Bridge lEMF-16*

2EMF-3*

Containment Noble Gas lEMF-39*

2EMF-39*

Applies to Mode 6 and No Mode Only.

(Continued)

Enclosure 4.3 RP/O/Al5700/000 Abnormal Rad Levels/Radiological Effluent Page 4 of5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.2-2 Plant personnel report that water level drop in reactor refueling cavity, spent fuel pool, or fuel transfer canal has or will exceed makeup capacity such that any irradiated fuel will become nncovered.

4.3.A.2-3 NC system wide range level

<358 inches after initiation ofNC system make-up.

AND Any irradiated fuel assembly not capable of being lowered into spent fuel pool or reactor vessel.

4.3.A.2-4 Spent Fuel Pool or Fuel Transfer Canal level decrease of>2 feet after initiation of makeup.

AND Any irradiated fuel assembly not capable of being fully lowered into the spent fuel pool racks or transfer canal fuel transfer system basket.

(Continued)

Enclosure 4.3 RP/O/N5700/000 Abnormal Rad Levels/Radiological Effluent Page 5 of5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.

OPERATING MODE: ALL 4.3.A.3-1 Valid reading on EMF-12 greater than 15 mR/hr in the Control Room.

4.3.A.3-2 Valid indication of radiation levels greater than 15 mR/hr in the Central Alarm Station (CAS) or Secondary Alarm Station (SAS).

4.3.A.3-3 Valid area EMF reading exceeds the levels shown in Enclosure 4.10.

END

Enclosure 4.4 RP/OIAl57001000 Loss of Shutdown Functions Page 1 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY END 4.4.A.l Failure of Reactor 4.4.S.1 Failure of Reactor 4.4.G.l Failure of the Reactor Protection System Protection System Protection System to Instrumentation to Complete Instrumentation to Complete Complete an Automatic Trip or Initiate an Automatic or Initiate an Automatic and Manual Trip WAS NOT Reactor Trip Once a Reactor Trip Once a Successful and There is Reactor Protection System Reactor Protection System Indication of an Extreme Setpoint Has Been Exceeded Setpoint Has Been Exceeded Challenge to the Ability to and Manual Trip WAS and Manual Trip WAS NOT Cool the Core.

Successful. Successful.

OPERATING MODE: 1 OPERATING MODE: 1,2,3 OPERATING MODE: 1 4.4.G.l-l The following conditions exist:

4.4.A.l-l The following conditions exist: 4.4.S.1-1 The following conditions exist:

Valid reactor trip signal Valid reactor trip signal Valid reactor trip signal received or required and received or required and received or required and automatic reactor trip automatic reactor trip automatic reactor trip was not successful.

was not successful. was not successful.

AND AND AND Manual reactor trip from the Manual reactor trip from the Manual reactor trip from the control room was not control room is successful and control room was not successful in reducing reactor reactor power is less than 5% successful in reducing reactor power to less than 5% and and decreasing. power to less than 5% and decreasing.

decreasing.

(Continued) AND (Continued) EITHER of the following conditions exist:

Core Cooling CSF-RED

Heat Sink CSF-RED.

END

Enclosure 4.4 RP/OIAl57001000 Loss of Shutdown Functions Page 20[3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.A.2 Inability to Maintain Plant 4.4.S.2 Complete Loss of Function in Cold Shutdown. Needed to Achieve or Maintain Hot Shutdown.

OPERATING MODE: 5,6 OPERATING MODE: 1,2,3,4 4.4.A.2-1 Total loss ofND and/or RN and/orKC. 4.4.S.2-1 Sub criticality eSP-RED.

AND 4.4.S.2-2 Heat Sink eSP-RED.

One of the following: 4.4.S.3 Loss of Water Level in the Reactor Vessel That Has or

Inability to maintain Will Uncover Fuel in the reactor coolant temperature Reactor Vessel.

below200oP OPERATING MODE: 5,6 OR 4.4.S.3-1 Pailure of heat sink causes loss of cold shutdown conditions.

Uncontrolled reactor coolant temperature rise to AND

>180 oP.

Lower range Reactor Vessel END Level Indication System (RVLIS) decreasing after initiation ofNe system makeup.

