McGuire Initial Exam 2011-301 Final SRO Written Exam. · 2013. 7. 8. · the inadvertent closure of IRN-277B (RB Non Ess Ret Cont Outside Isol) At 1400, the following NC pumps indications

U.S. Nuclear Regulatory Commission

Site-Specific SRO Written Examination

Applicant Information

Date: Facility/Unit: MCGUIRE

Region: I (jj III IV Reactor Type: CE BW GE

Start Time: Finish Time:

Instructions

Use the answer sheets provided to document your answers. Staple this cover sheeton top of the answer sheets. To pass the examination you must achieve a final grade of at least80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunctionwith the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. Youhave 8 hours to complete the combined examination, and 3 hours 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

Resu Its

RO/SRO-OnIy/Total Examination Values / I PcJilts

Applicant’s Score I I Points

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McGuire Nuclear Station• 21111 MNS SRO NRC Examination

Question: II point)

Given the following conditions on Unit 1:

• Unit is operating at 100% RTP• Ops Test Group is performing a slave relay test and a procedural error results in

the inadvertent closure of IRN-277B (RB Non Ess Ret Cont Outside Isol)• At 1400, the following NC pumps indications are noted:

NC Pump IA lB IC ID

Current Stator 275°F 260°F 266°F 270°FWinding Temp.

Temp. Increase 1.1°F/mm 1.6°F/mm 1.5°F/mm 1.35°F/mm

Based on the conditions above, which ONE (1) of the following lists the FIRST NC pumpthat would have to be manually secured in accordance with AP-08, (Malfunction of NCPump)?

A. IA

B. lB

C. 1C

D. 1D

Page 1 of 100

McGuire Nuclear Station2011 MNS SliO NRC Examination

Question: 21 point)


• An NC pump is to be started for a unit heatup to normal operating temperatureand pressure

• The 1A2 Oil Lift pump has been started in preparation for starting the IA NCP• Oil Lift pressure is 550 PSIG

Based on the conditions above, the 1A NCP (1) start.

The MINIMUM required #1 Seal differential pressure for starting the NC pump is(2) PSID.

Which ONE (1) of the following completes the statements above?

A. 1. will2. 200

B. I. will2. 350

C. 1. willnot2. 200

D. 1. willnot2. 350

Page 2 of 100

McGuire Nuclear Station• 2011 MNS SRO NRCExamination

Question: 31 point)

Given the following condition on Unit 1:

• Unit is in Mode 3 recovering from a reactor trip.• Due to a PZR Pressure control malfunction, the crew had to establish NV Aux

Spray per Generic Enclosure G-1 Enc 3 (Establishing NV Aux Spray) tofacilitate repairs.

• B/U PZR heaters A and B are energized and pressure is stable• During repair activities, 1 NV-i 3B (NV Supply to A NC Loop) opens

inadvertently.

Based on the conditions above, which ONE (1) of the following describes the effect onPZR pressure and what action is required per Gen Enc G-1 Enc 3 to control pressure?

A. Pressure would INCREASEManually throttle open 1NC-29C (B NC Loop Spray Control)

B. Pressure would DECREASEManually throttle open 1NC-29C (B NC Loop Spray Control)

C. Pressure would INCREASEDeenergize PZR heaters as required to control pressure

D. Pressure would DECREASEEnergize additional PZR Heaters as required to control pressure

Page 3 of 100


Question: 41 point)


• The unit is in MODE 4 with the “A” Train of ND in service• LTOP key switches for PORVs are selected to “NORM” position• Annunciator 1AD-9 I Al (1 ND-2A Open and NC Hi Press) has just come in

alarm• NC WR pressure is indicating 460 PSIG

Which ONE (1) of the following describes the redundant indication which would beused to verify that this annunciator is valid?

A. PRT Level increasing ONLY

B. CF&E sump level increasing ONLY

C. I ND-I B (“C” HL Suction to ND Isolation) would have auto closed

D. PORVs NC-32B and NC-34A would be lifting

Page4oflOO

McGuire Nuclear Station• 2011MNS SRO NRC Examination

Question: 5‘1 point)

Unit I is shutdown with the following conditions:

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

• The flow rate on each ND train is currently 1500 GPM• 1 ND-34 (IA & I B ND Hx Byp Isol) is initially 50% OPEN• NC system temperature is being maintained at 170° F• The air line for 1 ND-34 breaks off

Which ONE (1) of the following describes the indications that will be observed by theOperators in the Main Control Room over the next 15 minutes?

IA ND Train Total Flowrate lB ND HX Inlet Temperature

A. Increase Increase

B. Increase Decrease

C. Decrease Increase

D. Decrease Decrease

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McGu ire Nuclear Station2U11MNSSRONRCExamiiatioi



A Safety Injection has occurred due to a Small-Break LOCA

Which ONE (I) of the following identifies the valves or groups of valves that willreceive a signal to automatically open on the Safety Injection (Se) Signal?

A. 1 ND-14 and I ND-29 (A/B ND Heat Exchanger Outlets) AND1 NI-9A and I NI-I OB (CCP Pump Isolation to Cold Legs)

B. I ND-I4 and I ND-29 (A/B ND Heat Exchanger Outlets) ANDI NI-I 73A and I NI-I 78B (Train A/B ND to Loop C and D Cold Legs)

C. I NI-i 62A (NI Pump Discharge Common Isolation to Cold Legs) AND1 Nl-9A and I NI-i OB (CCP Pump Isolation to Cold Legs)

D. 1 NI-i 62A (NI Pump Discharge Common Isolation to Cold Legs) AND1 NI-i 73A and I NI-I 78B (Train A/B ND to Loop C and D Cold Legs)

Page 6 of 100

McGuire Nuclear Static2011MNSSRONRCExamInar

Question: 7(1 point)

Given the following conditions on Unit 1:/7)

• A LOCA has/occurred inside Containment• E-1 (Loss or Reactor or Secondary Coolant) has been implem• NC system pressure is 400 PSIG• One of the Cold Leg Accumulators (CLA) injects N2 into the NL

the crew reaching the step in E-1 which directs isolating the CLAs

Which ONE (1) of the following would result in the highest probability of N2 injectioninto the NC system?

Accumulator pressure was initially (1) than the Tech Spec limit and level wasinitially (2) than the Tech Spec limit.

A. 1. higher2. higher

B. 1. lower2. higher

C. 1. higher2. lower

D. 1. lower2. lower

Page 7 of 100

McGuire Nuclear Station• 2011 MNSSRONRCExaininationQuestion: 8

‘1 point)


• The unit is operating at 100% RTP• The following annunciators are received:

1AD-6 D-9 (PRT HI PRESS)1AD-6 E-9 (PRT ABNORMAL LEVEL)

• The following OAC Alarms are received:

UI PZR RELIEF TANK LEVEL HI-HIUi PZR RELIEF TANK PRESS HI

• PRT pressure is 9 PSIG and RISING SLOWLY• PRT level is 95% and RISING SLOWLY

The PRT rupture disc is designed to discharge to Containment when PRT pressurerises to a MAXIMUM of (1)

To prevent PRT rupture disc operation the operating crew must (2)


A. 1. 85PSIG2. drain the PRTto the NCDT

B. 1. 100PSIG2. drain the PRT to the NCDT

C. 1. 85PSIG2. have Chemistry vent the PRT to the Waste Gas system

D. 1. 100 PSIG2. have Chemistry vent the PRT to the Waste Gas system

Page8oflOO


Question: 91 point)

Unit I was operating at 100%. Given the following conditions and sequence of events:

• A plant transient results in a cooldown of the NC system causes the “BackupHeaters to Energize” and the “C” PZR heaters to be full “ON”

• Pressurizer Pressure Channel I fails offscale high

Subsequently, the following events occur:

• Pressurizer Pressure Channel 2 fails offscale low• The following annunciators are received simultaneously with the second

pressure channel failure:

o IAD-2 I E8 (DCS Trouble)o IAD-2 I F8 (DCS Alternate Action)

• The OATC notifies the rest of the crew that there is an Alternate Action onPressurizer Pressure Select 1 and Select 2

Assuming no operator actions, what effect does this combination of channel failureshave on the Pressurizer Pressure Control system?

A. The Pressurizer Pressure Master swaps to manual.Backup heaters deenergize.

B. The Pressurizer Pressure Master swaps to manual.Backup heaters remain energized.

C. The Pressurizer Pressure Master remains in automatic.Pressure is controlled based on the average of the 2 unaffected channels.

D. The Pressurizer Pressure Master remains in automatic.Pressure is controlled based on the highest of the 2 unaffected channels.

Page 9 of 100

McGu ire Nuclear Station2011 MNS SRO NRC Examination


If the indication for a Power Range Nuclear instrument fails low, it could be caused byalossof (1)

If either Instrument OR Control Power is lost to Power Range N-42 at 100% RTP, thecrew will implement (2)


A. 1. Instrument Power2. E-0 (Reactor Trip or Safety Injection)

B. 1. Instrument Power2. AP-16 (Malfunction of Nuclear Instrumentation)

C. 1. Control Power2. E-0 (Reactor Trip or Safety Injection)

D. 1. Control Power2. AP-16 (Malfunction of Nuclear Instrumentation)

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Given the following on Unit 1:

• Unit 1 is operating at 100% RTP

The OTAT trip setpoint for N-41 will decrease if (1)

The OTAT reactor trip is design to protect against exceeding (2)


A. 1. NC system pressure increases2. DNB limits

B. 1. NC system pressure increases2. fuel heat generation rate limits

C. 1. power is lost to N-41 lower detector2. DNB limits

D. 1. power is lost to N-41 lower detector2. fuel heat generation rate limits

Page 11 of 100

McGuire Nuclear Station• 2UiIMNSSRONRCExamination

Question: 12M point)


• The unit is at 100% RTP• Pressurizer Pressure channel II is in test with all bistables in the tripped position

for IAE testing• 2EKVA develops a fault and is now de-energized

Which ONE (1) of the following describes the effect on the plant due to the fault on2EKVA?

A. The reactor will automatically trip.SI signal equipment on both trains will automatically start due to redundantpower supplies.

B. The reactor will automatically trip.SI signal equipment on ‘A’ train ONLY will not automatically start due to theloss of power to the SSPS output cabinet.

C. The reactor will not automatically tripIf an SI signal is received, equipment on both trains will automatically start dueto redundant power supplies.

D. The reactor will not automatically trip.If an SI signal is received, equipment on ‘A’ train ONLY will not automaticallystart due to the loss of power to the SSPS output cabinet.

Page 12 of 100

McGuire Nuclear Station2011 MNSSRUNRCExammnation


The normal logic for a High Containment pressure Safety Injection actuation is(1) channels.

If one of the Containment pressure channels that provides input to the Safety Injectionlogic is placed in BYPASS, then the logic becomes (2) channels.


A. 1. 2of32. lof2

B. 1. 2of42. lof3

C. 1. 2of32. 2of2

D. 1. 2of42. 2of3

Page 13 of 100

McGuire Nuclear Station• 2011 MNS SRO NRC ExaminütionQuestion: 14

(1 point)


• Unit was initially operating at 100% RTP• Train B components are initially in service• A LOCA occurs inside Containment• Containment pressure peaks at 3.2 PSIG and stabilizes at 1.5 PSIG

Based on these conditions Lower Containment Ventilation Cooling is

_________

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

A. isolated

B. being supplied by the RV system ONLY

C. being supplied by the RN system ONLY

D. being supplied the RVAND RN systems

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McGuire Nuclear Station• 2U11 MNS SRO NRC Ecamlnatioh

Question: 151 point)

Given the following plant conditions:

• Uniti wasoperatingatlOO%RTP• Steamline Break INSIDE Containment occurs• “NS System CPCS Train A Inhibit” status light, located on Status Indication

Panel ISI-12, is (LIT)• Containment Pressure is 8 PSIG• The Crew is performing FR Z.1 (Response to High Cont Press)

1) What effect would the condition described above have on the operation of the NSSystem?

2) What is the basis for the minimum temperature of the FWST as it relates to a LOCAContainment Spray actuation?

Which ONE (1) of the following answers the questions above?

A. 1. There would be no “Train A” Containment Spray Flow2. To prevent exceeding the maximum external pressure load on the

containment shell due to an internal vacuum caused by overcooling of thePost-Accident Containment atmosphere.

B. 1. There would be no “Train A” Containment Spray Flow2. To prevent an excessive reduction in containment pressure, which could

decrease the rate at which steam can be vented out the break andincreases peak clad temperature.

C. 1. ONLY one of the two “Train A” NS Spray Headers will have flow2. To prevent exceeding the maximum external pressure load on the

containment shell due to an internal vacuum caused by overcooling of thePost-Accident Containment atmosphere.

D. 1. ONLY one of the two “Train A” NS Spray Headers will have flow2. To prevent an excessive reduction in containment pressure, which could

decrease the rate at which steam can be vented out the break andincreases peak clad temperature.

Page 15 of 100

McGu ire Nuclear Station• 2011 MNS SRO NRC Examination


Given the following events and conditions on Unit 1:

• Unitisholdingat3o% RTP• Control rods are in manual• Turbine control is “MW OUT”• One MSR 2nd stage relief valve fails open

1) What trends will occur for Tavg and reactor power 10 minutes after the leak starts?

2) Per AP-Ol (Steam Leak) what will be required to isolate the steam leak?

A. 1. Tavg is HIGHER - reactor power is LOWER2. Reactor Trip and Turbine Trip ONLY

B. 1. Tavg is HIGHER - reactor power is LOWER2. Reactor Trip and closure of the MSIV’s

C. 1. Tavg is LOWER — reactor power is HIGHER2. Reactor Trip and Turbine Trip ONLY

D. 1. Tavg is LOWER — reactor power is HIGHER2. Reactor Trip and closure of the MSIV’s

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McGu ire Nuclear Station• 2U11 MNS SRO NRC Examination



• A LOCA has occurred inside Containment• Containment pressure is 3.4 PSIG• The crew is preparing to initiate a cooldown per ES 1.2 (Post LOCA Cooldown

and Depressuration)

Which ONE (1) of the following must occur to allow reopening the MSIV’s for the givenconditions?

A. Reset the Main Steam Isolation signal ONLY.

B. Reset the Phase BAND Main Steam Isolation signals ONLY.

C. Containment pressure must be reduced below 3 PSIG AND reset the MainSteam Isolation signal ONLY.

D. Containment pressure must be reduced below 3 PSIG AND reset BOTH theMain Steam Isolation signal and Phase B Isolation signal.

Page 17 of 100

McGuire Nuclear Station2OJIMNS SRO NRCExaminátion


Which ONE (1) of the following lists automatic actions that occur as a result of aP-14 (Feedwater Isolation) signal?

A. Both Feedwater Pumps tripCF to CA Nozzle Isolations close

B. Both Feedwater Pumps tripMain Steam Isolation Valves close

C. Both Feedwater Pumps speed goes to 2800 RPMCF to CA Nozzle Isolations close

D. Both Feedwater Pumps speed goes to 2800 RPMMain Steam Isolation Valves close

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McGuire Nuclear Station• 2U1JMNSSRONRCExam1natIonQuestion: 19

‘1 point)


• A reactor trip from 100% power has occurred• Bus 1ETA locked out• The TD CA Pump tripped on overspeed upon starting

CA is being supplied to (1) at a maximum flow rate of (2) ?

Which ONE of the following completes the statement above?

A. 1.AandBSGs2. 450 GPM

B. 1. CandDSGs2. 450 GPM

C. 1.AandBSGs2. 900 GPM

D. 1. CandDSGs2. 900 GPM

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McGuire Nuclear Station• 2011 MNS SRO NRC ExaminationQuestion: 20

(1 point)


• Bus 1EVDA has lost power

The breakerfor 1A NV pump

______


A. automatically trips open but can be closed locally after manually charging thesprings

B. can only be operated locally but the springs must be manually charged prior toeach opening and closing operation

C. can only be operated locally but only closing operations require manuallycharging the springs

D. can be opened one time from the control room but all further operations mustbe performed locally

Page 20 of 100


Question: 21I point)


• Both Units are operating at 100% RTP• Battery 1DP is aligned for equalizing charge• The AC input breaker for Charger 1 DP has tripped open

Based on the conditions above, which ONE (1) of the following describes the currentsource(s) of power for Bus IDP?

A. Charger 1DS and Battery IDP ONLY

B. Charger 2DS and Battery 2DP ONLY

C. Charger2DP and Battery2DP ONLY

D. Charger 2DS, Charger 2DP, AND Battery 2DP

Page 21 of 100




. 125 VDC Control Power for the 1 B DG has been lost

Which ONE (1) of the following lists equipment associated with the lB DG that will nolonger have power?

A. DG Sump Pump AND Starting Air Solenoids

B. Electronic Governor AND Starting Air Solenoids

C. DG Sump Pump AND Diesel Building dampers

D. Electronic Governor AND Diesel Building dampers

Page 22 of 100

McGu ire Nuclear Station• 2(111 MNS SRO NRC Examination



. The 1A DG was started in MANUAL MODE at time 14:04 for surveillance

The following trend of DG Crankcase Vacuum is observed:

DG 1A Crankcase Vacuum

TI ME

14:00 14:15 14:30 14:45

00O.5

E[1

dJ 1.5

I c 2 —

L)

2.5 -

Based on the trend above, the EARLIEST time that IA DG will trip on Low CrankcaseVacuum is...

A. 14:10

B. 14:15

C. 14:20

D. 14:25

Page 23 of 100

McGuire Nuclear Station2OI1MNSSRONRCExam1naIion



• A MANUAL MODE start of the 1 B DG was performed for surveillance testing• The DG is currently loaded to 4000 KW

1) Which ONE (1) of the following is a heat load that is currently being cooled by theDG Cooling Water (KD) System?

2) Which ONE (1) of the following is the MIMIMUM Surge Tank level that was requiredto perform the MANUAL MODE start?

A. 1. VG After Coolers2. 11.5 inches

B. 1. Air Intake System Intercooler2. 25 inches

C. 1. Air Intake System Intercooler2. 11.5 inches

D. 1. VG After Coolers2. 25 inches

Page 24 of 100

McGu ire Nuclear Station2011 MNS SRO NRC Exambwtioñ


With Unit I at 100% RTP, a SIG tube leaks develops on the lB Steam Generator.

The following radiation monitors start to slowly increase:

• IEMF33 (Unit I Condenser Air Ejector Monitor)• 1EMF34(L) (Unit I Steam Generator Sample (Low Range))• 1EMF34(H)(Unit I Steam Generator Sample (High Range))• IEMF72 (Unit I Steam Line lB Monitor)

Subsequently, both 1EMF33 and 1EMF34(L) reach their TRIP 2 setpoints.