(Continued)

Enclosure 4.4 RP/OIAJ57001000 Loss of Shutdown Functions Page 3 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.S.3-2 Failure of heat sink: causes loss of cold shutdown conditions.

AND Reactor Coolant (NC) system narrow range level less than 6 inches and decreasing after initiation ofNC system makeup.

4.4.S.3-3 Failure of heat sink: causes loss of cold shutdown conditions.

AND Either train ultrasonic level indication less than 6 inches and decreasing after initiation ofNC system makeup.

END

Enclosure 4.5 RP/OIAl57001000 Loss of Power Page 1 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.5.U.l Loss of All Offsite 4.5.A.l Loss of All Offsite 4.5.S.1 Loss of All Offsite 4.5.G.l Prolonged Loss of All Power to Essential Power and Loss of All Power and Loss of All (Offsite and Onsite) AC Busses for Greater Than Onsite AC Power to Onsite AC Power to Power.

15 Minutes. Essential Busses During Essential Busses.

Cold Shutdown Or OPERATING MODE: 1,2,3,4 OPERATING MODE: 1,2,3,4 Refueling Mode. OPERATING MODE: 1,2,3,4 4.5.G.l-l Prolonged loss of all 4.5.U.l-l The following conditions OPERATING MODE: 5,6, No 4.5.S.1-1 Loss of all offsite and offsite and onsite AC exist: Mode onsite AC power as power as indicated by:

indicated by:

Loss of offsite power to 4.5.A.l-l Loss of all offsite and Loss of power on essential essential buses ETA and onsite AC power as Loss of power on essential buses ETA and ETB for ETB for greater than indicated by: buses ETA and ETB. greater than 15 minutes.

15 minutes.

Loss of power on essential AND AND AND buses ETA and ETB.

Failure to restore power to Standby Shutdown Both emergency diesel AND at least one essential bus Facility (SSF) fails to generators are supplying within 15 minutes. supply NC pump seal power to their respective Failure to restore power to injection OR CA supply essential busses. at least one essential bus to Steam Generators.

within 15 minutes. (Continued)

AND (Continued)

(Continued)

Enclosure 4.5 RP/OIA/57001000 Loss of Power Page 2 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY OPERATING MODE: 5,6, No 4.5.A.2 AC Power to Essential 4.5.S.2 Loss of All Vital DC At least one of the Mode Busses Reduced to a Power. following conditions Single Power Source for exist:

4.5.U.I-2 The following conditions Greater Than 15 OPERATING MODE: 1,2,3,4 exist: Minutes Such That An

Restoration of at least Loss of offsite power to Additional Single 4.5.S.2-1 The following conditions one essential bus essential buses ETA and Failure Could Result in exist: within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is ETB for greater than Station Blackout. NOT likely 15 minutes. Loss of both unit related OPERATING MODE: 1,2,3,4 EVDA andEVDD busses

Indication of AND as indicated by bus continuing 4.5.A.2-1 The following condition voltage less than degradation of core One emergency diesel exists: 110 VDC. cooling based on generator is supplying Fission Product power to its respective AC power capability has AND Barrier monitoring.

essential bus. been degraded to one essential bus powered Failure to restore power to END from a single power at least one required DC source for> 15 min. due bus within 15 minutes Continued to the loss of all but one from the time ofloss.

of:

END SATA SATB ATC ATD DIG A DIG B.

END

Enclosure 4.5 RP/O/A/57001000 Loss of Power Page 3 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.5.U.2 Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes.

OPERATING MODE: 5,6 4.5.U.2-1 The following conditions exist:

Unplanned loss of both unit related EVDA and EVDD busses as indicated by bus voltage less than 110 VDC.

AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

END

Enclosure 4.6 RP/O/A/57001000 Fire/Explosion and Security Events Page 1 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.6.U.l Fire Within Protected Area 4.6.A.l Fire or Explosion Affecting 4.6.S.1 Security Event in a Plant 4.6.G.l Security Event Resulting in Boundary NOT the Operability of Plant Vital Area. Loss of Physical Control of Extinguished Within Safety Systems Required to the Facility.