Assuming NO action is taken by the crew, which ONE (1) of the following identifies themonitor or monitors that are expected to DECREASE over the next several minutes?

A. IEMF33ONLY

B. IEMF34(L) ONLY

C. 1EMF34(L) and 1EMF 34(H) ONLY

D. IEMF33, 1EMF34(L) and 1EMF34(H)

Page 25 of 100


1 point)


• A loss of power had occurred on 2ETA and 2ETB• The 2A and 2B DG started and all loads sequenced normally

Based on the conditions above, which ONE (1) of the following loads is currently beingsupplied by the Nuclear Service Water (RN) system?

A. Safety Injection Pump motor cooler

B. Containment Spray heat exchanger

C. Auxiliary Feedwater Pump motor cooler

D. Residual Heat Removal Pump motor cooler

Page 26 of 100

McGuire Nuclear Station• 2011 MNS SRUNKCExamination


Given the following initial conditions on Unit 1:

• A LOCA has occurred inside Containment• NC pressure is 1700 PSIG

With Containment pressure at 2.5 PSIG and increasing rapidly, the crew MANUALLYinitiates Containment Spray.

Assuming all automatic and manual actions have functioned normally, which ONE (1)of the following lists the ESF actuation(s) that has/have occurred?

A. PhaseAONLY

B. PhaseAand Phase B ONLY

C. Phase A and Containment Ventilation Isolation ONLY

D. Phase A, Phase B, and Containment Ventilation Isolation

Page 27 of 100

McGu ire Nuclear Station2O1IMNSSRUNRCExamInatIon

Question: 28M point)

In accordance with FR-Z.1 (Response to High Containment Pressure), to place NDSpray in service Containment pressure must be greater than a MINIMUM of (1)AND the time since the Reactor Trip must be greater than a MINIMUM of (2)


A. 1. 3PSIG2. 50 minutes

B. 1. 1OPSIG2. 50 minutes

C. 1. 3PSIG2. 60 minutes

D. 1. 1OPSIG2. 60 minutes

Page 28 of 100



Given the following:

• Unit 2 was initially operating at 100% RTP• Due to Main condenser cooling water issues, the Main Turbine controls are

placed in manual and load reduced by 400 MW over a 5 minute period

Which ONE (1) of the following describes the INITIAL effect on the pressurizer and theinherent response to the transient described above?

A. PZR outsurgePZR Backup heaters would energize to limit the pressure decrease.

B. PZRoutsurgeSome of the liquid in the PZR would flash to steam which will limit the pressuredecrease.

C. PZRinsurgeSome of the steam volume in the PZR would condense limiting the pressureincrease.

D. PZR insurgePZR Backup heaters would de-energize to limit the pressure increase.

Page 29 of 100

McGuire Nuclear Station• 2011 MNS SRO NRUffxainiithtionQuestion: 30

1 point)


• A Loss of Offsite Power has occurred• 1ETA and 1ETB are energized from their respective DGs

Based on the conditions above, power can be restored to Pressurizer Heater Group(s)(1)

If control is transferred to the Auxiliary Shutdown Panel (ASP), the Pressurizer Heatersoperated from the ASP would trip on a (2)


A. 1.A&BONLY2. Safety Injection

B. 1.A&BONLY2. Pressurizer Low-Low Level

C. 1.A,B1&D2. Safety Injection

D. 1.A,B,&D2. Pressurizer Low-Low Level

Page 30 of 100

McGuire Nuclear Station• 2011 MNS SRU NRC ExaminatIonQuestion: 31

‘1 point)

Given the following indications:

• Unit I is operating at 100% RTP• An electrical fault occurred in the DRPI circuitry located in the CIR• The fault has affected the DRPI programming causing Rod H-8 to indicate 200

steps with the remainder of Control Bank D at 216 steps• The “gray” code from the DRPI Data Panels has NOT been affected• ACTUAL ROD position for Rod H-8 is 216 steps

The RPI (1) Failure annunciator would be in LIT.

The OAC “Rods” program (2) indicate the correct rod position for Rod H-8.

Based on the conditions described, which ONE (1) of the following completes thestatements above?

A. 1. Non-Urgent2. will

B. 1. Non-Urgent2. will not

C. 1.Urgent2. will

D. 1.Urgent2. will not

Page3l of 100

McGuire Nuclear Station2011 MNS SRONRCExan,inatkinQuestion: 32

1 point)


• The unit is operating at 100% RTP• Rod Control is in AUTOMATIC

Subsequently, rod control is placed in MANUAL due to a DCS Afternate Action.

1) Which ONE (1) of the following describes the MINIMUM number of Loop Tavedetector failures required to cause a DCS Alternate Action on the Tavg ReactorTemperature Control unit?

2) In accordance with 0P111A161001003 (Controlling Procedure for Unit Operation),after DCS is repaired, what is the MAXIMUM deviation between Tavg and Tref prior toshifting CRD Select from MANUAL to AUTO?


A. 1.22. ±1°F

B. 1.22. ±3°F

C. 1.32. ±1°F

D. 1.32. ±3°F

Page 32 of 100

McGu ire Nuclear Station2011 MNS SRO NRC ExamthationQuestion: 33

1 point)


• Unit is Mode 6• VP is in service and refueling is in progress• The CIR receives the following alarms and indications

o VP Supply Damper “CLOSED”o VP Exhaust Damper “CLOSED”o VP Supply Fans “OFF”o VP Exhaust Fans “OF F”

• The Containment evacuation alarm is reported to be sounding

A Trip 2 alarm on which ONE (1) of the following would have caused ALL of thealarms and indications described above?

A. 1 EMF-35(L) (Unit Vent Particulate)

B. 1EMF-36(L) (Unit Vent Gas)

C. 1 EMF-38(L) (Containment Particulate)

D. I EMF-39(L) (Containment Gas)

Page 33 of 100

McGuire Nuclear Station2U11 MNS SRO NRCExmilnation



• The unit is operating at 100% RTP• The “SPENT FUEL POOL LEVEL LO” alarm is received on the Unit 2 OAC

1) Which ONE (1) of the following is the FIRST EMF that will alarm to confirm a leakon the Spent Fuel Pool Cooling system?

2) Which ONE (1) of the following AUTOMATIC actions associated with FuelHandling Building Ventilation (VF) system is caused by a Spent Fuel Pool radiationmonitor?

A. 1. 2EMF-42 (Unit 2 FUEL BUILDING VENTILATION)2. The VF Supply and Exhaust Fans will stop on a 2EMF-4 Trip 2.

B. 1. 2EMF-4 (SPENT FUEL BLDG REFUEL BRDG)2. The VF Supply and Exhaust Fans will stop on a 2EMF-42 Trip 2.

C. 1. 2EMF-42 (Unit 2 FUEL BUILDING VENTILATION)2. The Exhaust Filter Bypass Damper (D-5) will close on a 2EMF-4 Trip 2.

D. 1. 2EMF-4 (SPENT FUEL BLDG REFUEL BRDG)2. The Exhaust Filter Bypass Damper (D-5) will close on a 2EMF-42 Trip 2.

Page 34 of 100

McGuire Nuclear Station2tfll MNS SRUNRCExamination


Which ONE (1) of the following combinations describes the MINIMUM ND requirementsand additional License requirements for Fuel Movement?

ND Additional Requirements

A. One Loop OPERABLE and Reactor subcritical for a minimum ofin operation 70 hours

B. Two Loops OPERABLE and A minimum of one door in each airlockONE in operation closed

C. One Loop OPERABLE and Equipment Hatch held in place by ain operation minimum of four bolts

D. Two Loops OPERABLE and One OPERABLE VCJYC trainONE in operation

Page 35 of 100

McGuire Nuclear Station21111 MNS SRONRCExaminationQuestion: 36

1 point)


• The unit was initially operating at 100% RTP• A SGTR occurs on 2C SG• The crew initiated a manual Reactor Trip and Safety Injection• The 2C SG MSIV would NOT close during isolation• E-3 (Steam Generator Tube Rupture) has been implemented• The crew is preparing to cooldown to the target temperature of 520°F

Which ONE (1) of the following describes how the plant coold own to the targettemperature will be accomplished?

A. Steam Dumps

B. SM PORVs from all intact S/Gs

C. SM PORVS from 2A and 2D SGs ONLY

D. Start the TDCA pump and align the steam line drains

Page 36 of 100

McGu ire Nuclear Station• 2011 MNS SRONRC ExaminationQuestion: 37

1 point)

Given the following related to Waste Gas (WG) Decay Tank “A”:

• H2 concentration is 4.4%

Per station Selected License Commitment (SLC) 16.11 .19 (Explosive Gas Mixture):

WG Decay tank 02 concentration must be maintained less than a MAXIMUM of(1) until H2 concentration is less than a MAXIMUM of (2)


A. 1.2%2. 2%

B. 1.4%2. 4%

C. 1.2%2.4%

D. 1.4%2. 2%

Page 37 of 100

McGu ire Nuclear Station• 2011 MNS SRONRCExaminätionQuestion: 38

‘1 point)


• A leak developed in the VS system• VI pressure dropped to 80 PSIG before the leak was isolated• VI pressure has returned to 100 PSIG

Assuming NO operator action, 1VI-820 (VI Supply to VS Control) auto-closed when VIpressure reached a MINIMUM of (1) AND currently is (2) ?

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

A. 1. 85P51G2. open

B. 1. 85PSIG2. closed

C. 1. 9OPSIG2. open

D. 1. 9OPSIG2. closed

Page 38 of 100

McGuire Nuclear Station• 2tfli MNS SRO NRC ExamimaionQuestion: 39

I point)


• A Reactor Trip has occurred• E-O (Reactor Trip or Safety Injection) has been implemented

The crew will FIRST determine if the turbine has automatically tripped by checking all(1) valves closed.

In accordance with E-O, if the turbine has not automatically tripped and can NOT bemanually tripped, the crew will NEXT attempt to (2)


A. 1. throttle2. close the MSIVs AND MSIV bypasses

B. 1. governor2. place the turbine in MANUAL AND close the governor valves

C. 1. throttle2. place the turbine in MANUAL AND close the governor valves

D. 1. governor2. close the MSIV5 AND MSIV bypasses

Page 39 of 100

McGuire Nuclear Station2OlIMNSSRONRCExainination



• The unit is in MODE 3 with a plant cooldown in progress• The crew has entered AP-1 I (Pressurizer Pressure Anomalies) due to

Pressurizer pressure decreasing very slowly• Pressurizer pressure is 1185 PSIG• PRT pressure is 5 PSIG

Given the above conditions, determine which ONE (1) of the following would indicatea leaking PORV and the state of the fluid in the PORV discharge?

REFERENCE PROVIDED

PORV Discharge Temperature State of the Effluent

A. 280-300°F Wet Vapor

B. 300-320°F Wet Vapor

C. 280-300°F Superheated Steam

D. 300-320°F Superheated Steam

Page 40 of 100

McGuire Nuclear Station• 2011 MNS SRUNRCExaminätion



• The unit has experienced a Reactor Trip and Safety Injection due to aSmall-Break LOCA

• The crew has just completed the actions of E-O (Reactor Trip or SafetyInjection)

• NV pump flow to the NC system Cold Legs is 390 GPM• NC system pressure is 1300 PSIG and stable• SG pressures are 1092 PSIG and stable• NC system subcooling on the ICCM is 22°F and stable

Which ONE (1) of the following describes plant conditions upon transition toE-1 (Loss of Reactor or Secondary Coolant)?

A.

B.

C.

D.

SG5NC Pumps Required forRunning? Heat

Removal?

YES YES

YES NO

NO YES

NO NO

Page 41 of 100

McGuire Nuclear Station2011 MNS SRO NRC ExaminationQuestion: 42

1 point)


• The unit has experienced a Large-Break LOCA• The crew has implemented ECA-1.1 (Loss of Emergency Coolant Recirculation)• NCS subcooling is 0 degrees• FWSTlevelis3%• NCPs are not running• There is no indication of natural circulation

The core is currently being cooled by steam entering the (1) of SIG U-tubeswhere the steam condenses and re-enters the core area via the SIG (2)


LflA. hot leg cold leg

B. hot leg hot leg

C. cold leg hot leg

D. cold leg cold leg

Page 42 of 100

McGuire Nuclear Station• 2IlIlMNSSRONRCExaminatiönQuestion: 43

‘1 point)


• The unit is in MODE 3 with Shutdown Banks withdrawn• 2B, 2C, and 2D NC pumps are in service• Bus 2TA is de-energized with its supply breaker racked out for maintenance

Subsequently, the 2B NC pump 6.9KV NC pump feeder breaker trips.

Based on the conditions above, the 2C and 2D NC pumps SAFETY BREAKERSindicate (1)

The Unit 2 Reactor Trip breakers (2)


A. 1. CLOSED2. are TRIPPED

B. 1. CLOSED2. remain CLOSED

C. 1. OPEN2. are TRIPPED

D. 1. OPEN2. remain CLOSED

Page 43 of 100

McGuire Nuclear Station• 2011 MNS SRUNRUExaminationQuestion: 44

1 point)

Given the following events and conditions:

• Unit I was operating at 96% power• 1NV-241 (Seal Water Injection Flow) failed shut• The operators entered AP-12 (Loss of Charging)• Excess LID has been placed in service

With 1 NV-26B (U 1 Excess LID Hx Outlet Cntrl) fully open maximum indicated flowwould be (1)

INV-94AC (NC Pump Seal Ret Cont Inside Isol) can be controlled from the C/RANDthe (2)


A. 1. 26GPM2. SSF

B. 1. 26GPM2. AuxSID Panel

C. I. 35GPM2. SSF

D. 1. 35GPM2. Aux S/D Panel

Page 44 of 100

McGu ire Nuclear Station• 21111 MNS SRO NRC ExaminationQuestion: 45

1 point)

Given the following initial conditions:

• Unit I is in MODE 5 and drained to mid-loop for NC Pump seal repair• ND Train IA is in service• ND system flow rate is 3200 GPM• NC System level is ÷10 inches• Current (amps) indication for the 1A ND pump begins to fluctuate

ThelANDpumpis (1)

In accordance with AP-19 (Loss of ND or ND System Leakage) NC system subcooling isdetermined by monitoring (2)


A. 1. cavitating2. Core ExitT/C’s

B. 1. cavitating2. Wide Range T-hot

C. 1. operating at a runout condition2. Core ExitT/C’s

D. 1. operating at a run out condition2. WideRangeT-hot

Page 45 of 100

McGuire Nuclear Station2011 MNSSRONRCExamination


Unit I was operating at 100% RTP when a pipe break occurred on the ID SIC steamheader.

The following sequence of events occurred:

• The 1D SIC is isolated• Pzr level dropped to 0% and was restored to 20%• NC system pressure is 1900 PSIG• The Safety Injection and Sequencers have been reset• ES-I .1(SI Termination) has been implemented

Which ONE (1) of the following describes the correct MCB panel actions to re-energizethe PZR Back-up Heater Bank B?

A. Select MANUAL on the heater mode switch on 1MC-5 (vertical section);Select CLOSED on the heater breaker switch on 1MC-10 (horizontal section);Select ON for the heater control switch on 1 MC-5 (vertical section).

B. Select MANUAL on the heater mode switch on IMC-10 (horizontal section);Select ON for the heater control switch on I MC-1 0 (horizontal section).

C. Select MANUAL on the heater mode switch on 1MC-5 (vertical section);Select ON for the heater control switch on I MC-5 (vertical section).

D. Select MANUAL on the heater mode switch on 1MC-10 (horizontal section);Select CLOSED on the heater breaker switch on 1MC-5 (vertical section);Select ON for the heater control switch on 1 MC-1 0 (horizontal section).

Page 46 of 100

McGu ire Nuclear Station• 2Ol1MNSSROi’TRCExaminàtionQuestion: 47

I point)


• The crew determines that a valid automatic Reactor Trip has failed to occur• A manual Reactor Trip is attempted unsuccessfully• FR-S.1 (Response to Nuclear Generation/ATWS) is implemented• The crew manually trips the Main Turbine• NC system pressure begins to increase rapidly

In accordance with Tech Spec 2.1.2 (RCS Pressure Safety Limit):

• NC system pressure shall be maintained less than or equal to a MAXIMUMpressure of (1) PSIG.

• In accordance with Tech Spec 2.1.2 Basis, the (2) are credited forpreventing the RCS from exceeding the Design Pressure Safety Limit.


A. 1. 25002. Main Steam Safety Valves

B. 1. 27352. Main Steam Safety Valves

C. 1. 25002. Condenser Steam Dumps

D. 1. 27352. Condenser Steam Dumps

Page 47 of 100

McGuire Nuclear Station• 2011 MNSSRO NRCExaminatiimQuestion: 48

1 point)


• A SGTR has occurred on the 2D SG• A Steam Line break occurred on the 2B SG• Containment pressure peaked at 3.1 PSIG and is now 0.9 PSIG and STABLE

In accordance with E-3 (Steam Generator Tube Rupture):

• 2D SG NR level shall be maintained greater than a MINIMUM of (1)

• During steps to depressurize the NC system, if it is determined that PressurizerSprays are ineffective at reducing NC system pressure, the crew shall FIRSTattempt to depressurize the NC system using (2)


A. 1. 11%2. Auxiliary Spray

B. 1. 32%2. Auxiliary Spray

C. 1. 11%2. one PZR PORV

D. 1. 32%2. one PZR PORV

Page 48 of 100

McGuire Nuclear Station• 2OITMNS SRONRUExaminätion



• Unit 2 is in Mode 3 performing a plant cooldown• NC System pressure is 1300 PSIG• All S/G pressures are 800 PSIG• Low Pressure SI and Low Pressure Steamline Isolation have been blocked

Subseq uently:

• A steamline rupture occurs inside containment from the 2A S/G• Containment pressure peaked at 1.5 PSIG• All MSIV’s indicate closed

Based on the conditions above, the Main Steam Isolation Valves closed because SGpressure decreased a MINIMUM of 100 PSIG in (1)

In addition to the closure of the MSIV’s, the (2) will also receive a signal to close.


A. 1. 1 second2. MSIV bypass valves ONLY

B. 1. 1 second2. MSIV bypass valves AND the S/G PORVs

C. 1. 2seconds2. MSlVbypass valves ONLY

D. 1. 2seconds2. MSIV bypass valves AND the S/G PORVs

Page 49 of 100

McGuire Nuclear Station2011 MNS SROJVRCExaminationQuestion: 50

I point)


• A unit shutdown is in progress• 0200 both Main Feedwater pumps trip

Subsequently, the following conditions are observed:

TIME

CONDITION 0200 0205 0210 0215

NCS Temp (°F) 557 558 558 559

NCSPress(PSIG) 1965 1960 1976 1991

NRSGA(%) 19 18 19 19

NRSGB(%) 20 18 17 16

NRSGC(%) 20 19 18 16

NRSGD(%) 18 16 18 19

Based on the conditions above:

The EARLIEST time that the MD CA pumps will be running is (1)

The EARLIEST time that the TD CA pump will be running is (2)


A. 1. 02002. 0205

B. 1. 02052. 0205

C. 1. 02002. 0215

D. 1. 02052. 0215

Page 50 of 100

McGuire Nuclear Station• 2011 i1INS SRO NRCExamiñätion


Given the following initial conditions:

• Unit I and 2 have been in a Loss of Offsite Power condition for 10 hours• BOTH Diesel VI compressors are unavailable• ECA-0.0 (Loss of All AC Power) was implemented on Unit 1 and the crew has

transitioned to ECA-0.1 (Loss of All AC Power Recovery Without S/I Required)

Current conditions:

• NC Th0ts are STABLE• S/G pressures are STABLE at 775 PSIG• S/G levels are decreasing and approaching 11% NR• NC T-colds are 480°F• VI headerpressureisO PSIG

Based on the indication above, which ONE (1) of the following describes the currentstate of Natural Circulation flow and the actions that must be taken per ECA-0.1?

REFERENCE PROVIDED

A. Natural Circulation flow has been established.Increase CA flow using flow controllers in the control room.

B. Natural Circulation flow has been established.Increase CA flow by notifying NLO to throttle CA valves locally.

C. Natural Circulation flow has NOT been established.Increase dumping steam using SM PORV controller on main control boardAND increase CA flow using flow controllers in the control room.

D. Natural Circulation flow has NOT been established.Dispatch an operator to locally increase flow from the SM PORVAND increase CA flow by notifying NLO to throttle CA valves locally.

Page 51 of 100

McGuire Nuclear Station- 2UHMNSSRONRCEaminatiôñ



• The unit is in MODE 3 preparing for a reactor startup• At 0400 on February 1, 1 B NV pump is tagged for maintenance

Subsequently:

• At 0800 hours on February 1 a fault occurs on Main Transformer 1A• Bus 1TA automatically transfers to Auxiliary Transformer IATB

Based on the conditions above, which ONE (1) of the following describes the requiredactions in accordance with Tech Spec 3.8.1 (AC Sources - Operating)?

REFERENCE PROVIDED

A. Declare the IA NV pump INOPERABLE by 0400 on February 2 ONLY.

B. Declare the IA NV pump INOPERABLE by 0800 on February 2 ONLY.

C. Restore Off-Site Power via the 1A Main Transformer by 0800 on February 4ANDDeclare the IA NV pump INOPERABLE by 0400 on February 2.

D. Restore Off-Site Power via the 1A Main Transformer by 0800 on February 4ANDDeclare the IA NV pump INOPERABLE by 0800 on February 2.

Page 52 of 100

McGuire Nuclear Station2OIIMNSSRONRCExámination



• The unit was operating at 100% RTP• A Loss of All AC Power occurs• ECA-0.0 (Loss of All AC Power) has been implemented

Subsequently, the following events occur:

• Both DGs have been started• Power has been restored to IETA and IETB

In accordance with ECA-0.0, the vital battery chargers must be re-started within aMAXIMUM of (1) minutes to prevent (2)


A. 1. 40 minutes2. failure of components supplied by the battery

B. 1. 60 minutes2. failure of components supplied by the battery

C. 1. 40 minutes2. complete loss of the vital battery

D. 1. 60 minutes2. complete loss of the vital battery

Page 53 of 100

McGuire Nuclear Station2011 MNSRO NRC ExaminationQuestion: 54

I point)

Given the following indications on Unit 1:

• The Unit is operating at 100% RTP• 1AD-8 D-l (Sump A Groundwater Drainage Hi Hi Lvl) just came in alarm• The crew is attempting to determine the source of the input to the sump• An NEO sent to investigate reports that the sump is overflowing

Which ONE(1) of the following describes the source of the flooding?

A. 1 B RN strainer basket shaft seal failure

B. 2A RN Pump Suction piping weld failure

C. RF piping break in the Unit 2 CA pump Room

D. IA RN strainer automatic backwash valve has failed open

Page 54 of 100

McGuire Nuclear Station• 2011MNS SRO CExaiinàtioiQuestion: 55

I point)


• VI pressure has been lost due to a pipe rupture• AP-22 (Loss of VI) has been implemented

Which ONE (1) of the following describes the RN alignment directed by AP-22 and thereason for those actions?

A. The RB” Train is aligned to the SNSWPTo prevent pump runout due to high flow conditions

B. The “B” Train is aligned to the SNSWPTo prevent cavitation due having all 4 RN pumps suction aligned to LLI

C. The “A” Train is aligned to the SNSWPTo prevent pump runout due to high flow conditions

D. The “A” Train is aligned to the SNSWPTo prevent cavitation due having all 4 RN pumps suction aligned to LLI

Page 55 of 100

McGuire Nuclear Station• 2011 MNS SRONRCExaminatlonQuestion: 56

I point)

Given the following condition associated with Unit 1:

• SIG’s 1A & lB are faulted and indicating <5% WR Level• SIG’s IC & 1D are indicating 30% NR Level• All CA is unavailable• Containment Pressure is 3.5 PSIG• E-0 (Reactor Trip or Safety Injection) has been completed

Based on the indications described above, which ONE (1) of the following describesthe NEXT procedure to be implemented and action(s) required?

A. Go to EP-E2 (Faulted SIG Isolation)Isolate C & D S/G’s

B. Go to FR-HA (Response to Loss of Secondary Heat Sink)Commence NCS feed and bleed

C. Go to FR-H.l (Response to Loss of Secondary Heat Sink)Restore feed water flow to C & D S/C’s

D. Go to EP-E2 (Faulted S/C Isolation)Close all MSIVs and MSIV bypasses

Page 56 of 100

McGuire Nuclear StationZO1IMNS SEQ NRCExaminatlon



• The unit is operating at 70% RTP• Rod control is in AUTOMATIC• Annunciator 1AD-2 I D9 (RPI AT BOTTOM ROD DROP) alarms• DRPI indicates that rod M14 (adjacent to Power Range N-44) has dropped

The dropped rod will be confirmed by the following indications:

• Control rods will move OUT f Tave decrease below Tf by a MINUMUM of(1)

• Over the next several hours the overall core QPTR will (2)

Which ONE (1) of the following corn pletes the statements above?

A. 1. 1°F2. decrease

B. 1. 1.5°F2. decrease

C. 1. 1°F2. increase

D. 1. 1.5°F2. increase

Page 57 of 100

McGuire Nuclear Station2011 MNS SRO NRCExamination


If an Emergency Boration flowpath via I NV-265B (Boric Acid to NV Pumps) can NOTbe established:

• AP-38 directs the Operators to (1)

• FR-S.1 directs (2)


A. 1. OPEN I NV-269 (BA to Charging Pumps Block) ONLY

2. manually aligning the NV pumps suction to the FWST

B. 1. OPEN 1NV-269 (BA to Charging Pumps Block) AND INV-267A (BoricAcid To Blender Control).

2. manually aligning the NV pumps suction to the FWST.

C. 1. OPEN 1 NV-269 (BA to Charging Pumps Block) ONLY

2. manually initiating Safety Injection to align the NV pumps suction to theFWST.

D. I. OPEN I NV-269 (BA to Charging Pumps Block) AND 1NV-267A (BoricAcid To Blender Control).

2. manually initiating Safety Injection to align the NV pumps suction to theFWST.

Page 58 of 100

McGuire Nuclear Station• - 2OHMNSSRONRCExamInàtIbnQuestion: 59

‘1 point)


• A Ventilation Unit Condensate Drain Tank (VUCDT) release to the RCDischarge is in progress

• A 1 EMF-44(L) Trip 2 alarm is received

Which ONE (1) of the following describes the automatic response to the EMF alarm?

A. All Unit 1 VUCDT pumps trip ONLY.

B. 1WL-322B (Cont. Vent Drn Otsd Isol) closes ONLY.

C. IWL-320 (VUCDT Rad Monitor Outlet) AND 1WP-35 (WMT/VUCDT to RCCNTRL) close.

D. 1WL-320 (VUCDT Rad Monitor Outlet) AND IWL-322B (Cont. Vent Drn OtsdIsol) close.

Page 59 of 100


Question: 60point)

Given the following indication on Unit 1:

• Unit is operating at 100% RTP• 1EMF-18 (Reactor Coolant Filter IA) is in Trip 2 alarm• lEMF48 (Reactor Coolant Hi Rad) is also in Trip 2 alarm• The crew has implemented AP-l 8 (High Activity In Reactor Coolant)• Chemistry has reported that the cause of the high activity is due to FAILED

FUEL

Which ONE (1) of the following describes the action required per AP-1 8 and reason forthat action?

A. Increase LID to 120 GPMIncreases the removal rate of fission products resulting from the failed fuel.

B. Increase L/D to 120 GPMIncreases the effectiveness of the fission product gas removal by the VCT byincreasing the flow rate through the NV system.

C. Place cation bed demineralizer in serviceFacilitates the removal of fission products resulting from the failed fuel.

D. Place cation bed demineralizer in serviceThe cation bed will remove lithium causing a pH change that prevents furtherfuel degradation.

Page 60 of 100

McGu ire Nuclear Station• 2OIIMNSSRONRCExáminatiwiQuestion: 61

I point)

The following conditions exist on Unit 1:

• The crew is responding to a Small-Break LOCA• Due to a failure to establish adequate safety injection flow FR-C.2 (Response to

Degraded Core Cooling) has been implemented• NCP’s “B” and “C” are running

Which ONE (1) of the following RVLIS DIP conditions would indicate Degraded CoreCooling conditions exist?

REFERENCE PROVIDED

A TRAIN”A”=18%TRAIN “B” = 27%

B TRAIN”A”=25%TRAIN “B” = 27%

C. TRAIN”A”=23%TRAIN “B” = Failed

D. TRAIN “A” = FailedTRAIN “B” = 21%

Page 61 of 100

McGuire Nuclear Station•

2011 MNS SRO NRCExaiiiihationQuestion: 62‘1 point)

The NC inventory requirements for SI termination during performance of FR-P.1(Response to Imminent Pressurized Thermal Shock Condition) are (1)restrictive than the termination criteria during performance of E-O (Reactor Trip orSafety Injection).

One evolution that would be allowed DURING a “soak” in FR-P.1 is (2)


A. 1. less2. initiate an NC system cooldown

B. 1. more2. initiate an NC system cooldown

C. 1. less2. placing Auxiliary Spray in service

D. 1. more2. placing Auxiliary Spray in service

Page 62 of 100

McGu ire Nuclear Station2OlIMNSSROiVRCExamination


Unit 1 was in MODE 3 when a steam generator over pressure event occurred. Giventhe following events and conditions:

• The crew entered FR-H.2 (Response to SIG Overpressure)• lB SIG pressure is 1235 PSIG• 1BS/GNRlevelis95%• The 1A, 1C and 1 D SIG pressures are all 850 PSIG and 50% NR level• All feedwater isolation status lights are DARK

Which ONE (1) of the following statements describes the FIRST action to be takenper this procedure, and the reason for this action?

A. Open the lB S/G PORVto immediately reduce pressure in the lB S/G

B. Dump steam from the I B S/G using CA pump #1 to immediately reducepressure in the lB SIC

C. Manually isolate feedwater to the 1 B SIG to prevent additional feedwater fromfurther pressurizing the lB SIG

D. Dump steam from the 1A, 1C and 1D SIGs to reduce NC system temperatureand reduce pressure in the lB SIC

Page 63 of 100

McGuire Nuclear Station• 2011 MNSSRO NRCExaminaiidn



• A LOCA has occurred inside Containment• Containment Pressure indicates 4 PSIG and rising• NC Pressure indicates 1400 PSIG and decreasing• FWST level is 200 inches• The breaker for 2NS-3B, (Train ‘B’ NS Pump Suction valve from FWST)

has tripped, and cannot be reclosed

Which ONE (1) of the following describes the effect on Control Board operationsof the NS system?

A. 2B NS Pump can NOT be started.

B. ONLY 2NS-IB (Train ‘B’ Containment Sump Isolation Valve) can beopened.

C. ONLY 2Nl-1 84B (Train ‘B’ Containment Sump Isolation Valve) can beopened open.

D. 2N5-1 B AND 2N1-184B (Train ‘B’ Containment Sump Isolation Valves)can be opened.

Page 64 of 100

McGuire Nuclear Station• 2011 MNS SRO NRCExaminatibn



• A Unit 1 RO is performing PT111A146001003 A (Semi-Daily Surveillance ItemsChecklist)

• The operator observes that 1 EMF 38, 39 & 40 sample points are aligned toupper containment ONLY

1) Which ONE (1) of the following describes the MODE(s) where IEMF 38(L) isrequired to be capable of NC system leak detection?

2) What effect (if any) does the sample point alignment described above have on thesampling capability?

REFERENCE PROVIDED

A. 1. M0deIONLY2. Due to the limited amount of communication between atmosphere in upper

and lower containment, 1 EMF-38 will not be able to detect a I GPM NCsystem leak within one hour.

B. 1. M0deIONLY2. Due to the normal mixing of containment atmosphere by the ventilation

system, this does not significantly affect the ability of this sample packageto sample the containment atmosphere.

C. 1. Modes 1-42. Due to the limited amount of communication between atmosphere in upper

and lower containment, 1 EMF-38 will not be able to detect a 1 GPM NCsystem leak within one hour.

D. 1. Modes 1-42. Due to the normal mixing of containment atmosphere by the ventilation

system, this does not significantly affect the ability of this sample packageto sample the containment atmosphere.

Page 65 of 100

McGuire Nuclear Station• 2011 MNSSRONRCExamination

Question: 66i point)


• Maintenance is performing PLANNED work on equipment in the Unit 2 TurbineBuilding

• The work results in an OAC Priority Alarm (below the line) that the OATC hasseen before during this particular maintenance activity

• The alarm was not specifically discussed prior to the work starting and this isthe first time the alarm has been received during the shift

Per OMP 2-2 (Conduct of Operations) which ONE (1) of the following would be theREQUIRED response by the OATC?

A. The alarm is NOT required to be announced to the crew.The alarm response is REQUIRED to be reviewed.

B. The alarm is NOT required to be announced to the crew.The alarm response is NOT REQUIRED to be reviewed.

C. Announce the noun name of the alarm which shall be repeated back by theC/R Supervisor.The alarm response is REQUIRED to be reviewed.

D. Announce the noun name of the alarm which shall be repeated back by theC/R Supervisor.The alarm response is NOT REQUIRED to be reviewed.

Page 66 of 100

McGuire Nuclear Station2(JII MNS SR0 Ni?C Exäminatibn


In accordance with Tech. Spec. 3.4.16 (RCS Specific Activity, the LCO limits forReactor Coolant System Dose Equivalent 1-13 1 (DEl) AND gross specific activity(GSA) are:

1) Less than or equal to

______

pCi/gm DEl2) Less than or equal to

______

pCi/gm GSA


A. 1. 0.12. 100/E-bar

B. 1. 100/E-bar2. 0.1

C. 1. 1.02. 100/E-bar

D. 1. 100/E-bar2. 1.0

Page 67 of 100


1 point)


• A reactor startup is in progress on the unit

In accordance with SOMP 01-02 (Reactivity Management):

One active licensed SRO and (I) shall be dedicated to the reactor startup, withno concurrent responsibilities.

The DEDICATED SRO is (2) to perform peer checks related to the reactorstartup.


A. 1. one active licensed RO2. allowed

B. 1. one active licensed RO2. NOT allowed

C. 1. two active licensed ROs2. allowed

D. 1. twoactive licensed ROs2. NOT allowed

Page 68 of 100

McGuire Nuclear Station- 2011 MI’TS SRO NRCExamination


In accordance with NSD 700 (Verification Techniques):

To verify the position of a normally closed valve by turning the valve in the opendirection, it requires (1)

To verify the position of a LOCKED OPEN valve, the operator shall (2)


A. 1. no Red or White tags ONLY2. remove locking device and physically verify valve position then replace the

locking device

B. 1. no Red or White tags AND supervisor permission2. remove locking device and physically verify valve position then replace the

locking device

C. 1. no Red or White tags ONLY2. ensure the locking device is properly installed and locked ONLY

D. 1. no Red or White tags AND supervisor permission2. ensure the locking device is properly installed and locked ONLY

Page 69 of 100

McGuire Nuclear Station• 2011 MNS SRU NRC Examination



• Both units are operating at 100% RTP• The following alarms are received on both units:

o lAD-Il /C5(XFMRAURGENTALARM)o 2AD-11 /C5(XFMRAURGENTALARM)

• Operations Test Group is performing ‘B’ Train RN Valve Stroke Timing• Each alarm is the result of a loss of ONE Cooling Group to the respective

transformer

To prevent a turbine runback to <56%, cooling must be restore to the (1) withinaMAXIMUM0f (2)


A. 1. 1A Main Transformer2. 8min45sec

B. 1. lAMainTransformer2. 28min45sec

C. 1. 2A Main Transformer2. 8min45sec

D. 1. 2A Main Transformer2. 28min45sec

Page 70 of 100

McGu ire Nuclear Station• 2UlIMNSSRONRCExamination


Regarding the use of Electronic Dosimeters (ED):

• If a DOSE alarm setpoint is exceeded, the alarm will (1)• If a DOSE RATE alarm setpoint is exceeded, the alarm will (2)


A. 1. not clear until the ED is reset2. clear when the dose rate drops below the alarm setpoint

B. 1. not clear until the ED is reset2. not clear until the ED is reset

C. 1. automatically clear after 10 seconds2. clear when the dose rate drops below the alarm setpoint

D. 1. automatically clear after 10 seconds2. not clear until the ED is reset

Page 71 of 100

McGu ire Nuclear Station• 2OliMNSSRONRCExáininatioñ


Given the following conditions:

• A team of Operators is preparing to enter the RCA to perform a valve lineup in ahigh radiation area

• The Radiation Work Permit (RWP) specifies that the individual workers must notexceed their annual “ALERT” exposure limits

Based on the conditions above, the individual worker’s exposure must remain LESSTHAN a MAXIMUM of (1) of the annual (2)


A. 1. 80%2. NRC exposure limit

B. 1. 80%2. Duke administrative exposure limit

C. 1. 90%2. NRC exposure limit

D. 1. 90%2. Duke administrative exposure limit

Page 72 of 100

McGu ire Nuclear Station• 2OlIMNSSRONRCExamination



• Unit 1 has experienced a Large Break LOCA• Containment pressure is currently 4.5 PSIG

Which ONE (1) of the following describes the correct parameters and logic whichwould require entry into the RED path procedure FR-C.1 (Response to InadequateCore Cooling) per the CSF status trees?