15 Minutes of Detection OR Establish or Maintain Safe OPERATING MODE: ALL Explosion Within the Shutdown. OPERATING MODE: ALL Protected Area Boundary. 4.6.S.1-1 Intrusion into any of the OPERATING MODE: 1,2,3,4,5,6 following plant areas by a 4.6.G.l-l A HOSTILE FORCE has OPERATING MODE: ALL HOSTILE FORCE: taken control of plant 4.6.A.l-1 The following conditions exist:

Reactor Building equipment such that plant 4.6.U.l-l Fire in any of the following (includes non-security events)

Auxiliary Building personnel are unable to operate areas NOT extinguished Fire or explosion in any of the

Diesel Generator Rooms equipment required to within 15 minutes of control following areas:

Control Room maintain safety functions.

room notification or

Reactor Building

Standby Shutdown Facility verification of a control room

Auxiliary Building

Doghouses END fire alarm.

Diesel Generator Rooms

CAS

Control Room

SAS.

Reactor Building

Standby Shutdown Facility

Auxiliary Building

CAS 4.6.S.1-2 Security confirmed bomb

Diesel Generator Rooms

SAS discovered/exploded in a vital

Control Room

FWST area.

Standby Shutdown Facility

Doghouses (Applies in

CAS Mode 1,2,3,4 only). 4.6.S.1-3 Security confirmed sabotage in

SAS a plant vital area.

Doghouses AND

FWST 4.6.S.1-4 Other security events as

Turbine Building determined from Safeguards

Service Building (Continued) Contingency Plan and reported

Interim Radwaste Building by the security shift supervision.

Equipment Staging Building (Continued)

ISFSI.

(Continued)

Enclosure 4.6 RP/O/A/57001000 Fire/Explosion and Security Events Page 2 of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY One of the following: 4.6.S.2 Site Attack 4.6.U.I-2 Report by plant personnel of an unanticipated explosion Note: Only one train of a system OPERATING MODE: ALL within the protected area needs to be affected or boundary resulting in visible damaged in order to satisfY 4.6.S.2-1 A notification from the site damage to permanent this condition. security force that an armed structures or equipment or a attack, explosive attack, airliner loaded cask in the ISFSI.

Affected safety system impact, or other HOSTILE parameter indications show ACTION is occurring or has 4.6.U.2 Confirmed Security Event degraded performance occurred within the protected Which Indicates a Potential Plant personnel report area.

Degradation in the Level of visible damage to Safety of the Plant. permanent structures or END equipment within the OPERATING MODE: All specified area.

4.6.U.2-1 Security events as detennined 4.6.A.2 Fire or Explosion Affecting from Safeguards Contingency the Operability of Plant Plan and reported by the Safety Systems Required to security shift supervision. Establish or Maintain Safe Shutdown.

4.6.U.2-2 A credible site-specific security threat notification. OPERATING MODE: No Mode 4.6.U.2-3 A validated notification from 4.6.A.2-1 The following conditions exist:

NRC providing information of (includes non-security events) an aircraft threat. Fire or explosion in any of the following areas:

4.6.U.2-4 Hostage situation/extortion.

Spent Fuel Pool

Auxiliary Building.

4.6.U.2-5 A violent civil disturbance within the owner controlled AND area.

(Continued)

END

Enclosure 4.6 RP/OIAl57001000 Fire/Explosion and Security Events Page 3 of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY One of the following:

Note: Only one train of a system needs to be affected or damaged in order to satisfy this condition.

Spent Fuel Pool level and/or temperature show degraded performance

Plant personnel report visible damage to permanent structures or equipment supporting Spent Fuel Pool Cooling.

4.6.A.3 Other Security Events as Determined from Safeguards Contingency Plan.

OPERATING MODE: ALL 4.6.A.3-1 Other security events as determined from Safeguards Contingency Plan and reported by the security shift supervision.

4.6.A.4 Notification of an Airborne Attack Threat.

OPERATING MODE: ALL 4.6.A.4-1 A validated notification from NRC of airliner attack threat less than 30 minutes away.

(Continued)

Enclosure 4.6 RP/O/A/5700/000 Fire/Explosion and Security Events Page 40[4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.6.A.S Notification of HOSTILE ACTION within the OCA.

OPERATING MODE: ALL 4.6.A.5-1 A notification from the site security force that an armed attack, explosive attack, airliner impact or other HOSTILE ACTION is occurring or has occurred within the OCA.