A. WR TH > 7200 F and UR RVLIS Level indicates “less than 60

B. Core Exits> 7000 F and UR RVLIS Level indicates “less than 60”

C. WRTH >700°F and LR RVLIS Level <39%

D. Core Exits> 700°F and LR RVLIS Level < 39%

Page 73 of 100



The following conditions on Unit 1:

• Due to a fire in the Auxiliary Building control of the plant has been transferred tothe SSF

• PZR level is 85% and increasing slowly• PZR pressure is 2100 PSIG and STABLE• The SSF PZR Heaters are ON

1) In accordance with AP-24 (Loss of Plant Control Due to Fire) what actions shouldbe taken with regards to controlling PZR level?

2) What indications available at the SSF could be used to verify natural circulation?

A. 1. Allow the PZR to fill2. NC Loop WIR Pressure and Steam Generator Pressure

B. 1. Allow the PZR to fill2. Core Exit Thermocouples and NC Loop WIR Pressure

C. 1. Open the Rx Head vents to lower PZR level2. NC Loop WIR Pressure and Steam Generator Pressure

D. 1. Open the Rx Head vents to lower PZR level2. Core Exit Thermocouples and NC Loop WIR Pressure

Page 74 of 100

McGuire Nuclear Station• 2011 MNSSRO NRCExainination



• The unit is stable at 47% RTP• Turbine Impulse Pressure is 350 PSIG

Which ONE (1) of the following is the MINUMUM cond Won that would require entry intoAP-03 (Load Rejection)?

A. CF Pump IA ONLY TRIPs

B. CF Pumps IAAND lB TRIP

C. Busline 1A ONLY DEENERGIZES

D. Buslines IA AND lB DEENERGIZE

Page 75 of 100

McGuire Nuclear Station• 21111 MNS SROI’TRC Examination



• The unit was initially as 25% RTP with a power increase in progress• The unit experienced a complete loss of all onsite and offsite power• ECA-O.0 (Loss of All AC Power) has been implemented• Operators have been dispatched to the SSF to place the Standby Makeup

Pump in service• The crew has determined that a cooldown and depressurization of the NC

system is required using the S/G PORVs• No power has been restored from onsite or offsite sources yet

1) Which ONE (1) of the following is the major operational concern if the SGs areallowed to depressurize to less than 190 PSIG?

2) What is the MAXIMUM time the crew has to start the Standby Makeup Pump toprevent a postulated NC Pump Seal LOCA?

A. 1. Nitrogen injection from the CLAs2. 5 minutes

B. 1. Inadvertent criticality2. 5 minutes

C. I. Nitrogen injection from the CLAs2. 10 minutes

D. 1. Inadvertent criticality2. 10 minutes

Page 76 of 100

McGu ire Nuclear Station• 2011 MNS SRO NRCExamination



• Unit was operating at 100% RTP• 45 minutes ago a large break LOCA occurred on Unit 1• Containment pressure indicates 9 PSIG• Containment hydrogen concentration is 1%• CETs indicate 1100°F• RVLIS lower range level indicates 40%• 1EMF- 51A indicates 165 R/hr• Subcooling margin indicates -35°F• All SIG NR levels are off scale low with no CA flow indicated

Which ONE (1) of the following is the correct classification and associated EALNumber for this event?

REFERENCE 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 77 of 100

McGuire Nuclear Station• 2011 MNS SRO NRC ExamInatIon



• The unit was operating at 100% RTP• The TD CA pump is out-of-service for maintenance• A DCS failure results in ALL CF Control valves failing closed• As a result, both CF pumps trip on high discharge pressure• Both MD CA pumps fail to automatically start and cannot be started manually• FR-H.1 (Response to Loss of Secondary Heat Sink) has been implemented• SG WR level is 40% in all SG5

Based on the conditions above, NC system Feed and Bleed must be initiated within(1) minutes of three SG WR levels decreasing to less than a MAXIMUM of(2)


A. 1.42. 24%

B. 1.82. 24%

C. 1.42. 36%

D. 1.82. 36%

Page 78 of 100

McGuire Nuclear Station• 2U11MNSgRONRCExainihatioi



• Both units are at 100% RTP• Testing of Sequencer B on Unit I is in progress• Annunciator lAD-i I I El (SEQ B IN TEST) is illuminated

Which ONE (1) of the following describes the current operability status of the I B DGAND SI loading response if an automatic SI actuation occurs before any correctiveactions are taken?

A. The I B DG is OPERABLE ANDSI loads on both trains will start automatically.

B. The I B DG is OPERABLE ANDTrain A SI loads will start automatically, but Train B SI loads must be startedmanually.

C. The I B DG is INOPERABLE ANDSI loads on both trains will start automatically.

D. The I B DG is INOPERABLE ANDTrain A SI loads will start automatically, but Train B SI loads must be startedmanually.

Page 79 of 100

McGuire Nuclear Station2011 MNS SRO NRC Examination



• Unit 1 is releasing the Ventilation Unit Condensate Drain Tank (VUCDT) to theRC Discharge using the Continuous Release Method

• Shortly after the release was initiated, I EMF-44 (Ventilation Unit CondensateDrain Tank) count rate indication fails to a reading of less than background

In addition to stopping the release, which ONE (1) of the following statements describesadditional actions required for this LWR (Liquid Waste Release) in accordance with0P111A165001001 A (Ventilation Unit Condensate Drain Tank Operation)?

A. Document on Continuous Release paperwork that OEMF-49 (Waste LiquidDisc) is both OPERABLE and monitoring release flow path then restart therelease.

B. Request updated Continuous Release paperwork, update the OP paperworkthen restart the release.

C. Request paperwork for a Batch Release, update the OP paperwork and restartthe release.

D. Notify IAE to restore IEMF-44 to OPERABLE prior to any VUCDT release.

Page 80 of 100


Question: 81point)


• The unit is in MODE 6 with fuel movement in progress• A leak has developed which is causing Spent Fuel Pool level to decrease• The Spent Fuel Pool Level Low computer (OAC) alarm is activated

Which ONE (1) of the following will be the FIRST action(s) directed by the CRS tomitigate the loss of level in the Spent Fuel Pool in accordance with (AP-40 Loss ofRefueling Cavity Level)?

A. Makeup to the Spent Fuel Pool from the FWST.

B. Makeup to the Spent Fuel Pool from the RF Header.

C. Install the Spent Fuel Pool Weir Gate AND inflate the seals.

D. Move the fuel transfer cart to the spent fuel (pit) side AND close 1 KF-122 (FuelTransfer Tube Block valve).

Page 81 of 100



Regarding the use of FR-Z.3 (Response To High Containment Radiation):

1) At what MINIMUM reading on IEMF 51A (Containment High Range) is theYELLOW path for Containment High Radiation valid?

2) What actions does the CRS direct to mitigate the consequences of the event?

A. 1. 35R/hr2. Start the Containment Auxiliary Charcoal Filter Unit.

B. 1. 15R/hr2. Start the Containment Auxiliary Charcoal Filter Unit.

C. 1. 35RIhr2. Ensure the VE system is in service and purge containment to the annulus

D. 1. l5RIhr2. Ensure the VE system is in service and purge containment to the annulus

Page 82 of 100

McGu ire Nuclear Station• 2011 MNS SRO NRC Examination



• Unit I was operating at 100% RTP• While performing maintenance activities in the 1A DIG 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 affectedAND what MINIMUM actions are required per SLC 16.9.3 (Halon Systems)?

A. Auto actuation ONLY is blockedEstablish an Hourly fire watch within one hour.

B. Manual Electric AND Auto actuation are blocked.Establish an Hourly fire watch within one hour.

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

D. Manual Electric AND Auto actuation are blocked.Establish a Continuous Fire Watch within one hour.

Page 83 of 100




• The unit is operating at 30% RTP• NC pump IC trips due to mis-operation during l&E testing• Subsequently, a lockout occurs on IA Busline due to a fault• The Reactor Trip breakers remain closed

1) Which ONE (1) of the following describes the plant response?

2) For the conditions described above what actions are required and the basis forthose procedure actions?

A. 1. 1TA and 1TC auto-swap.2. Restart 1C NC pump within 6 hours to comply with TS 3.4.4 (RCS Loops

Modes 1 &2).

B. 1. ITA and 1TC auto-swap.2. Place the unit in MODE 3 within 6 hours to comply with TS 3.4.4 (RCS

Loops Modes I & 2).

C. 1. AnATWSisinprogress.2. Manually trip the turbine to conserve SG inventory.

D. I. An ATWS is in progress.2. Manually trip the turbine to generate a redundant reactor trip signal.

Page 84 of 100




• Unit 2 is operating at 50% RTP• A NC system leak has developed inside Containment• Containment pressure has increased from 0.1 PSIG to 0.4 PSIG• Lower Containment temperature has increased to 118°F

Based on the conditions above, which ONE (1) of the following describes theLIMITING concern for exceeding a Containment design limit per the basis for TS 3.6.4(Containment Pressure) and TS 3.6.5 (Containment Temperature)?

A. Design peak Containment temperature could be exceeded during a LOCA.

B. Design peak Containment pressure could be exceeded during a LOCA.

C. Design peak Containment temperature could be exceeded during a SteamLine Break.

D. Design peak Containment pressure could be exceeded during a Steam LineBreak.

Page 85 of 100

McGuire Nuclear Station• 2OllMNSSRONRCExamination


Given the following events and conditions on Unit 1:

• Unit was operating at 100% RTP and experienced a LOOP at 0200• The resulting transient caused a lB feedwater line break inside Containment• lB DIG failed to start and the TD CA pump is unavailable• IA CA pump tripped on overcurrent• FR H.1 (Response to Loss of Secondary Heat Sink) has been entered and feed and

bleed of the NC system was initiated at 0230• The lB CA pump has been returned to service and available as a source of feedwater

Given the following conditions at 2:45 AM:

• Containment pressure = 3.5 PSIG• Core exit T/Cs are stable at an average value = 560 °F

Indication A Loop B Looi C Loop D Loop

SIG WR level (%) 19 0 16 8

Into which Steam Generator(s) should CA flow be restored and what limitation, if any isrequired on CA flow rates?

A. SIGC ONLYNo limitation on CA flow rate

B. SIGsAORS/GCNo limitation on CA flow rate

C. SIGC ONLY100 gpm limitation on CA flow rate

D. S/GsCANDS/GD100 gpm limitation on CA flow rate

Page 87 of 100

McGuire Nuclear Station• 2O1XMNSSRONRCExaminaüôn



• The unit is operating at 100% RTP

The following indications are observed by the crew:

• Annunciator lAD-Il IGI (120 VAC ESS PWR Channel ATRBL) is LIT• All Channel I instruments have failed low• Power Range N-41 has failed low

1) Based on the conditions above, the CRS will direct the crew to implement

_______

per Annunciator Response lAD-li I Gi.

2) The indication used to verify that those actions have been successful is the


A. 1. AP-1 5 (Loss of Vital or Aux Control Power)2. top row of channel status lights - NORMAL

B. 1. AP-1 5 (Loss of Vital or Aux Control Power)2. switch indication on any pump powered from IETA - LIT

C. 1. OPIOIA/63501001A (I25VDCII2OVAC Vital I&C Power System)2. top row of channel status lights - NORMAL

D. 1. OP/0/A163501001A (I25VDCI12OVAC Vital l&C Power System)2. switch indication on any pump powered from IETA - LIT

Page 88 of 100


(1 point)


• A complete Loss of Off-Site power has occurred• Unit 2 DIGs started and sequenced normally• Unit 1 DIGs did NOT automatically start and attempts to start them have been

unsuccessful• Subsequently, a SGTR occurs on the 1A SIG

Which ONE (1) of the following describes the procedure that will be used to isolate theruptured S/C in this situation AND the procedural guidance regarding WHEN theruptured S/G will be isolated?

A. E-3 (Steam Generator Tube Rupture) is used to isolate the ruptured S/GAS SOON AS it is identified.

B. E-3 (Steam Generator Tube Rupture) is used to isolate the ruptured S/GONLY AFTER ruptured SIG NR level is greater than 11%.

C. ECA-O.O (Loss of All AC Power) is used to isolate the ruptured S/GAS SOON AS it is identified.

D. ECA-O.O (Loss of All AC Power) is used to isolate the ruptured S/CONLY AFTER ruptured S/C NR level isgreaterthan 11%.

Page 89 of 100

McGuire Nuclear Station• 2011 iWNS SEQ NRC ExaminationQuestion: 90

1 point)


• Unit I is at 100% RTP• Control Bank D Rods are at 220 steps• A single rod in Control Bank D drops to 115 steps

In accordance with AP-14 (Rod Control Malfunction) thermal power must be reducedto less than a MAXIMUM of (1) power within 2 hours.

The basis for restricting thermal power to less than this limit is to (2)


A. 1. 50%2. limit core peaking factors

B. 1. 50%2. minimize AFD swings

C. 1. 75%2. limit core peaking factors

D. 1. 75%2. minimize AFD swings

Page 90 of 100

McGuire Nuclear Station• 2011 MNS SRO NRCExaminationQuestion: 91

‘1 point)


• The unit is in MODE 6 with core reload in progress• NC system boron concentration is 2705 PPM• The following surveillances are being performed:

o PTI1IAI4600IIOO (Surveillance Requirements For Shutdown Conditions)o PT111A146001003 C (Weekly Surveillance Items Checklist)

The surveillance for NC system boron concentration performed during PTIIIAI4600I100(SR 3.9.1.1) ensures that keff during MODE 6 remains less than a MAXIMUM of

(1)

The MINIMUM Boric Acid Tank level required to meet the surveillance requirements ofPT/i /A/4600/003 C (TR 16.9.14.3) is (2)


REFERENCE PROVIDED

A. 1. 0.952. 8.7%

B. 1. 0.952. 13.6%

C. 1. 0.982. 8.7%

D. 1. 0.982. 13.6%

Page 91 of 100

McGuire Nuclear Station• 2OlJMNSSRONRCExaminationQuestion: 92

1 point)

Unit I is operating at 100% RTP. Gien the following:

• 1EMF-33 (Condenser Air Ejector Exhaust) is in Trip 2 alarm• 1EMF-71 (S/G A Leakage) is in Trip 2 alarm• Pressurizer level has been stabilized using AP-1 0 (NC Leakage Within the

Capacity of Both NV Pumps)• Letdown flow is 45 GPM• Charging flow is 78 GPM

The MAXIMUM time that AP-lO allows for the unit to reach MODE 3 for the conditionsspecified is (1)

In accordance with SLC 16.9.7 (Stby SID System) Condition C (Leakage), theStandby Makeup Pump (2) have to be declared INOPERABLE.


A. 1. 3 hours2. will

B. 1. 3 hours2. will not

C. 1. 6 hours2. will

D. 1. 6 hours2. will not

Page 92 of 100


1 point)


• Both units are operating at 100% RTP• A fire occurs in the Auxiliary Building on the 733’ Elevation in the Battery Room• AP-45 (Plant Fire) has been implemented• The crew has just reached the steps in Enclosure 13 (AB 733’ Battery Room

Fire Unit 1 and 2 Actions) which directs closing the PZR PORV Block valves onboth units

In accordance with the AP-45 Background document, these valves must be closedwithin (1) minutes of thestart of the fire.

In accordance with Tech Spec 3.4.11 (Pressurizer PORVs) Basis ONLY, after the PZRPORV Block valves are closed, the PZR PORVs are (2)


A. 1. 102. INOPERABLE

B. 1.302. INOPERABLE

C. 1. 102. OPERABLE

D. 1.302. OPERABLE

Page 93 of 100


(1 point)

With Unit I at 100% RTP the following conditions exist:

• The 1C SIG develops a 10 GPM tube leak• AP-1 0 (NC System leakage Within the Capacity of Both NV Pumps) has been

implemented• Plant load is reduced using AP-04 (Rapid Downpower), and manually tripped at

15% power• The crew has just transitioned to ES-0.1 (Reactor Trip Response)

Subsequently:

• NCS Pressure is 2210 PSIG, and slowly DECREASING• The crew arrives at Step 23.a of ES-0.1 and is directed to “Maintain Pzr

pressure - AT 2235 PSIG”• Simultaneously, the crew arrives at Step 18.a of AP-1 0 and is directed to

“Depressurizeto between 1900-1 955 PSIG using normal Pzr spray.”

Which ONE (1) of the following describes how the implementation of AP-lO andES-0.1 is coordinated?

A. Suspend actions in AP-lO until ES-0.1 is complete; THENReturn to AP-lO and complete all required actions.

B. Suspend actions in ES-0.1 until AP-lO is complete; THENReturn to ES-0.1 and complete all required actions.

C. Continue simultaneous implementation of ES-0.1 and AP-1 0;If conflicting guidance is provided, AP-lO actions will have priority.

D. Continue simultaneous implementation of ES-0.1 and AP-lO;If conflicting guidance is provided, ES-0.1 actions will have priority.

Page 94 of 100

McGu ire Nuclear Station• 2011 MNSSRO NRC ExüminattonQuestion: 95

‘1 point)


• The unit is in MODE 6 with core alterations in progress• It is determined that a Fuel Handling interlock must be bypassed to insert the

next fuel assembly into the core

• The interlock which must be bypassed is NOT specified in a procedure

In accordance with NSD-414 (Fuel Handling), which the individuals listed below isallowed to approve bypassing the interlock?

1. Fuel Handling SRO2. Reactor Engineering3. Operations Shift Manager

A. IONLY

B. 1 AND 2 ONLY

C. lAND 3ONLY

D. 1,2,AND3

Page 95 of 100

McGuire Nuclear Station2011 MNS SRO NRC ExaminationQuestion: 96

(1 point)

In accordance with SOMP 02-02 (Operations Roles In The Risk Management Process):

• In MODE 3, risk assessment during non-core business hours for emergentmaintenance activities is performed ONLY by the (1)

• Defense in Depth (DID) assessments (2) performed during MODE 3.


A. 1.WCCSRO2. are

B. 1. CRS2. are

C. 1.WCCSRO2. are NOT

D. 1. CRS2. are NOT

Page 96 of 100

McGuire Nuclear Station• 2011 MNSSRONRCExamiñationQuestion: 97

(1 point)


• Several upcoming maintenance activities to be worked concurrently have beenevaluated and it is determined that a Risk Management Plan is required

• The Risk Management Plan does NOT meet the exemption requirements ofNSD 213 (Risk Management Process)

In accordance with SOMP 02-02 (Operations Role In The Risk Management Process):

• The Electronic Risk Assessment Tool color associated with the LOWEST riskcondition where a Risk Management Plan must be generated is (1)

• The Risk Management Plan shall be approved by the (2)


A. 1. RED2. Site Vice-President

B. 1. ORANGE2. Site Vice-President

C. 1. RED2. Plant Operations Review Committee

D. 1. ORANGE2. Plant Operations Review Committee

Page 97 of 100


1 point)


• A Large-Break LOCA has occurred on Unit 2• The 2A ND pump tripped after starting• The ECCS is in cold leg recirculation using the 2B ND pump• A SITE AREA EMERGENCY (SAE) is in effect• The OSM has determined that an effort must be made to make the 2A ND pump

functional• A team will be dispatched from the OSC to repair the 2A ND pump• The maximum dose rate in the area of the 2A ND pump is 20 REM/hour

Considering ONLY the dose rate received at ND Pump ‘A’, the MAXIMUM stay time is(1) minutes.

In accordance with RP-003 (Site Area Emergency), Enclosure 4.4 (Request forEmergency Exposure) the emergency exposure is required to be APPROVED by the

(2)


A. 1.302. Radiation Protection Manager

B. 1.302. Emergency Coordinator

C. 1.752. Radiation Protection Manager

D. 1.752. Emergency Coordinator

Page 98 of 100

McGuire Nuclear Station• 2011 MNS MW NRCExamination



• The unit is operating at 100% RTP• lB NC pump #1 seal flow increases to 8 GPM• AP-08 (Malfunction of NC Pump) has been implemented• The crew is preparing to trip the NC pump

How will the remaining actions of AP-08 be addressed when implementing theEmergency Procedures after the reactor trip?

A. Suspend AP-08 action until directed by ES-0.1 (Reactor Trip Response) toevaluate implementing applicable abnormal procedures.

B. Direct an operator to perform AP-08, Enclosure 2 (NC Pump Post Trip ActionsFor #1 Seal Failure).

C. OMP 4-3 (Use Of Abnormal And Emergency Procedures) does not allow anyAP usage during implementation of Emergency Procedures (EP) unlessdirected by the EP.

D. Direct the opposite unit to complete the remainder of AP-08 since it was theprocedure in effect at the time of the trip.

Page 99 of 100

McGu ire Nuclear Station• 2U11 MNS SRO NRC ExaMinationQuestion: 100

(1 point)


• Following a LOCA with subsequent equipment failures, ECA-1 .1, Loss ofEmergency Coolant Recirculation has been implemented

• Subsequently, FR-Z.1 (Response to High Containment Pressure) isimplemented due to high Containment pressure

1) In accordance with the Emergency Response Guidelines (ERG5), which ONE (1) ofthe following describes the basis for why ECA-1.1 has priority over FR-Z.1regarding the operation of the NS pumps?

2) Under what condition does ECA-1 I require both NS pumps to be in operation?

A. 1. Per ERG rules of usage, once an ECA procedure is implemented, noactions in any other procedures should be performed until the ECA iscomplete.

2. Containment pressure exceeds a MINIMUM of 10 PSIG.

B. 1. Since there is no recirculation flow to the NC system, the importance ofconserving FWST inventory takes precedence over maximum availableheat removal from Containment.


C. 1. Per ERG rules of usage, once an ECA procedure is implemented, noactions in any other procedures should be performed until the ECA iscomplete.


D. 1. Since there is no recirculation flow to the NC system, the importance ofconserving FWST inventory takes precedence over maximum availableheat removal from Containment.


Page 100 of 100

Reference Listfor: 2011 MNS SRO NRC Examination

Steam Tables including Mollier Diagram

ECAO.1 Enc4

Tech Spec 3.8.1 (AC Sources - Operating)

FR-C.2 Page 10

Tech Spec 3.4.15 (RCS Leakage Detection Instrumentation)

RPIOIAI5700I000

COLR - 2.16 (Borated Water Sources - Shutdown)

Printed 6/9/2011 1:05:26 PM

Pagel of1

Reference Listfor: 2011 MNS SRO NRC Examination

Steam Tables including Mollier Diagram

ECAO.1 Enc4

Tech Spec 3.8.1 (AC Sources - Operating)

FR-C.2 Page 10

Tech Spec 3.4.15 (RCS Leakage Detection Instrumentation)

RPIOIAI5700I000

COLR - 2.16 (Borated Water Sources - Shutdown)

Printed 6/10/2011 12:13:45PM

F t.OSS OF ALL AG POWER RECOVERY WITHOUT I1PAGE NO:•MI4S REQUIRED I 31 of 34

iEPIIIAI5000IECA-O.1 Enclosure 4 - Page 1 of 1 Rev. 10UNIT 1 Natural Circulation Parameters -

1. The following conditions support or indicate natural circulation flow:

• NC subcooling - GREATER THAN 0°Fa SIG pressures - STABLE OR GOING DOWN• NC T-I-iots - STABLE OR GOING DOWN• Core exit TICs - STABLE OR GOING DOWN• NC T-Colds - AT SATURATION TEMPERATURE FOR SIG PRESSURE

(WITHIN THE LIMITS OF THE GRAPH BELOW).

2. Natural Circulation flow is not established. THEN raise dumping steam toestablish Natural Circulation fl-ow.

1200

1100

SG 1000

pRESSURF

pSI 500G

400

300

200

1 QO

900

800

700

600

0

200 250 300 350 400 450 500 550 600NC T-COLD(°F)

AG Sources —Operating3.8.1

3.8 ELECTRICAL POWER SYSTEMS

3.8.1 AC Sources-.--Operating

LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network andthe Onsite Essential Auxiliary Power System; and

b. Two diesel generators (DGs) capable of supplying the OnsiteEssential Auxiliary Power Systems;

AND

The automatic load sequencers for Train A and Train B shall beOPERABLE.

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

ACTIONSNOTE

LCO 3.O.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME

A. One offsite circuit A.1 Perform SR 3.8.1.1 for 1 hourinoperable. OPERABLE offsite circuit.

AND

Once per 8 hoursthereafter

AND

A.2 Declare required feature(s) 24 hours fromwith no offsite power discovery of noavailable inoperable when offsite power to oneits redundant required train concurrent withfeature(s) is inoperable. inoperability of

redundant requiredfeature(s)

AND

(continued)

McGuire Units 1 and 2 3.8.1-1 Amendment Nos. 221/203

• AC Sources - Operating3.8.1


A. (continued) A.3 Restore offsite circuit to 72 hoursOPERABLE status.

AND

6 days fromdiscovery of failure tomeet LCO

B. One DG inoperable. B.1 Perform SR 3.8.1.1 for the 1 houroffsite circuit(s).

AND

Once per 8 hoursthereafter

AND

B.2 Declare required feature(s) 4 hours fromsupported by the discovery ofinoperable DG inoperable Condition Bwhen its required concurrent withredundant feature(s) is inoperability ofinoperable, redundant required

feature(s)

AND

B.3.1 Determine OPERABLE DG 24 hoursis not inoperable due tocommon cause failure.

OR

B.3.2 Perform SR 3.8.1.2 for 24 hoursOPERABLE DG.

AND

(continued)

ACTIONS

McGuire Units I and 2 3.8.1-2 Amendment Nos. 184/1 66

ACTIONS

AC Sources—Operating3.8.1

(continued)

* For Unit I only, the Completion Time that the IA EDG can be inoperable as specified byRequired Action B.4 may be extended beyond the “72 hours and 6 days from discovery offailure to meet the LCO” up to a total of 10 days as part of the IA EDG Jacket/Intercooler WaterPump Motor repair. Upon completion of the repair and restoration, this footnote is no longerapplicable and will expire at 1741 hours on June 15, 2007.


B. (continued) B.4 Restore DG to OPERABLE 72 hours *

status.AND

6daysfromdiscovery of failure tomeet LCO *

C. Two offsite circuits C.1 Declare required feature(s) 12 hours frominoperable, inoperable when its discovery of

redundant required Condition Cfeature(s) is inoperable, concurrent with

inoperability ofredundant requiredfeature(s)

AND

C.2 Restore one offsite circuit 24 hoursto OPERABLE status.

McGuire Units I and 2 3.8.1-3 Amendment Nos. 241/-,

- ASourcss - Operat3.8.1

ACTIONS (continued)


D. Oneoffsitecircuit -NOTE-inoperable. Enter applicable Conditions and

Required Actions of LCO 3.8.9,AND “Distribution Systems—

Operating, when Condition D isOne DG inoperable, entered with no AC power source

to any train.

D.1 Restore offsite circuit to 12 hoursOPERABLE status.

OR

D.2 Restore DG to OPERABLE 12 hoursstatus.

E. Two DGs inoperable. E.1 Restore one DG to 2 hoursOPERABLE status.

F. One automatic load F.1 Restore automatic load 12 hourssequencer inoperable, sequencer to OPERABLE

status.

G. Required Action and G.1 Be in MODE 3. 6 hoursassociated CompletionTime of Condition A, B, ANDC, D, E, or F not met.

G.2 Be in MODE 5. 36 hours

H. Three or more AC H.1 Enter LCO 3.0.3. Immediatelysources inoperable.

McGuire Units I and 2 3.8.1-4 Amendment Nos. 184/166

- AG Sources — Operating3.8.1

SURVEILLANCE_REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.8.1.1 Verify correct breaker alignment and indicated power 7 daysavailability for each offsite circuit.

SR 3.8.1.2 NOTES1. Performance of SR 3.8.1.7 satisfies this SR.

2. All DG starts may be preceded by an engineprelube period and followed by a warmup periodprior to loading.

3. A modified DG start involving idling and gradualacceleration to synchronous speed may be usedfor this SR as recommended by the manufacturer.When modified start procedures are not used, thetime, voltage, and frequency tolerances ofSR 3.8.1.7 must be met.

Verify each DG starts from standby conditions and 31 daysachieves steady state voltage 3740 V and 4580 V,and frequency 58.8 Hz and 61.2 Hz.

(continued)


- AC Sources - Operating3.8.1

SURVEILLANCE REQUIREMENTS (continued)

NOTES1. DG loadings may include gradua[ loading as

recommended by the manufacturer.

2. Momentary transients outside the load range donot invalidate this test.

3. This Surveillance shall be conducted on only oneDG at a time.

4. This SR shall be preceded by and immediatelyfollow without shutdown a successful performanceof SR 3.8.1.2 or SR 3.8.1.7.

Verify each DG is synchronized and loaded and operatesfor 60 minutes at a load 3600 kW and 4000 kW.

SR 3.8.1.3


31 days

SR 3.8.1.4 Verify each day tank contains 39 inches of fuel oil. 31 days

SR 3.8.1.5 Check for and remove accumulated water from each day 31 daystank.

SR 3.8.1.6 Verify the fuel oil transfer system operates to 31 daysautomatically transfer fuel oil from storage tank to the daytank.

(continued)

McGuire Units 1 and 2 3.8.1-6 Amendment Nos. 254 I 234

AC Sources — Operating3.8.1


____________


SR 3.8.1.7 NOTESAll DG starts may be preceded by an engine prelubeperiod.

Verify each DG starts from standby condition and 184 daysachieves in 11 seconds voltage of 3740 V andfrequency of 57 Hz and maintains steady state voltage

3740 V and 4580 V, and frequency 58.8 Hz and61.2 Hz.

SR3.8.1.8 NOTESThis Surveillance shall not be performed in MODE I or 2.

Verify automatic and manual transfer of AC power 18 monthssources from the normal offsite circuit to each alternateoffsite circuit.

(continued)


AC Seurce = Operating3.8.1


_______________


SR 3.8.1.9 Verify each DG, when connected to its bus in parallel with 18 monthsoffsite power and operating with maximum kVAR loadingthat offsite power conditions permit, rejects a load greaterthan or equal to its associated single largest post-accident load, and:

a. Following load rejection, the frequency is 63 Hz;

b. Within 3 seconds following load rejection, thevoltage is 3740 V and 4580 V; and

c. Within 3 seconds following load rejection, thefrequency is 58.8 Hz and 61.2 Hz.

SR 3.8.1.10 Verify each DG does not trip and voltage is maintained 18 months4784 V during and following a load rejection of

3600 kWand4000 kW.

(continued)

McGuire Units I and 2 3.8.1-8 Amendment Nos. 192 (Unit 1)173 (Unit 2)

AC-Sources — Operating3.8.1


______________


SR 3.8.1.11 NOTES1. All DG starts may be preceded by an engine

prelube period.

2. This Surveillance shall not be performed inMODE 1,2,3, or4.

Verify on an actual or simulated loss of offsite power 18 monthssignal:

a. De-energization of emergency buses;

b. Load shedding from emergency buses;

c. DG auto-starts from standby condition and:

1. energizes the emergency bus in11 seconds,

2. energizes auto-connected blackout loadsthrough automatic load sequencer,

3. maintains steady state voltage3740 V and 4580 V,

4. maintains steady state frequency58.8 Hz and 61.2 Hz, and

5. supplies auto-connected blackout loads for5 minutes.

(continued)


•ACSouroes — Operating3.8.1


______________


SR3.8.1.12 . .

All DG starts may be preceded by prelube period.

Verify on an actual or simulated Engineered Safety 18 monthsFeature (ESF) actuation signal each DG auto-starts fromstandby condition and:

a. In 11 seconds after auto-start signal achievesvoltage of 3740 and during tests, achievessteady state voltage 3740 V and 4580 V;

b. In 11 seconds after auto-start signal achievesfrequency of 57 Hz and during tests, achievessteady state frequency 58.8 Hz and 61.2 Hz;

c. Operates for 5 minutes; and

d. The emergency bus remains energized from theoffsite power system.

(continued)


AC Sources —Operating-3.8.1


(continued)


SR 3.8.1.13 Verify each DGs non-emergency automatic trips are 18 monthsbypassed on actual or simulated loss of voltage signal onthe emergency bus concurrent with an actual orsimulated ESF actuation signal.

SR 3.8.1.14 NOTES1. Momentary transients outside the load range do

not invalidate this test.

2. DG loadings may include gradual loading asrecommended by the manufacturer.

Verify each DG, when connected to its bus in parallel with 18 monthsoffsite power and operating with maximum kVAR loadingthat offsite power conditions permit, operates for

24 hours:

a. For 2 hours loaded 4200 kW and 4400 kW;and

b. For the remaining hours of the test loaded3600 kW and 4000 kW.

McGuire Units 1 and 2 3.8.1-11 Amendment Nos.242/223

- AC Sources Opersting3.8.1


_______________


SR 3.8.1.15 NOTES1. This Surveillance shall be performed within

5 minutes of shutting down the DG after the DGhas operated 2 hours loaded 3600 kW and

4000 kW.

Momentary transients outside of load range donot invalidate this test.

2. All DG starts may be preceded by an engineprelube period.

Verify each DG starts and achieves, in 11 seconds, 18 monthsvoltage 3740 V, and frequency 57 Hz and maintainssteady state voltage 3740 V and 4580 V andfrequency 58.8 Hz and 61.2 Hz.

SR 3.8.1.16 NOTESThis Surveillance shall not be performed in MODE 1, 2, 3,or4.

Verify each DG: 18 months

a. Synchronizes with offsite power source whileloaded with emergency loads upon a simulatedrestoration of offsite power;

b. Transfers loads to offsite power source; and

c. Returns to standby operation.

(continued)



AC Sourcs=Operating3.8.1

SR 3.8.1.17

SURVEILLANCE

NOTESThis Surveillance shall not be performed in MODE 1, 2, 3,or4.

Verify, with a DG operating in test mode and connectedto its bus, an actual or simulated ESF actuation signaloverrides the test mode by:

FREQUENCY

18 months

a. Returning DG to standby operation; and

b. Automatically energizing the emergency load fromoffsite power.

SR 3.8.1 .18 Verify interval between each sequenced load block is 18 monthswithin design interval for each automatic load sequencer.

(continued)

McGuire Units 1 and 2 3.8.1-13 Amendment Nos. 184/1 66

AC.Srces — Operatin3.8.1


______________


SR 3.8.1.19 -NOTES1. All DG starts may be preceded by an engine

prelube period.

2. This Surveillance shall not be performed inMODE 1,2,3, or4.

Verify on an actual or simulated loss of offsite power 18 monthssignal in conjunction with an actual or simulated ESFactuation signal:

a. De-energization of emergency buses;

b. Load shedding from emergency buses; and

c. DG auto-starts from standby condition and:

1. energizes the emergency bus in 11seconds,

2. energizes auto-connected emergencyloads through load sequencer,

3. achieves steady state voltage 3740 Vand 4580 V,

4. achieves steady state frequency 58.8 Hzand 61.2 Hz, and

5. supplies auto-connected emergency loadsfor 5 minutes.

(continued)


- AC Sources - Operating3.8.1


______________


SR 3.8.1.20 NOTESAll DG starts may be preceded by an engine prelubeperiod.

Verify when started simultaneously from standby 10 yearscondition, each DG achieves, in 11 seconds, voltage of

3740 V and frequency of 57 Hz and maintains steadystate voltage 3740 V and 4580 V, and frequency

58.8 Hz and 61.2 Hz.


MNS RESPONSE TO DEGRADED CORE COOLING PAGE NO.EPI1IAI5000IFR-C.2 10 of 44

UNIT 1 Rev. 7

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

7. Check Reactor Vessel level as follows:

a. Determine required reactor vessel deltaP from table below:

Required “REACTOR VESSEL D/P”

Number of

NC Pumps TRN A TRN B

On With 1A NC Pump With 1C NC Pump

On Off On Off

4 44% N/A 44% N/A

3 30% 24% 30% 24%

2 23% 15% 23% 15%

1 16% 10% 16% 10%

— b. Check “REACTOR VESSEL DIP” - b. Perform the following:GREATER THAN REQUIRED DELTAP. — 1) if “REACTOR VESSEL DIP” going

up, THEN observe Note prior toStep 1 and RETURN ] Step 1.

_2) TQStep8.

— c. RETURN] procedure and step ineffect.

8. Check if one NC pump should bestopped:

— a. All NC pumps - ON. — a. ]Q.Step 10.

— b. Stop lB NC pump.

— c. Place 1 NC-29C (B NC Loop PZR SprayControl) in manual and CLOSED.

_d. jQ.Step10.

RCS Leakage Detection instrumentation

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.15 RCS Leakage Detection Instrumentation

LCO 3.4.15 The following RCS leakage detection instrumentation shall beOPERABLE:

a. The containment floor and equipment sump level monitors and theincore instrument sump level alarm;

b. The containment atmosphere particulate radioactivity monitor; and

c. The containment ventilation unit condensate drain tank level monitor.

APPLICABILITY: MODE 1 for all instrumentation,MODES 2, 3, and 4 for all instrumentation except the containmentatmosphere particulate radioactivity monitor.

ACTIONS

NOTESeparate Condition entry is allowed for each leakage detection instrument.


A. One or both containment A.1floor and equipment - NOTE -

sump level monitor(s) Not required until 12 hoursinoperable, after establishment of

steady state operation.