END

Enclosure 4.7 RP/O/A/57001000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 1 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.U.l Natural and Destructive 4.7.A.l Natural and Destructive 4.7.S.1 Control Room Evacuation 4.7.G.l Other Conditions Existing Phenomena Affecting the Phenomena Affecting the Has Been Initiated and Plant Which in the Judgement of Protected Area. Plant Vital Area. Control Cannot Be the Emergency Established. Coordinator/EOF Director OPERATING MODE: ALL OPERATING MODE: ALL Warrant Declaration of OPERATING MODE: ALL General Emergency.

4.7.U.l-l Tremor felt and valid alarm on 4.7.A.l-l Valid "OBE Exceeded" Alarm the Syscom Seismic on IAD-13, E-7 4.7.S.1-1 The following conditions OPERATING MODE: ALL Monitoring System (OAC exist:

MlD2422). 4.7.A.I-2 Tornado or high winds: 4.7.G.l-l Other conditions exist which Control Room evacuation has in the Judgement of the 4.7.U.I-2 Report by plant personnel of Tornado striking plant been initiated per Emergency tornado striking within structures within the vital AP/l(2)/AJ5500/017, or CoordinatorlEOF Director protected area area: AP/l(2)/AJ5500/024. {3] indicate:

boundarylISFSI.

Reactor Building AND (1) actual or imminent 4.7.U.I-3 Vehicle crash into plant

Auxiliary Building substantial core degradation structures or systems within

FWST Control of the plant cannot be with potential for loss of protected area

Diesel Generator Rooms established from the Auxiliary containment, boundarylISFSI.

Control Room Shutdown Panel or the

Standby Shutdown Standby Shutdown Facility OR 4.7.U.I-4 Report of turbine failure Facility within 15 minutes.

resulting in casing penetration (2) potential for

Doghouses or damage to turbine or uncontrolled radionuclide

CAS releases. These releases can generator seals. (Continued)

SAS.

reasonably be expected to (Continued) OR exceed Environmental Protection Agency Sustained winds;:::: 74 mph for Protective Action Guideline

> 15 minutes. {4} levels outside the site boundary.

(Continued)

END

Enclosure 4.8 RP/O/A/5700/000 Definitions/Acronyms Page 2 of3 HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

IMMINENT - Expected to occur within 1-3 hours.

INABILITY TO DIRECTLY MONITOR - Operational Aid Computer data points are unavailable or gauges/panel indications are not readily available to the operator.

INTRUSION/INTRUDER - Suspected hostile individual present in a protected area without authorization.

ISFSI - Independent Spent Fuel Storage Installation.

NO MODE - Defue1ed.

PROLONGED - a duration beyond norma11imits, defined as "greater than 15 minutes" or as determined by the judgement of the Emergency Coordinator.

PROTECTED AREA - Encompasses all owner controlled areas within the security perimeter fence.

REACTOR COOLANT SYSTEM (RCSINCS) LEAKAGE - RCS Operational Leakage as defined in the Technical Specification Basis B 3.4.13.

RUPTURED - (As relates to Steam Generator) - Existence of primary to secondary leakage of a magnitude sufficient to require or cause a reactor trip and safety injection.

SABOTAGE - Deliberate damage, misalignment, or misoperation of plant equipment with the intent to render the equipment unavailable.

SECURITY EVENT - A security related emergency situation for which prompt response by the Security Force, immediate action by plant personnel, and/or assistance from offsite agencies may be required to apprehend intruders and mitigate the effects of or prevent radiological sabotage.

SIGNIFICANT TRANSIENT-An unplanned event involving one or more of the following: (1) automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full electrical load; (3) reactor trip, (4) safety injection, (5) thermal power oscillations 2:10%.

SITE AREA EMERGENCY - Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile action that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Enclosure 4.8 RP/O/A/57001000 Definitions/Acronyms Page 3 of3 SITE BOUNDARY - That area, including the protected area, in which Duke Power Company has the authority to control all activities, including exclusion or removal of personnel and property.

SLC - Selected Licensee Commitments.

SUSTAINED - A duration of time long enough to confirm that the CSF is valid (not momentary).

TOTAL EFFECTNE DOSE EQUNALENT (TED E) - The sum of external dose exposure to a radioactive plume, to radionuclides deposited on the ground by the plume, and the internal exposure from inhaled radionuclides deposited in the body.

TOXIC GAS - A gas that is dangerous to life or health by reason of inhalation or skin contact (e.g.

chlorine).

UNCONTROLLED - Event is not the result of planned actions by the plant staff.

UNPLANNED - An event or action is UNPLANNED if it is not the expected result of normal operations, testing, or maintenance. Events that result in corrective or mitigative actions being taken in accordance with abnormal or emergency procedures are UNPLANNED.

UNUSUAL EVENT - . Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

VALID - An indication or report or condition is considered to be VALID when it is conclusively verified by: (1) an instrument channel check, or (2) indications on related or redundant instrumentation, or (3) by direct observation by plant personnel such that doubt related to the instrument's operability, the condition's existence or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

VIOLENT - Force has been used in an attempt to injure site personnel or damage plant property.

VISIBLE DAMAGE - Damage to equipment or structure that is readily observable without measurements, testing, or analyses. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage:

deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering.

VITAL AREA - Areas within the PROTECTED AREA that house equipment important for nuclear safety. Access to a VITAL AREA is allowed only if an individual has been authorized to be in that area per the Security plan, therefore VITAL AREA is a Security term.

Enclosure 4.9 RP/O/A/5700/000 Emergency Declaration Guidelines Page 1 of3 THE FOLLOWING GUIDANCE IS TO BE USED BY THE EMERGENCY COORDINATOR IN ASSESSING EMERGENCY CONDITIONS.

The Emergency Coordinator shall review all applicable initiating events to ensure proper classification.

The BASIS Document (located in Section D of the McGuire Nuclear Site Emergency Plan) is available for review if any questions arise over proper classification.

If an event occurs on more than one unit concurrently, the event with the higher classification will be classified on the emergency notification form. Information relating to the problem on the other unit will be captured on the emergency notification form line 13 remarks section.

The Affected Unites) on Line 11 is tied to the EAL (IC) Number and EAL (IC) Description on Line

4. Certain events could occur at the plant site such that multiple units are affected. These may include Abnormal Rad Levels/Radiological Effluents; Fire/Explosion and Security Events; and Natural Disasters, Hazards, and Other Conditions Affecting Plant Safety. This shall be considered when evaluating the accuracy of the Unit designation. {PIP 0-M97-4638} If the initiating event puts more than one unit in the same Emergency Classification (example Alert), then the unit designation may be either the Affected Unit numbers or "All." If the initiating event drives one unit to a higher classification, then the unit with the higher classification should be listed as Affected Unit. {PIPs M-03-3294 and C-04-2586}

The EAL (IC) Number and EAL (IC) Description provided on Line 4 of the emergency notification form should be based on the highest emergency classification that applies. Other classifiable events should be included on Line 13, Remarks, on the emergency notification form, but not given an EAL number. {PIPs M-03-3294 and C-04-2586}

If an event occurs, and a lower or higher plant operating mode is reached before the classification can be made, the classification shall be based on the mode that existed at the time the event occurred.

The fission product barrier matrix is applicable only to those events that occur at hot shutdown or higher. An event that is recognized at cold shutdown or lower shall not be classified using the fission product barrier matrix. Reference would be made to the additional enclosures that provide emergency action levels for specific events (e.g. severe weather, fire, security).

If a transient event should occur, the following guidance is provided.

1. Some emergency action levels specify a specific duration. For these EALs, the classification is made when the Emergency Coordinator assessment concludes that the specified duration is exceeded or will be exceeded (i.e. condition cannot be reasonably corrected before the duration elapses), whichever is sooner.

Enclosure 4.9 RP/O/A/5700/000 Emergency Declaration Guidelines Page 2 of3

2. If a plant condition exceeding EAL criteria is corrected before the specified duration time is exceeded, the event is NOT classified by that EAL. Lower Severity EALs, if any, shall be reviewed for possible applicability in these cases.

3. If a plant condition exceeding EAL criteria is not recognized at the time of occurrence, but is identified well after the condition has occurred (e.g. as a result of routine log or record review) and the condition no longer exists, an emergency shall NOT be declared. Reporting under 10CFR50.72 may be required. Such a condition could occur, for example, if a follow-up evaluation of an abnormal condition uncovers evidence that the condition was more severe than earlier believed.