Perform SR 3.4.13.1. Once per 24 hours

AND

A.2 Restore inoperable 30 dayscontainment floor andequipment sump levelmonitor(s) to OPERABLEstatus.

(continued)

McGuire Units I and 2 3.4.15-1 Amendment Nos. 235 /217

ACTIONS (continued)

RCS Leakage Detection instrumentation3t5


B. Containment B.1atmosphere particulate - NOTE -

radioactivity monitor Not required until 12 hoursinoperable, after establishment of

steady state operation


OR

B.2 Analyze grab samples of Once per 24 hoursthe containmentatmosphere.

C. Containment ventilation C.1.1unit condensate drain - NOTE -

tank level monitor Not required until 12 hoursinoperable, after establishment of

steady state operation


OR

C. 1.2 Analyze grab samples of Once per 24 hours.the containmentatmosphere.

OR

C.1.3 Perform SR 3.4.15.1. Once per8 hours.

AND

C.2 During Modes 2, 3, and 4, 30 daysrestore inoperablecontainment ventilation unitcondensate drain tanklevel monitor toOPERABLE status.

(continued)

McGuire Units I and 2 3.4.15-2 Amendment Nos. 235 /217

RCS Leakage Detection instrumentation3.4:15

ACTIONS (continued)


D. Containment D.1 Restore containment 30 daysatmosphere particulate atmosphere particulateradioactivity monitor radioactivity monitor toinoperable in MODE 1. OPERABLE status.

AND OR

Containment ventilation D.2 Restore containmentunit condensate drain ventilation unit condensate 30 daystank level monitor drain tank level monitor toinoperable in MODE 1. OPERABLE status.

E. Incore instrument sump E.1level alarm inoperable. - NOTE -

Not required until 12 hoursafter establishment ofsteady state operation.


F. Required Action and F.1 Be in MODE 3. 6 hoursassociated CompletionTime not met. AND

F.2 Be in MODE 5. 36 hours

G. All required monitors G.1 Enter LCO 3.0.3. Immediatelyinoperable.

McGuire Units 1 and 2 3.4.15-3 Amendment Nos. 235 /217

RCS Leakage Detection instrumentation- 3.4t5

SURVEILLANCE_REQUIREMENTS


SR 3.4.15.1 Perform CHANNEL CHECK of the containment 12 hoursatmosphere particulate radioactivity monitor.

SR 3.4.15.2 Perform COT of the containment atmosphere particulate 92 daysradioactivity monitor.

SR 3.4.15.3 Perform CHANNEL CALIBRATION of the containment 18 monthsfloor and equipment sump level monitors.

SR 3.4.15.4 Perform CHANNEL CALIBRATION of the containment 18 monthsatmosphere particulate radioactivity monitor.

SR 3.4.15.5 Perform CHANNEL CALIBRATION of the containment 18 monthsventilation unit condensate drain tank level monitor.

SR 3.4.15.6 Perform CHANNEL CALIBRATION of the incore 18 monthsinstrument sump level alarm.

McGuire Units 1 and 2 3.4.15-4 Amendment Nos. 235 /217

Duke Energy Procedure No.

McGuire Nuclear Station RPIOIA/5700I000

Classification Of Emergency Revision No.

017

Electronic Reference No.

Reference Use MC0048M3

PERFORMANCE* * * * * * * * * * UNCONTROLLED FORPRINT * * * * * * * * * *

(ISSUED) - PDF Format

RP/OIA/5700J00UPage 2 of 3

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 degradationof the level of safety of the plant or indicate a security threat to facilityprotection has been initiated.

1.1.2 No releases of radioactive material requiring offsite response or monitoringare 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 potentialsubstantial degradation of the level of safety of the plant or a security eventthat involves probable life threatening risk to site personnel or damage to siteequipment because of HOSTILE ACTION.

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

1.3 Site Area Emergency

1.3.1 Events are in process or have occurred which involve actual or likely majorfailures of plant functions needed for protection of the public or HOSTILEACTION that results in intentional damage or malicious acts; (1) toward sitepersonnel or equipment that could lead to the likely failure of or; (2) thatprevent effective access to equipment needed for the protection of the public.

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

1.4 General Emergency

1.4.1 Events are in process or have occurred which involve actual or imminentsubstantial core degradation or melting with potential for loss of containmentintegrity or HOSTILE ACTION that results in an actual loss of physicalcontrol of the facility.

1.4.2 Releases can be reasonably expected to exceed EPA Protective ActionGuidelines 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 anyprotection system or operator action initiated in response of the event.

RP/OIAJ57001IQ0Page 3 of 3

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 IFa General Emergency is NOT declared in Step 2.2, Q the condition does not affectfission product barriers, THEN review the listing of enclosures to determine if the eventis applicable to one of the categories shown.

2.4 Compare actual plant conditions to the Emergency Action Levels evaluated in 2.2 and/or2.3, then declare the appropriate Emergency Class as indicated.

2.4.1 Event Declaration time

________________________

2.5 linpiement the applicable Emergency Response Procedure (RP) for that classificationand continue with subsequent steps of this procedure.

Notification of Unusual Event RP/0/A15700/00 1Alert RP/0/A/5700/002Site Area Emergency RP/0/A/5700/003General Emergency RP/0/A/5700/004.

3. Subsequent Actions

3.1 To escalate, de-escalate, or terminate the Emergency, compare plant conditions to theInitiating 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 ofemergencies as needed.

4.0 Enclosures4.1 Fission Product Barrier Matrix4.2 System Malfunctions4.3 Abnormal Rad Levels/Radiological Effluent4.4 Loss of Shutdown Functions4.5 Loss of Power4.6 Fire/Explosion and Security Events4.7 Natural Disasters, Hazards and Other Conditions Affecting Plant Safety4.8 Definitions/Acronyms4.9 Emergency Declaration Guidelines4.10 Radiation Monitor Readings for Enclosure 4.3 EALs4.11 Commitment Reference for Emergency Action Levels

En .re 4.1

Fission Product Barrier MatrixRP/O/A/57U. OPage 1 of5

Use EALs to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all 3 barriers. Classif,r according to the tableon 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 isimminent (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 fqrthe Fuel Clad Barrier loss or potential loss EALs were met previously (validated and sustained) during the event, even if the conditions do notcurrently exist.

Note 4: Critical Safety Function (CSF) indications are not meant to include transient alarm conditions which may appear during the start-up ofengineered safeguards equipment. A CSF condition is satisfied when the alarmed state is valid and sustained. The STA should be consulted toaffirm if any CSF has been validated prior to that CSF being used as the basis to classify an emergency. {1} 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 emergencyprior to any function restoration procedure being implemented within the confines of ECA-O.O.EAL # Unusual Event

4.1.U.1 Potential Loss ofContainment

4.1.U.2 Loss of Containment

EAL # Alert

4.1 A. 1 Loss OR Potential Lossof

Nuclear Coolant System

4.1.A.2 Loss OR Potential Lossof

Fuel Clad

4.l.A.3 Potential Loss ofContainment

ANDLoss OR Potential Loss

of Any Other Barrier

EAL # Site Area Emergency

4.l.S.l Loss OR Potential Lossof Both

Nuclear Coolant SystemAND

Fuel Clad4.l.S.2 Loss

ANDPotential Loss

Combinations of BothNuclear Coolant System

ANDFuel Clad

4.l.S.3 Loss of ContainmentAND

Loss OR Potential Lossof Any Other Barrier

EAL # General Emergency

4.l.G.1 Loss of All Three Barriers

4.1 .G.2 Loss of Any Two BarriersAND

Potential Loss of the Third

Elk re4.1

Fission Product Barrier Matrixi/O!A/57. 0Page 2 of 5

NOTE: If a barrier is affected, it has a single point value based on a “potential loss” or a “loss”. “Not Applicable” isincluded in the matrix as a place holder only, and has no point value assigned.

Barrier Points (1-5) Potential Loss (X) Loss (X)

Containment 1 3

NCS 4 5

Fuel Clad 4 5

Total Points

Total Points Classification

1 —3 Unusual Event

4—6 Alert

7—10 SiteAreaEmergency

1 1 - 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.l.A.l for Loss of Nuclear CoolantSystem) that best fits the loss of barrier description.

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

EL Are 4.1 1u!O/A157u..Fission Product Barrier Matrix Page3of5

POTENTIAL LOSS -

(1 Point)

• Containment-RED.

• Core Cooling -

RED Path isindicated for >15minutes.

2. Containment Conditions

ContainmentPressure> 15PSIG.

• H2 concentration>9%.

• Containmentpressure greater than3 psig with less thanone full train of NSand a VX-CARFoperating.

• Rapid unexplaineddecrease incontainmentpressure followinginitial increase.

• Containmentpressure or sumplevel response notconsistent withLOCA conditions.

POTENTIAL LOSS -

(4 Points)

• NCS Integrity-RED.

• Heat Sink-RED.

2. NCS Leak Rate

• Unisolable leakexceeding thecapacity of onecharging pump inthe normalcharging modewith letdownisolated.

• GREATER THANavailable makeupcapacity asindicated by a lossof NCS subcooling.

POTENTIAL LOSS -

(4 Points)

• Core Cooling-ORANGE.

• Heat Sink-RED.

• Coolant ActivityGREATER THAN300 i.tCi/cc DoseEquivalent Iodine(DEl) 1-13 1.

4.1.C CONTAINMENT BARRIER 4.1.N NCS BARRIER

LOSS -

(3 Points)

1. Critical Safety Function Status

4.1.F

LOSS -

(5 Points)

FUEL CLAD BARRIER

Not applicable.


LOSS -

(5 Points)

• Not applicable.


• Core Cooling-RED.

2. Primary Coolant Activity Level

• Not applicable.

CONTINUED CONTINUED CONTINUED

En .re 4.1

Fission Product Barrier Matrix

RP/OIA/57(. .jOPage 4 of 5

Containnientisolation isincomplete and arelease path fromcontainment exists.

Release ofsecondaiy side tothe environmentwith primary-to-secondary leakageGREATER THANTech Specallowable.

4.1.N

POTENTIAL LOSS -

(4 Points)

3. SG Tube Rupture

• Primary-to-Secondary leakrate exceeds thecapacity of onecharging pump inthe normalcharging modewith letdownisolated.

• Indication that aSO is Ruptured andhas a Non-Isolablesecondary linefault.

• Indication that aSG is ruptured anda prolonged releaseof contaminatedsecondary coolantis occurring fromthe affected SG tothe environment.

• Not applicable.

• Containmentradiation monitorEMF 51 A or 51 BReading at timesince shutdown

0-0.5 hrs > 99 RJhr0.5-2 hrs > 43 R/hr2-4hrs>31 R/hr4-8 hrs >22 RJhr>8 hrs> 13 RJhr

4.1.C CONTAINMENT BARRIER

POTENTIAL LOSS - LOSS -

(1 Point) (3 Points)

NCS BARRIER

3. Containment Isolation Valves Status AfterContainment Isolation Actuation

Not applicable.

4.1.F FUEL CLAD BARRIER

LOSS -

(5 Points)


(4 Points) (5 Points)

3. Containment Radiation Monitoring

Not applicable.

4. SG Secondary Side Release With Primary-to-Secondary Leakage- Not applicable.

4. Containment Radiation Monitoriiw

Not applicable.

4. Emcrency Coordinator/EOF DirectorJudement

• Any condition, including inability to monitorthe barrier, that in the opinion of theEmergency Coordinator/EOF Directorindicates LOSS or POTENTIAL LOSS ofthe fuel clad barrier.

END

CONTINUED CONTINUED

En re4.1 RP/O/A/57U OFission Product Barrier Matrix Page 5 of5

4.1.C CONTAINMENT BARRIER

POTENTIAL LOSS -

(1 Point)

5. Significant Radioactive Inventory InContainment

Containment Rad. • Not applicable.Monitor EMF5 1Aor 51BReading @ timesince shutdown:>39ORJhr@o - 0.5 hr> l70RJhr(0.5 - 2 hr> 125 RJhr @2 - 4 hr> 9ORJhr @4 - 8 hr> 53 RJhr @> 8 hr.

6. Emergency Coordinator /EOF Director

4.1.N NCS BARRIER

POTENTIAL LOSS -

(4 Points)

Emergency CoordinatorlEOF DirectorJudement

Any condition, including inability to monitorthe barrier, that in the opinion of theEmergency Coordinator/EOF Directorindicates LOSS or POTENTIAL LOSS ofthe NCS barrier.

• Any condition, including inability to monitorthe barrier, that in the opinion of theEmergency Coordinator/EOF Directorindicates LOSS or POTENTIAL LOSS ofthe containment barrier.

LOSS -

(3 Points)

LOSS -

(5 Points)

4.1.F FUEL CLAD BARRIER


(4 Points) (5 Points)

END

Judgement

END

Enclosure 4.2 RPIO/A!57001000

UNUSUAL EVENT

System Malfunctions

ALERT

Page 1 of2

SITE AREA EMERGENCY GENERAL EMERGENCY

4.2.U.1 Inability to Reach RequiredShutdown Within TechnicalSpecification Limits.

OPERATING MODE: 1,2,3,4

4.2U.2 Unplanned Loss of Most or All SafetySystem Annunciation or Indication inthe Control Room for Greater Than15 Minutes.

4.2.A.1 Unplanned Loss of Most or All 4.2.S.1Safety System Annunciation orIndication in Control RoomWith Either (1) a SignificantTransient in Progress, or (2)Compensatory Non-AlarmingIndicators Unavailable.

4.2.A.1-1 The following conditions exist:

Unplanned loss of most (>50%)annunciators associated with safetysystems for greater than 15 minutes.

Inability to Monitor aSignificant Transient inProgress.

exist:

Loss of most (>50%)annunciators associated withsafety systems.

AND

Unplanned loss of most (>50%)annunciators associated with safetysystems for greater than 15 minutes.

In the opinion of the Operations ShiftManager/Emergency Coordinator/EOFDirector, the loss of the annunciatorsor indicators requires additionalpersonnel (beyond normal shiftcompliment) to safely operate the unit.

In the opinion of the OperationsShift Manager/EmergencyCoordinator/EOF Director, theloss of the annunciators orindicators requires additionalpersonnel (beyond normal shiftcompliment) to safely operate theunit.

EITHER of the following:A significant plant transient isin progress.

A significant planttransient is in progress.

AND

Inability to provide manualmonitoring of any of thefollowing Critical SafetyFunctions:

• subcriticality• core cooling• heat sink• containment.

END

4.2.U.1-1 Plant is brought to requiredoperating mode within TechnicalSpecifications LCO Action Statement OPERATING MODE: 1, 2, 3, 4Time.


4.2.S.1-1 The following conditions

OPERATING MODE: 1,2, 3, 4

4.2.U.2-1 The following conditions exist:

AND

AND

AND

Loss of the OAC.

AND

CONTINUED

OR

• Loss of the OAC.

END

Enclosure 4.2 /OIAI57oo/oooSystem Malfunctions Page 2 of 2

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 1-13 1 greater than theTechnical Specification allowablelimit. (*mode 3 with Tavg 5 00°F)

4.2.U.4 Reactor Coolant System (NCS)Leakage.


4.2.U.4-l Unidentified leakage> 10 gpm.

4.2.U.4-2 Pressure boundary leakage> 10 gpm.

4.2.U.4-3 Identified leakage> 25 gpm.

4.2.U.5 Unplanned Loss of ALL Onsite orOffsite Communications.

OPERATING MODE: ALL

4.2.U.5-1 Loss of all onsite communicationscapability (internal phone system, PAsystem, onsite radio system) affectingthe ability to perform routineoperations.

4.2.U.5-2 Loss of all offsite communicationscapability (Selective Signaling, NRCETS lines, offsite radio system,commercial phone system) affectingthe ability to communicate with offsiteauthorities.

END

Enclosure 4.3 1u/O/A/5700/000Abnormal Rad Levels/Radiological Effluent Page 1 of5


4.3.U.1 Any Unplanned Release of Gaseous orLiquid Radioactivity to theEnvironment that Exceeds Two Timesthe SLC Limits for 60 Minutes orLonger.

OPERATING MODE: ALL

Note: (This applies to all EALs in the 4.3.U.1IC). If the monitor reading is sustainedfor the time period indicated in the EAL

Aiii the required assessments(procedure calculations) cannot becompleted within this time period,declaration must be made based on thevalid radiation monitor reading.

4.3.U.1-1 A valid indication on radiation monitorEMF- 49L, EMF.44L or EMF-3 1(when aligned to RC) of

5.45E+06 cpm for 60 minutes or willlikely continue for 60 minutes, whichindicates that the release may haveexceeded the initiating condition andindicates the need to assess the releasewith procedure HP/0/B/1 009/010,HP/0/B/1 009/029, or SH/0/B/2005/00 1.

Any Unplanned Release ofGaseous or LiquidRadioactivity to theEnvironment that Exceeds200 Times the SLC limitsfor 15 Minutes or Longer.

OPERATING MODE: ALL

Note: (This applies to all EALs in the4.3.A.1 IC). If the monitorreading is sustained for the timeperiod indicated in the EALAND the required assessments(procedure calculations) cannotbe completed within this timeperiod, declaration must be madebased on the valid radiationmonitor reading.

4.3.A.1-1 A valid indication onradiation monitor EMF- 49Hof1. 56E+03 cpmfor

15 minutes or will likelycontinue for 15 minutes, whichindicates that the release mayhave exceeded the initiatingcondition and indicates the needto assess the release withprocedure HP/0/B/1 009/010,HP/0/B/1009/029, orSHJO/B/2005/00 1.

4.3.S.l Boundary DoseResulting from anActual or ImminentRelease ofRadioactivity thatExceeds 100 mRemTEDE or 500 mRemCDE Adult Thyroidfor the Actual orProjected Durationof the Release.

OPERATING MODE: ALL

Note 1: These EMF readings arecalculated based onaverage annualmeteorology, siteboundary dose rate, anddesign unit vent flow rate.Calculations by the doseassessment team useactual meteorology,release duration, and unitvent flow rate. Therefore,these EMF readingsshould not be used if doseassessment teamcalculations are available.

(Continued)

4.3.G.1 Boundary DoseResulting from anActual or ImminentRelease ofRadioactivity thatExceeds 1000mRem TEDE or5000 mRem CDEAdult Thyroid forthe Actual orProjected Durationof the Release.

OPERATING MODE: ALL

Note 1: These EMF reading arecalculated based onaverage annualmeteorology, siteboundary dose rate, finddesign unit vent flowrate. Calculations by thedose assessment teaiji useactual meteorology,release duration, and unitvent flow rate.Therefore, these EMIFreadings should not beused if dose assessmentteam calculations areavailable.

4.3.A.l

(Continued)

(Continued)(Continued)

Enclosure 4.3

UNUSUAL EVENT

Abnormal Rad Levels/Radiological Effluent

ALERT SITE AREA EMERGENCY

Ip/O!A!57oo/oooPage 2 of 5

GENERAL EMERGENCY

4.3.U.1-2 A valid indication on radiation monitorEMF- 36L of 2.05E+04 cpm for

60 minutes or will likely continue for60 minutes, which indicates that the

release may have exceeded the initiatingcondition and indicates the need to assessthe release with procedureHP/0/B/1009/0 10, HP/0/B/l009/029, orSH/0/B/2005/00 1.

4.3.U.l-3 A valid indication on radiation monitorEMF-3 1 (when aligned to WC orWWCB) of 9.174 E+03 cpm for

60 minutes or will likely continue for60 minutes which indicates that the

release may have exceeded the initiatingcondition and indicates the need to assessthe release with procedureHP/0/B/1 009/010, HP/0/B/1 009/029, orSH/0/B/2005/00 1.

4.3.U.1-4 Gaseous effluent being released exceedstwo times SLC 16.1 1-6 for

60 minutes as determined by RadiationProtection (RP) procedure.

4.3.U.1-5 Liquid effluent being released exceedstwotimesSLC 16.11-1 for

60 minutes as determined by RadiationProtection (RP) procedure.

(Continued)

4.3.A.1-2 A valid indication onradiation monitor EMF- 36Lof 2.05E+06 cpm for> 15 minutes or will likelycontinue for 15 minutes,which indicates that the releasemay have exceeded the initiatingcondition and indicates the needto assess the release withprocedure HP/0/B/1 009/010,HP/0/B/1009/029, orSHJO/B/2005/00 1.

4.3.A.1-3 Gaseous effluent being releasedexceeds 200 times the level ofSLC l6.ll-6for>l5minutesasdetermined by RadiationProtection (RP) procedure.

4.3.A.1-4 Liquid effluent being releasedexceeds 200 times the level ofSLC 16.11-1 for> 15 minutes asdetermined by RadiationProtection (RP) procedure.

(Continued)

Note 2: If dose assessment teamcalculations cannot becompleted in 15 minutes,then valid monitor readingshould be used foremergency classification.

4.3.S.1-1 A valid indicationon radiation monitorEMF-36H of3.4E+03cpmsustained for> 15 minutes.

4.3.S.1-2 Dose assessment teamcalculations indicatedose consequencesgreater than 100mRem TEDE or 500mRem CDE AdultThyroid at the siteboundary.

4.3.S.1-3 Analysis of fieldsurvey results or fieldsurvey samplesindicates doseconsequences greaterthan 100 mRemTEDE or 500 mRemCDE Adult Thyroidat the site boundary.

END

Note 2: If dose assessment tamcalculations cannot lecompleted in 15 minutes,then valid monitorreading should be usedfor emergencyclassification.

4.3.G.1-1 A valid indicationon radiation monitorEMF-36H of3.4 E + 04 cpmsustained for>15 minutes.

4.3.G.1-2 Dose assessmentteam calculationsindicate doseconsequencesgreater than 1000mRem TEDE or5000 mRem CDEAdult Thyroid at thesite boundary.

4.3.G.1-3 Analysis of fieldsurvey results orfield survey samplesindicates doseconsequencesgreater than 1000mRem TEDE or5000 rnRemCDEAdult Thyroid at thesite boundary.

END

Enclosure 4.3

UNUSUAL EVENT

Abnormal Rad Levels/Radiological Effluent

ALERT

R1:/O/A/57oo/oooPage3of5

SITE AREA EMERGENCY GENERAL EMERGENCY

4.3.U.2 Unexpected Increase in PlantRadiation or Airborne Concentration.

OPERATING MODE: ALL

4.3 .U .2-1 Indication of uncontrolled water leveldecrease of greater than 6 inches in thereactor refueling cavity with allirradiated fuel assemblies remainingcovered by water.

4.3.A.2 Major Damage toIrradiated Fuel or Loss ofWater Level that Has orWill Result in theUncovering of IrradiatedFuel Outside the ReactorVessel.

Does not apply to spent fuel in drycask storage. Refer to EPLANsection D basis document.

4.3.U.2-2 Uncontrolled water level decrease ofgreater than 6 inches in the spent fuelpool and fuel transfer canal with allirradiated fuel assemblies remainingcovered by water.

4.3.U.2-3 Unplanned valid area EMF readingexceeds the levels shown in Enclosure4.10.

OPERATING MODE: ALL

4.3.A.2-1 An unplanned valid trip IIalarm on any of thefollowing radiationmonitors:

Spent Fuel BuildingRefueling Bridge1EMF-172EMF-4

END Spent Fuel Pool Ventilation1 EMF422EMF-42

Reactor Building RefuelingBridgelEMFl6*

2EMF-3 *

Containment Noble Gas1EMF39*2EMF39*

*Applies to Mode 6 and No Mode Only.(Continued)

Enclosure 4.3 RP/O/A/5700/000Abnormal Rad Levels/Radiological Effluent Page 4 of 5


4.3.A.2-2 Plant personnel report thatwater level drop in reactorrefueling cavity, spent fuelpool, or fuel transfer canalhas or will exceed makeupcapacity such that anyirradiated fuel will becomeuncovered.

4.3.A.2-3 NC system wide range level<358 inches after initiationof NC system make-up.

AND

Any irradiated fuel assembly notcapable of being lowered intospent fuel pool or reactor vessel.

4.3.A.2-4 Spent Fuel Pool or FuelTransfer Canal leveldecrease of >2 feet afterinitiation of makeup.

AND

Any irradiated fuel assemblynot capable of being fullylowered into the spent fuelpool racks or transfer canalfuel transfer system basket.

(Continued)

Enclosure 4.3 /O/A/s7oo/oooAbnormal Rad Levels/Radiological Effluent Page 5 of 5


4.3.A.3 Release of RadioactiveMaterial or Increases inRadiation Levels Withinthe Facility That ImpedesOperation of SystemsRequired to Maintain SafeOperations or to Establishor Maintain ColdShutdown.

OPERATING MODE: ALL

4.3.A.3-1 Valid reading on EMF-12greater than 15 mR/hr in theControl Room.

4.3.A.3-2 Valid indication of radiationlevels greater than 15 mR/hrin the Central Alarm Station(CAS) or Secondary AlarmStation (SAS).

4.3.A.33 Valid area EMF readingexceeds the levels shown inEnclosure 4.10.

END

Enclosure 4.4

UNUSUAL EVENT

Loss of Shutdown Functions


Iu!O/A/5700/000Page 1 of3

GENERAL EMERGENCY

4.4.A.1 Failure of ReactorProtection SystemInstrumentation to Completeor Initiate an AutomaticReactor Trip Once aReactor Protection SystemSetpoint Has Been Exceededand Manual TripSuccessful.

4.4.S.1 Failure of ReactorProtection SystemInstrumentation to Completeor Initiate an AutomaticReactor Trip Once aReactor Protection SystemSetpoint Has Been Exceededand Manual Trip WAS NOTSuccessful.

4.4.G.1 Failure of the ReactorProtection System toComplete an Automatic Tripand Manual Trip {jSuccessful and There isIndication of an ExtremeChallenge to the Ability toCool the Core.

OPERATING MODE: 1

Manual reactor trip from thecontrol room is successful andreactor power is less than 5%and decreasing.

Manual reactor trip from thecontrol room was notsuccessful in reducing reactorpower to less than 5% anddecreasing.

Manual reactor trip from thecontrol room was notsuccessful in reducing reactorpower to less than 5% anddecreasing.

(Continued) EITHER of the followingconditions exist:

• Core Cooling CSF-RED

• Heat Sink CSF-RED.

END

OPERATING MODE: 1,2,3 OPERATING MODE: 1

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

Valid reactor trip signalreceived or required andautomatic reactor tripwas not successful.

AND


AND

4.4.G.1-1 The following conditions exist:


AND

(Continued) AND

END

Enclosure 4.4 RP/O/A/5700/000Loss of Shutdown Functions Page 2 of 3


4.4.A.2 Inability to Maintain Plant 4.4.S.2 Complete Loss of Functionin Cold Shutdown. Needed to Achieve or

Maintain Hot Shutdown.OPERATING MODE: 5,6

OPERATING MODE: 1,2,3,44.4.A.2—1 Total loss of ND and/or RN

and/or KC. 4.4.S.2-1 Subcriticality CSF-RED.

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

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

• Inability to maintain Will Uncover Fuel in thereactor coolant temperature Reactor Vessel.

below 200°FOPERATING MODE: 5,6

OR— 4.4.S.3-1 Failure of heat sink causes loss

of cold shutdown conditions.• Uncontrolled reactor

coolant temperature rise to AND>180°F. —

Lower range Reactor VesselLevel Indication System(RVLIS) decreasing afterinitiation of NC systemmakeup.

(Continued)

Enclosure 4.4 1u/O/A!5700/000Loss of Shutdown Functions Page 3 of 3


4.4.S.3-2 Failure of heat sink causes lossof cold shutdown conditions.

AND

Reactor Coolant (NC) systemnarrow range level less than6 inches and decreasing afterinitiation of NC systemmakeup.

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

AND

Either train ultrasonic levelindication less than 6 inchesand decreasing after initiationof NC system makeup.

END

Enclosure 4.5 ip/O/A/57oo/ooo

UNUSUAL EVENT ALERT

Loss of Power

SITE AREA EMERGENCY

Page 1 of3

GENERAL EMERGENCY

4.5.U.1 Loss of All OffsitePower to EssentialBusses for Greater Than15 Minutes.


4.5.U.1-1 The following conditionsexist:

Loss of offsite power toessential buses ETA andETB for greater than15 minutes.

AND

Both emergency dieselgenerators are supplyingpower to their respectiveessential busses.

4.5.A.1 Loss of All OffsitePower and Loss of AllOnsite AC Power toEssential Busses DuringCold Shutdown OrRefueling Mode.

OPERATING MODE: 5,6, NoMode

4.5.A.1-1 Loss of all offsite andonsite AC power asindicated by:

Loss of power on essentialbuses ETA and ETB.

AND

Failure to restore power toat least one essential buswithin 15 minutes.

4.5.S.1 Loss of All OffsitePower and Loss of AllOnsite AC Power toEssential Busses.


4.5.S.1-1 Loss of all offsite andonsite AC power asindicated by:

Loss of power on essentialbuses ETA and ETB.

AND

Failure to restore power toat least one essential buswithin 15 minutes.

(Continued)

4.5.G.1 Prolonged Loss of All(Offsite and Onsite) ACPower.


4.5.G.1-1 Prolonged loss of alloffsite and onsite ACpower as indicated by:

Loss of power on essentialbuses ETA and ETB forgreater than 15 minutes.

AND

Standby ShutdownFacility (SSF) fails tosupply NC pump sealinjection OR CA supplyto Steam Generators.

AND

(Continued)(Continued)

(Continued)

Enclosure 4.5

Loss of Power1u/O/A/5700/000Page 2 of 3


OPERATING MODE: 5,6, NoMode

4.5.U.1-2 The following conditionsexist:Loss of offsite power toessential buses ETA andETB for greater than15 minutes.

AND

One emergency dieselgenerator is supplyingpower to its respectiveessential bus.

Continued

4.5.A.2 AC Power to EssentialBusses Reduced to aSingle Power Source forGreater Than 15Minutes Such That AnAdditional SingleFailure Could Result inStation Blackout.


4.5.A.2-1 The followmg conditionexists:

AC power capability hasbeen degraded to oneessential bus poweredfrom a single powersource for> 15 mm. dueto the loss of all but oneof:

SATASATBATCATDDIG ADIG B.

4.5.S.2 Loss of All Vital DCPower.


4.5.S.2-1 The following conditionsexist:

Loss of both unit relatedEVDA and EVDD bussesas indicated by busvoltage less than110 VDC.

AND

Failure to restore power toat least one required DCbus within 15 minutesfrom the time of loss.

END

At least one of thefollowing conditionsexist:

• Restoration of at leastone essential buswithin 4 hours is

j likely

• Indication ofcontinuingdegradation of corecooling based onFission ProductBarrier monitoring.

END

END

Enclosure 4.5 RP/O/A/5700/000Loss of Power Page 3 of 3


4.5.U.2 Unplanned Loss ofRequired DC PowerDuring Cold Shutdownor Refueling Mode forGreater than15 Minutes.

OPERATING MODE: 5,6

4.5.U.2-1 The following conditionsexist:

Unplanned loss of bothunit related EVDA andEVDD busses as indicatedby bus voltage less than110 VDC.

AND

Failure to restore power toat least one required DCbus within 15 minutesfrom the time of loss.

END

Enclosure 4.6

Fire/Explosion and Security Events

I.P!O/A/57oo/oooPage 1 of4


4.6.U.1 Fire Within Protected AreaBoundary NOTExtinguished Within15 Minutes of Detection ORExplosion Within theProtected Area Boundary.

OPERATING MODE: ALL

4.6.U.1-1 Fire in any of the followingareas NOT extinguishedwithin 15 minutes of controlroom notification orverification of a control roomfire alarm.

• Reactor Building• Auxiliary Building• Diesel Generator Rooms

Control Room• Standby Shutdown Facility• CAS

SASDoghousesFWSTTurbine BuildingService BuildingInterim Radwaste BuildingEquipment StagingBuilding

• ISFSL

4.6.A.1 Fire or Explosion Affectingthe Operability of PlantSafety Systems Required toEstablish or Maintain SafeShutdown.

OPERATING MODE: 1,2,3, 4, 5, 6

4.6.A.1-1 The following conditions exist:(includes non-security events)Fire or explosion in any of thefollowing areas:• Reactor Building• Auxiliary Building• Diesel Generator Rooms• Control Room• Standby Shutdown Facility• CAS• SAS• FWST• Doghouses (Applies in

Mode 1,2,3,4 only).

AND

4.6.S.1 HOSTILE ACTION withinthe PROTECTED AREA

OPERATING MODE: ALL

4.6.5.1-I A HOSTILE ACTION isoccurring or has occurred within thePROTECTED AREA as reported by theMiSTS Security Shift Supervision.

4.6.G.1 HOSTILE ACTIONResulting in Loss of PhysicalControl of the Facility.

OPERATING MODE: ALL

4.6.G.1-1 A HOSTILE ACTION hasoccurred such that plantpersonnel are unable to operateequipment required tomaintain safety functions.

4.6.G.1-2 A HOSTILE ACTION hascaused failure of Spent FuelCooling Systems andIMMINENT fuel damage islikely for a freshly off-loadedreactor core in pool.

END

END

•

•

(Continued)

(Continued)

Enclosure 4.6 RP/O/A/5700/000

UNUSUAL EVENT ALERT SITE AREA EMERGENCY

Page2of4

GENERAL EMERGENCY

4.6.U.l-2 Report by plant personnel ofan unanticipated explosionwithin the protected areaboundary resulting in visibledamage to permanentstructures or equipment or aloaded cask in the ISFSI.

4.6.U.2 Confirmed SECURITYCONDITION or ThreatWhich Indicates a PotentialDegradation in the Level ofSafety of the Plant.

OPERATING MODE: All

4.6.U.2-1 A SECURITY CONDITIONthat does NOT involve aHOSTILE ACTION asreported by the MNS SecurityShift Supervision.

4.6.U.2-2 A credible site specificsecurity threat notification.

4.6.U.2-3 A validated notification fromNRC providing information ofan aircraft threat.

END

One of the following:

Note: Only one train of a systemneeds to be affected ordamaged in order to satisfythis condition.

• Affected safety systemparameter indications showdegraded performance

• Plant personnel reportvisible damage topermanent structures orequipment within thespecified area.

4.6.A.2 Fire or Explosion Affectingthe Operability of PlantSafety Systems Required toEstablish or Maintain SafeShutdown.

OPERATING MODE: No Mode

4.6.A.2-1 The following conditions exist:(includes non-security events)Fire or explosion in any of thefollowing areas:• Spent Fuel Pool• Auxiliary Building.

AND

Fire/Explosion and Security Events

(Continued)

Enclosure 4.6 RP/O!A!5700/000Fire/Explosion and Security Events Page 3 of 4

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCYOne of the following:

Note: Only one train of a systemneeds to be affected ordamaged in order to satisfythis condition.• Spent Fuel Pool level

and/or temperature showdegraded performance

• Plant personnel reportvisible damage topermanent structures orequipment supportingSpent Fuel Pool Cooling.

4.6.A.3 HOSTILE ACTION Withinthe OWNERCONTROLLED AREA orAirborne Attack Threat.

OPERATING MODE: ALL

4.6.A.3-1 A HOSTILE ACTION isoccurring or has occurredwithin the OWNERCONTROLLED AREA asreported by the MNS SecurityShift Supervision.

(Continued)

Enclosure 4.6 RP,/O/A/5700/000Fire/Explosion and Security Events Page 4 of 4


4.6.A.3-2 A validated notification fromNRC of an airliner attackthreat within 30 minutes of thesite.

END

Enclosure 4.7 1P/O/A/57oo/ooo

UNUSUAL EVENT

Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety


Page 1 of4

GENERAL EMERGENCY

4.7.U.1 Natural and DestructivePhenomena Affecting theProtected Area.

OPERATING MODE: ALL

4.7.U.1-1 Tremor felt and valid alarm onthe Syscom SeismicMonitoring System (OACM1D2422).

4.7.U.1-2 Report by plant personnel oftornado striking withinprotected areaboundary/ISFSI.

4.7.U.1-3 Vehicle crash into plantstructures or systems withinprotected areaboundary/ISFSI.

4.7.U.l-4 Report of turbine failureresulting in casing penetrationor damage to turbine orgenerator seals.

(Continued)

4.7.A.1 Natural and DestructivePhenomena Affecting thePlant Vital Area.

OPERATING MODE: ALL

4.7.A.1-1 Valid “OBE Exceeded” Alarmon 1AD-l3, E-7

4.7.A.1-2 Tornado or high winds:

Tornado striking plantstructures within the vitalarea:

• Reactor Building• Auxiliary Building• FWST• Diesel Generator Rooms• Control Room• Standby Shutdown

Facility• Doghouses• CAS• SAS.

Sustained winds 74 mph for> 15 minutes. {4}

4.7.S.1 Control Room EvacuationHas Been Initiated and PlantControl Cannot BeEstablished.

OPERATING MODE: ALL

4.7.S.1.-1 The following conditionsexist:

Control Room evacuation hasbeen initiated perAP/l (2)/A!5500!O 17, orAP/l (2)!A!5500/024. {3]

Control of the plant carmot beestablished from the AuxiliaryShutdown Panel or theStandby Shutdown Facilitywithin 15 minutes.