4. If an emergency classification was warranted, but the plant condition has been corrected prior to declaration and notification, the following are applicable: {2}

a. For UNUSUAL EVENT, the emergency shall be declared and the condition shall be reported. The event should be terminated in a follow-up notification as soon as time permits, but within one hour.

b. For ALERT, SITE AREA EMERGENCY, and GENERAL EMERGENCY, the emergency shall be declared and the Emergency Response Organization shall be activated. The TSC Emergency Coordinator shall be responsible for terminating the emergency as soon as time permits when appropriate.

c. The Control Room Emergency Coordinator (Operations Shift Manager) shall ensure that any required follow-up notifications are conducted as required prior to activation of the TSC.

DETERMINATION OF "EVENT TIME" (TIME THE 15 MINUTE OFF SITE NOTIFICATION CLOCK STARTS)

1. Ifplant conditions require implementation ofEPIl or 2/A/5000/E-O (Reactor Trip or Safety Injection), increased emphasis shall be given to evaluation of plant conditions for determination of EAL(s) when "kickout" of the diagnostic procedure occurs. "Event Time" is the time at which the EAL(s) is determined to be valid by the Emergency Coordinator/EOF Director.

2. Ifplant conditions do not require implementation ofEPIl or 2/A/5000/E-O (Reactor Trip or Safety Injection), and conditions of a specific EAL are met, the "Event Time" is the time at which the EAL(s) is determined to be valid by the Emergency Coordinator/EOF Director.

3. The time the event is classified shall be entered on the initial emergency notification form.

Enclosure 4.9 RP/OIAl57001000 Emergency Declaration Guidelines Page 3 of3 MOMENTARY ENTRY INTO A HIGHER CLASSIFICATION If, while in an emergency classification, the specified EALs of a higher classification are met momentarily, and in the judgment of the Emergency Coordinator are not likely to recur, the entry into the higher classification must be acknowledged. Acknowledgment is performed as follows:

If this condition occurs prior to the initial notification to the emergency response organization and off site agencies, the initial message should note that the site is currently in the lower classification, but had momentarily met the criteria for the higher classification. It should also be noted that plant conditions have improved and stabilized to the point that the criteria for the higher classification are not expected to be repeated.

Enclosure 4.10 RP/O/Al5700/000 Radiation Monitor Readings for Enclosure 4.3 EALs Page 1 of 1 Note: These values are not intended to apply to anticipated temporary increases due to planned events (e.g. inc ore detector movement, radwaste container movement, depleted resin transfers, etc.)

Detector Elevation Column Identifier Unusual Alert Event mRlhr mRlhr lEMF-1 695' FF, GG-56 Aux. Bldg. Corridor 500 5000 1EMF-5 716' FF-54 Unit 1 NM Sample Room 600 5000 1EMF-8 733' HH-56 Aux. Bldg. Corridor 100 5000 1EMF-1O 750' LL-56 Aux. Bldg. Corridor 100 5000 lEMF-13 775' QQ-56 Shift Lab/Count Room 100 5000 1EMF-17 786' N/A Unit 1 Spent Fuel Pool Refueling Bridge 100 5000 2EMF-1 716' EE, FF-58 Unit 2 NM Sample Room 300 5000 2EMF-4 786' N/A Unit 2 Spent Fuel Pool Refueling Bridge 100 5000 2EMF-9 767' J!-5~ _ _ Aux. Bl~orridor -- _._- 100 5000

Enclosure 4.11 RP/OIAl57001000 Commitment Reference for Emergency Action Levels Page 1 of 1

{I} PIP-M-OO-2138, CA # 18

{2} PIP-M-02-0187, CA # 6

{3} PIP-M-OI-2860, CA #2

{4} PIP-M-03-4281, CA #3

{5} PIP-M-05-3403, CA#3, multiple changes in enclosure 4.6

Enclosure 4.8 RP/O/A/5700/000 Definitions/Acronyms Page 1 of3 ALERT - Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels ALL - (As relates to Operating Mode Applicability) - At all times.

BOMB - A fused explosive device.

CIVIL DISTURBANCE - A group of ten (10) or more people violently protesting station operations or activities at the site. A civil disturbance is considered to be violent when force has been used in an attempt to injure site personnel or damage plant property.

CONFINEMENT BOUNDARY - The barrier(s) between areas containing radioactive substances and the environment.