4.7.G.1 Other Conditions ExistingWhich in the Judgement ofthe EmergencyCoordinator/EOF DirectorWarrant Declaration ofGeneral Emergency.

OPERATING MODE: ALL

4.7.G.1-1 Other conditions exist whichin the Judgement of theEmergencyCoordinator/EOF Directorindicate:

(1) actual or imminentsubstantial core degradationwith potential for loss ofcontainnent,

OR

(2) potential foruncontrolled radionuclidereleases. These releases canreasonably be expected toexceed EnvironmentalProtection AgencyProtective Action Guidelinelevels outside the siteboundary.

AND

OR

(Continued)

(Continued)END

Enclosure 4.7

UNUSUAL EVENT

Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety


1P/O/A/57oo/oooPage2of4

GENERAL EMERGENCY

4.7.U.15 Independent Spent Fuel Casktipped over or dropped greaterthan 12 inches.

4.7.U.1-6 Uncontrolled flooding in theISFSI area.

4.7.A.1-3 Visible structural damagecaused by either:• Vehicle crashes, OR• Turbine failure generated

missiles, OR• Other catastrophic events

4.7.S.2 Other Conditions ExistingWhich in the Judgement ofthe EmergencyCoordinator/EOF DirectorWarrant Declaration of SiteArea Emergency.

4.7.U.l-7 Tornado generated missile(s)impacting the ISFSI.

4.7.U.2 Release of Toxic orFlammable Gases DeemedDetrimental to SafeOperation of the Plant.

OPERATING MODE: ALL

4.7.U.2-1 Report or detection of toxic orflammable gases that couldenter within the site boundaryin amounts that can affect safeoperation of the plant.

4.7.U.2-2 Report by Local, County orState Officials for potentialevacuation of site personnelbased on offsite event.

on any of the following plantstructures:

• Reactor Building• Auxiliary Building• FWST• Diesel Generator Rooms• Control Room• Standby Shutdown

FacilityDoghousesCAS

• Ultimate heat sink(Standby Nuclear ServiceWater Pond Dam andDikes).

OPERATING MODE: ALL

4.7.S.2-1 Other conditions exist whichin the Judgement of theEmergency Coordinator/EOFDirector indicate actual orlikely major failures of plantfunctions needed forprotection of the public.

END

•

•

• SAS

(Continued)

(Continued)

Enclosure 4.7 RP/O/A/5700/000Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 3 of 4

4.7.U.3 Other Conditions ExistingWhich in the Judgement ofthe EmergencyCoordinator/EOF DirectorWarrant Declaration of anUnusual Event.

OPERATING MODE: ALL

4.7.U.3-1 Other conditions exist whichin the judgernent of theEmergency Coordinator/EOFDirector indicate a potentialdegradation of the level ofsafety of the plant.

END

4.7.A.2 Release of Toxic orFlammable Gases Within aFacility Structure WhichJeopardizes Operation ofSystems Required toMaintain Safe Operations orto Establish or MaintainCold Shutdown.

OPERATING MODE: ALL

Note: Structures for the below EALs:• Reactor Building• Auxiliary Building• Diesel Generator Rooms• Control Room• Standby Shutdown

Facility• Doghouses• CAS• SAS.

4.7.A.2-1 Report or detection of toxicgases within a FacilityStructure in concentrationsthat will be life threatening toplant personnel.

4.7.A.2-2 Report or detection offlammable gases within aFacility Structure in concentrations that will affect the safeoperation of the plant.


(Continued)

Enclosure 4.7 RP/O/A/5700I000Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 4 of 4


4.7.A.3 Control Room EvacuationHas Been Initiated.

OPERATING MODE: ALL

4.7.A.3-l Control Room evacuation hasbeen initiated perAP/l (2)/A!5500/Ol 7, orAP/l (2)/A15500/024. {3}

4.7.A.4 Other Conditions ExistingWhich in the Judgement ofthe EmergencyCoordinator/EOF DirectorWarrant Declaration of anAlert.

OPERATING MODE: ALL

4.7.A.4-1 Other conditions exist whichin the Judgement of theEmergency Coordinator/EOFDirector indicate that plantsafety systems may bedegraded and that increasedmonitoring of plant functionsis warranted.

END

Enc1oure4.8 RP/O/A/5700/000Definitions/Acronyms Page 1 of 4

ALERT - Events are in process or have occurred which involve an actual or potential substantialdegradation of the level of safety of the plant or a security event that involves probable life threateningrisk to site personnel or damage to site equipment because of hostile action. Any releases are expectedto be limited to small fractions of the EPA Protective Action Guideline exposure levels

ALL - (As relates to Operating Mode Applicability) — At all times.

BOMB - Refers to an explosive device suspected of having sufficient force to damage plant systems orstructures.

CIVIL DISTURBANCE - A group of persons violently protesting station operations or activities at thesite.

CONFINEMENT BOUNDARY - The barrier(s) between areas containing radioactive substances andthe environment.

EXPLOSION - A rapid, violent unconfined combustion, or a catastrophic failure ofpressurizedlenergized equipment that imparts energy of sufficient force to potentially damage permanentstructures, systems or components.

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 oroverheated electrical equipment do not constitute fires. Observation of flames is preferred but is NOTrequired if large quantities of smoke and heat are observed. An electrical breaker flash that creates hightemperatures for a short duration and merely localized scorching to that breaker and its compartmentshould not be considered a fire.

GENERAL EMERGENCY - Events are in process or have occurred which involve actual or imminentsubstantial core degradation or melting with potential for loss of containment integrity or hostile actionthat results in an actual loss of physical control of the facility. Releases can be reasonably expected toexceed EPA Protective Action Guidelines exposure levels offsite for more than the immediate site area.

HOSTAGE - A person(s) held as leverage against the station to ensure demands will be met by thestation.

HOSTILE ACTION - An act toward a NPP or its personnel that includes the use of violent force todestroy equipment, take HOSTAGES, and / or intimidate the licensee to achieve an end. This includesattack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used todeliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILEACTION should not be construed to include acts of civil disobedience or felonious acts that are not partof a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities,(i.e., this may include violent acts between individuals in the owner controlled area).

- EneIosure4.8 - /O/A/57oo/oooDefinitions/Acronyms Page 2 of 4

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or bystealth and deception, equipped with suitable weapons capable of killing, maiming, or causingdestruction.

IMMINENT - Mitigation actions have been ineffective, additional actions are not expected to besuccessful, and trended information indicates that the event or condition will occur. WhereIMMINENT time frames are specified, they shall apply.

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

iNTRUSION - A person(s) present in a specified area without authorization. Discovery of a BOMB in aspecified area is indication of INTRUSION into that area by a HOSTILE FORCE.

ISFSI - Independent Spent Fuel Storage Installation.

NO MODE - Defueled.

PROJECTILE - An object directed toward a NPP that could cause concern for its continued operability,reliability, or personnel safety.

PROLONGED - A duration beyond normal limits, defined as “greater than 15 minutes!! or as determinedby the judgement of the Emergency Coordinator.

PROTECTED AREA - Typically the site specific area which normally encompasses all controlled areaswithin the security PROTECTED AREA fence.

REACTOR COOLANT SYSTEM (RCS/NCS) LEAKAGE - RCS Operational Leakage as defined inthe Technical Specification Basis B 3.4.13.

RUPTURED - (As relates to Steam Generator) - Existence of primary to secondary leakage of amagnitude sufficient to require or cause a reactor trip and safety injection.

SABOTAGE - Deliberate damage, misalignment, or mis-operation of plant equipment with the intent torender the equipment inoperable. Equipment found tampered with or damaged due to maliciousmischief may not meet the definition of SABOTAGE until this determination is made by securitysupervision.

SECURITY CONDITION - Any Security Event as listed in the approved security contingency plan thatconstitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation tothe level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

SIGNIFICANT TRANSIENT- An unplanned event involving one or more of the following: (I)automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full electricalload; (3) reactor trip, (4) safety injection, (5) thermal power oscillations 1O%.

Enclosure 4.8 iwOiA/57oo/oooDefinitions/Acronyms Page 3 of 4

SITE AREA EMERGENCY - Events are in process or have occurred which involve actual or likelymajor failures of plant functions needed for protection of the public or hostile action that results inintentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likelyfailure of or; (2) that prevent effective access to equipment needed for the protection of the public. Anyreleases are not expected to exceed EPA Protective Action Guideline exposure levels beyond the siteboundary.

SITE BOUNDARY - That area, including the protected area, in which Duke Power Company has theauthority 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 EFFECTIVE DOSE EQUIVALENT (TEDE) - The sum of external dose exposure to aradioactive plume, to radionuclides deposited on the ground by the plume, and the internal exposurefrom 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 normaloperations, testing, or maintenance. Events that result in corrective or mitigative actions being taken inaccordance with abnormal or emergency procedures are UNPLANNED.

UNUSUAL EVENT - . Events are in process or have occurred which indicate a potential degradation ofthe level of safety of the plant or indicate a security threat to facility protection has been initiated. Noreleases of radioactive material requiring offsite response or monitoring are expected unless furtherdegradation of safety systems occurs.

VALID - An indication or report or condition is considered to be VALID when it is conclusively verifiedby: (1) an instrument channel check, or (2) indications on related or redundant instrumentation, or (3) bydirect observation by plant personnel such that doubt related to the instrument’s operability, thecondition’s existence or the report’s accuracy is removed. Implicit in this definition is the need for timelyassessment.

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 withoutmeasurements, testing, or analyses. Damage is sufficient to cause concern regarding the continuedoperability or reliability of affected safety structure, system, or component. Example damage:deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering.

• Enchure‘ RP/O/A[5’700/000Definitions/Acronyms Page 4 of 4

VITAL AREA - Areas within the PROTECTED AREA that house equipment important for nuclearsafety. Access to a VITAL AREA is allowed only if an individual has been authorized to be in that areaper the Security plan, therefore VITAL AREA is a Security term.

Eneloure ‘‘• aP/O/Ai5700/000Emergency Declaration Guidelines Page 1 of 2

THE FOLLOWING GUIDANCE IS TO BE USED BY THE EMERGENCY COORDINATOR iNASSESSING EMERGENCY CONDITIONS.

• The Emergency Coordinator shall review all applicable initiating events to ensure properclassification.

• The BASIS Document (located in Section D of the McGuire Nuclear Site Emergency Plan) isavailable 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 beclassified on the emergency notification form. Information relating to the problem on the other unitwill be captured on the emergency notification form line 13 remarks section.

• The Affected Unit(s) on Line 11 is tied to the EAL (IC) Number and EAL (IC) Description on Line4. Certain events could occur at the plant site such that multiple units are affected. These mayinclude Abnormal Rad Levels/Radiological Effluents; Fire/Explosion and Security Events; andNatural Disasters, Hazards, and Other Conditions Affecting Plant Safety. This shall be consideredwhen evaluating the accuracy of the Unit designation. {PIP 0-M97-463 8 } If the initiating event putsmore than one unit in the same Emergency Classification (example Alert), then the unit designationmay be either the Affected Unit numbers or All. If the initiating event drives one unit to a higherclassification, then the unit with the higher classification should be listed as Affected Unit. {PIPs M03-3294 and C-04-2586}

• The EAL (IC) Number and EAL (IC) Description provided on Line 4 of the emergency notificationform should be based on the highest emergency classification that applies. Other classifiable eventsshould be included on Line 13, Remarks, on the emergency notification form, but not given an EALnumber. {PIPs M-03-3294 and C-04-2586}

• If an event occurs, and a lower or higher plant operating mode is reached before the classificationcan 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 orhigher. An event that is recognized at cold shutdown or lower shall not be classified using the fissionproduct barrier matrix. Reference would be made to the additional enclosures that provideemergency 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 classificationis made when the Emergency Coordinator assessment concludes that the specified durationis exceeded or will be exceeded (i.e. condition cannot be reasonably corrected before theduration elapses), whichever is sooner.

Ene1oure 4.9- RP/O/A/57OO/000

Emergency Declaration Guidelines Page 2 of 2

2. If a plant condition exceeding EAL criteria is corrected before the specified duration time isexceeded, the event is NOT classified by that EAL. Lower Severity EALs, if any, shall bereviewed for possible applicability in these cases.

3. If a plant condition exceeding EAL criteria is not recognized at the time of occurrence, but isidentified well after the condition has occurred (e.g. as a result of routine log or recordreview) and the condition no longer exists, an emergency shall NOT be declared. Reportingunder 1OCFR5O.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 moresevere than earlier believed.

4. If an emergency classification was warranted, but the plant condition has been correctedprior to declaration and notification, the following are applicable: {2}

a. For UNUSUAL EVENT, the emergency shall be declared and the condition shallbe reported. The event should be terminated in a follow-up notification as soon astime permits, but within one hour.

b. For ALERT, SITE AREA EMERGENCY, and GENERAL EMERGENCY, theemergency shall be declared and the Emergency Response Organization shall beactivated. The TSC Emergency Coordinator shall be responsible for terminatingthe emergency as soon as time permits when appropriate.

c. The Control Room Emergency Coordinator (Operations Shift Manager) shallensure that any required follow-up notifications are conducted as required prior toactivation of the TSC.

DETERMINATION OF “EVENT TIME” (TIME THE 15 MINUTE CLOCK STARTS)

Event Time is the time at which indications become available that an EAL has been exceeded.

2. Event Time is the time the 15 minute clock starts for classification.

3. The event classification time shall be entered on the emergency notification form.

MOMENTARY ENTRY INTO A HIGHER CLASSIFICATION

If while in an emergency classification, the specified EALs of a higher classification are metmomentarily, and in the judgment of the Emergency Coordinator are not likely to recur, the entry into thehigher classification must be acknowledged. Acknowledgment is performed as follows:

If this condition occurs prior to the initial notification to the emergency responseorganization and off site agencies, the initial message should note that the site is currentlyin the lower classification, but had momentarily met the criteria for the higherclassification. It should also be noted that plant conditions have improved and stabilizedto the point that the criteria for the higher classification are not expected to be repeated.

Enclosure 4.10 RP/O/A/5700/000Radiation 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. incore detector movement, radwastecontainer movement, depleted resin transfers, etc.)

Detector Elevation Column Identifier Unusual AlertEvent mR/hrmR/hr

1EMF-1 695’ FF, GG-56 Aux. Bldg. Corridor 500 50001EMF-5 716’ FF-54 Unit 1 NM Sample Room 600 50001EMF-8 733’ HH-56 Aux. Bldg. Corridor 100 50001EMF-10 750’ LL-56 Aux. Bldg. Corridor 100 50001EMF-13 775’ QQ-56 Shift Lab/Count Room 100 50001EMF-17 786’ N/A Unit 1 Spent Fuel Pool Refueling Bridge 100 50002EMF-1 716’ BE, FF-58 Unit 2 NM Sample Room 300 50002EMF-4 786’ N/A Unit 2 Spent Fuel Pool Refueling Bridge 100 50002EMF-9 767’ JJ-59 Aux. Bldg. Corridor 100 5000

Enoi& 4.11 RP/O/A/5’70O/000Commitment Reference for Emergency Action Levels Page 1 of 1

{1} PIP-M-OO-2138, CA# 18{2} PIP-M-02-O 187, CA # 6{3} PIP-M-O1-2860, CA #2{4} Pll-M-O3-4281, CA #3{5} PIP-M-05-3403, CA#3, multiple changes in enclosure 4.6

MCEI-0400..232PaeZ9of 32

Revision UMeGuire 1 Cyde 21 Côrt Operating Limits Report

2.16 Borated Water Sources — Shutdown (SLC 16.9.14)

2.1 .l VoJumc and boroe coacentzians for the &ñc Acid Tank (BAT) .nd üie RefuclIngWater Storage Tank (RWST) thiring MODE 4 with any RCS cold leg temperature

300 °F and MODES 5 and 6.

Parameter Limit

BAT tninimm contained berated wacr volume 10,599 gallons13.6% L.evel

[Not; When cycle boruup is> 455 EFPD Figure 6 maybe used to

Lh1 the rcquir’ed BAT mInimum IeveL -

BAT miiiiimuxn boron concentration 7,000 ppm

BAT minimum water votume reqiired to 2,Gô.gallons..maintain 5DM al 7,0130 ppm

RWST rninianum contained borated water 47,70 gallonsvolume 41 inches

RWST minimum boron concentration 2,675 ppm

RWST minimum water vourne required to 8200 gallonsmthitair SDM al 2,675 ppm

MCufre 1. Cyde 21 Core Operating Lhnit Report

figure 6

MJE.4J4UU-23ZPac3I af32

RevisioiO

• 25.0

41-J

rkirk Acid Serage Tank Indlcted Level VersusRCS Baron Concentration

(Valid When Cycle Eurnup Is> 455 EFPD)

This igureincbdas dditiona[volunieollsted in SLC L69t4 and 16.9ll

4I

5-oRS Boron

CQnceatrat1n BAT Luvel(ppi)00 370

300<500 33.0500<7GG 28.O

700 ID0 20.1000<13001 16.

Opvt4ion

1Jtab1ej

5O

0 200 400 600 600 1000 1200 1400 10O 1500 2000 2200 2400 3q50 2500

RGS Boron Concentration (ppmb)

Examination KEYfor: 2011 MNS SRO NRC Examination

Question AnswerNumber

C

2 C

3 C

4 A

5 A

6 A

7 C

8 B

9 A

10 B

11 C

12 B

13 C

14 A

15 B

16 C

17 A

18 A

19 B

20 C

21 C

22 B

23 D

24 C

25 C

Printed 6/9/2011 1:06:45 PM Page 1 of4



26 C

27 D

28 A

29 C

30 A

31 C

32 A

33 D

34 D

35 C

36 B

37 C

38 D

39 C

40 C

41 A

42 B

43 D

44 A

45 A

46 D

47 B

48 D

49 B

50 C

Printed 6/9/2011 1:06:45 PM Page 2 of 4



51 D

52 D

53 D

54 B

55 B

56 C

57 D

58 B

59 C

60 C

61 D

62 C

63 C

64 C

65 A

66 C

67 C

68 A

69 B

70 D

71 A

72 B

73 D

74 D

75 D

Printed 6/9/2011 1:06:45 PM Page 3 of 4



76 C

77 C

78 A

79 A

80 C

81 D

82 A

83 C

84 C

85 B

86 A

87 C

88 A

89 C

90 C

91 B

92 A

93 C

94 C

95 A

96 C

97 D

98 B

99 B

100 D

Printed 6/9/2011 1:06:45 PM Page 4 of4

Documents

McGuire Initial Exam 2011-301 Final SRO Written Exam. · 2013. 7. 8. · the inadvertent closure of IRN-277B (RB Non Ess Ret Cont Outside Isol) At 1400, the following NC pumps indications