EXPLOSION - A rapid, violent unconfined combustion, or a catastrophic failure of pressurized equipment (e.g., a steamline or feedwater line break) that imparts energy sufficient to potentially damage or creates shrapnel to actually damage permanent structures, systems or components.* An electrical breaker flash that creates shrapnel and results in damage to other components beyond scorching should also be considered.

EXTORTION - An attempt to cause an action at the site by threat of force.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flames is preferred but is NOT required if large quantities of smoke and heat are observed. An electrical breaker flash that creates high temperatures for a short duration and merely localized scorching to that breaker and its compartment should not be considered a fire.

GENERAL EMERGENCY - Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile action that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guidelines exposure levels offsite for more than the immediate site area.

HOSTAGE - A person or object held as leverage against the site to ensure demands will be met by the site.

HOSTILE ACTION - An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, takes hostages, and / or intimidates the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area).

Enclosure 4.7 RP/O/A/57001000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 4 of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.A.3 Control Room Evacuation Has Been Initiated.

OPERATING MODE: ALL 4.7.A.3-1 Control Room evacuation has been initiated per AP/l(2)/A/5500/017, or AP/l (2)/Al5500/024. {3}

4.7.A.4 Other Conditions Existing Which in the Judgement of the Emergency CoordinatorlEOF Director Warrant Declaration of an Alert.

OPERATING MODE: ALL 4.7.A.4-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.

END

Enclosure 4.7 RP/O/A/57001000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 3 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.U.3 Other Conditions Existing 4.7.A.2 Release of Toxic or Which in the Judgement of Flammable Gases Within a the Emergency Facility Structure Which Coordinator/EOF Director Jeopardizes Operation of Warrant Declaration of an Systems Required to Unusual Event. Maintain Safe Operations or to Establish or Maintain OPERATING MODE: ALL Cold Shutdown.

4.7.U.3-1 Other conditions exist which OPERATING MODE: ALL in the judgement of the Emergency Coordinator/EOF Note: Structures for the below EALs:

Director indicate a potential

Reactor Building degradation of the level of

Auxiliary Building safety of the plant.

Diesel Generator Rooms END

Control Room

Standby Shutdown Facility

Doghouses

CAS

SAS.

4.7.A.2-1 Report or detection oftoxic gases within a Facility Structure in concentrations that will be life threatening to plant personnel.

4.7.A.2-2 Report or detection of flannnable gases within a Facility Structure in concentra-tions that will affect the safe operation of the plant.

(Continued)

Enclosure 4.7 RP/O/A/57001000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 2 of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.U.1-5 Independent Spent Fuel Cask 4.7.A.1-3 Visible structural damage 4.7.S.2 Other Conditions Existing tipped over or dropped greater caused by either: Which in the Judgement of than 12 inches.

Vehicle crashes, OR the Emergency

Turbine failure generated Coordinator/EOF Director 4.7.U.1-6 Uncontrolled flooding in the missiles, OR Warrant Declaration of Site ISFSI area.

Other catastrophic events Area Emergency.

4.7.U.1-7 Tornado generated missile(s) on any of the following plant OPERATING MODE: ALL impacting the ISFSI. structures:

4.7.S.2-1 Other conditions exist which

Reactor Building in the Judgement of the 4.7.U.2 Release of Toxic or

Auxiliary Building Emergency Coordinator/EOF Flammable Gases Deemed Director indicate actual or

FWST Detrimental to Safe likely major failures of plant

Diesel Generator Rooms Operation of the Plant. functions needed for

Control Room protection of the public.

Standby Shutdown END OPERATING MODE: ALL Facility 4.7.U.2-1 Report or detection of toxic or

Doghouses flammable gases that could

CAS enter within the site boundary

SAS in amounts that can affect safe

Ultimate heat sink (Standby Nuclear Service operation of the plant. Water Pond Dam and Dikes and Cowan's Ford Dam and 4.7.U.2-2 Report by Local, County or associated Dikes).

State Officials for potential evacuation of site personnel based on offsite event.

(Continued)

ML091700317: Difference between revisions

Latest revision as of 11:32, 12 March 2020

Text

References:

Navigation menu

ML091700317: Difference between revisions

Latest revision as of 11:32, 12 March 2020

Text

References:

Navigation menu

Search