Text

Initial Exam 2011-301 Final SRO Written Exam
	ML112280604
Person / Time
Site:	Mcguire, McGuire
Issue date:	08/16/2011
From:	; NRC/RGN-II
To:	;
References
	Download: ML112280604 (167)
	v • d • e

U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

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

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

Applicants Signature Resu Its RO/SRO-OnIy/Total Examination Values / I PcJilts Applicants Score I I Points Applicants Grade I I Percent

McGuire Nuclear Station Question: I 21111 MNS SRO NRC Examination I 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°F Winding 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 pump that would have to be manually secured in accordance with AP-08, (Malfunction of NC Pump)?

A. IA B. lB C. 1C D. 1D Page 1 of 100

McGuire Nuclear Station Question: 2 2011 MNS SliO NRC Examination 1 point)

Given the following conditions on Unit 1:

An NC pump is to be started for a unit heatup to normal operating temperature and 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. will

2. 200 B. I. will

2. 350 C. 1. willnot

2. 200 D. 1. willnot

2. 350 Page 2 of 100

McGuire Nuclear Station Question: 3 2011 MNS SRO NRCExamination 1 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) to facilitate 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 on PZR pressure and what action is required per Gen Enc G-1 Enc 3 to control pressure?

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

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

C. Pressure would INCREASE Deenergize PZR heaters as required to control pressure D. Pressure would DECREASE Energize additional PZR Heaters as required to control pressure Page 3 of 100

McGuire Nuclear Station Question: 4 2011 MNS SRO NRC Examination 1 point)

Given the following conditions on Unit 1:

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 be used 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 Question:

5 2011MNS SRO NRC Examination 1 point)

Unit I is shutdown with the following conditions:

1A and lB ND trains are in service providing shutdown cooling in accordance with 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 the Operators 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 Page5oflOO

McGu ire Nuclear Station Question: 6 2U11MNSSRONRCExamiiatioi 1 point)

Given the following conditions on Unit 1:

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 will receive 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) AND 1 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) AND I 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) AND 1 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) AND 1 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 Static Question: 7 2011MNSSRONRCExamInar (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 N 2 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 N 2 injection into the NC system?

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

A. 1. higher

2. higher B. 1. lower

2. higher C. 1. higher

2. lower D. 1. lower

2. lower Page 7 of 100

McGuire Nuclear Station Question:

8 2011 MNSSRONRCExainination 1 point)

Given the following conditions on Unit 1:

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-HI Ui 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 pressure rises to a MAXIMUM of (1)

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

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

A. 1. 85PSIG

2. drain the PRTto the NCDT B. 1. 100PSIG

2. drain the PRT to the NCDT C. 1. 85PSIG

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

2. have Chemistry vent the PRT to the Waste Gas system Page8oflOO

McGuire Nuclear Station Question: 9 21111 MNS SRO NRC Examination 1 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 Backup Heaters 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 on Pressurizer Pressure Select 1 and Select 2 Assuming no operator actions, what effect does this combination of channel failures have 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 Station Question: 10 2011 MNS SRO NRC Examination 1 point)

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

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

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

A. 1. Instrument Power

2. E-0 (Reactor Trip or Safety Injection)

B. 1. Instrument Power

2. AP-16 (Malfunction of Nuclear Instrumentation)

C. 1. Control Power

2. E-0 (Reactor Trip or Safety Injection)

D. 1. Control Power

2. AP-16 (Malfunction of Nuclear Instrumentation)

PagelOoflOO

McGuire Nuclear Station Question:

11 2011 MNS SRO NRC Examination 1 point)

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)

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

A. 1. NC system pressure increases

2. DNB limits B. 1. NC system pressure increases

2. fuel heat generation rate limits C. 1. power is lost to N-41 lower detector

2. DNB limits D. 1. power is lost to N-41 lower detector

2. fuel heat generation rate limits Page 11 of 100

McGuire Nuclear Station Question: 12 2UiIMNSSRONRCExamination M point)

Given the following conditions on Unit 2:

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 on 2EKVA?

A. The reactor will automatically trip.

SI signal equipment on both trains will automatically start due to redundant power supplies.

B. The reactor will automatically trip.

SI signal equipment on A train ONLY will not automatically start due to the loss of power to the SSPS output cabinet.

C. The reactor will not automatically trip If an SI signal is received, equipment on both trains will automatically start due to redundant power supplies.

D. The reactor will not automatically trip.

If an SI signal is received, equipment on A train ONLY will not automatically start due to the loss of power to the SSPS output cabinet.

Page 12 of 100

McGuire Nuclear Station Question: 13 2011 MNSSRUNRCExammnation (1 point)

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 Injection logic is placed in BYPASS, then the logic becomes (2) channels.

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

A. 1. 2of3

2. lof2 B. 1. 2of4

2. lof3 C. 1. 2of3

2. 2of2 D. 1. 2of4

2. 2of3 Page 13 of 100

McGuire Nuclear Station Question:

14 2011 MNS SRO NRC Examinütion (1 point)

Given the following conditions on Unit 1:

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 Pagel4oflOO

McGuire Nuclear Station Question: 15 2U11 MNS SRO NRC Ecamlnatioh 1 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 NS System?

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

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

A. 1. There would be no Train A Containment Spray Flow

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

B. 1. There would be no Train A Containment Spray Flow

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

C. 1. ONLY one of the two Train A NS Spray Headers will have flow

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

D. 1. ONLY one of the two Train A NS Spray Headers will have flow

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

Page 15 of 100

McGu ire Nuclear Station Question:

16 2011 MNS SRO NRC Examination (1 point)

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 LOWER

2. Reactor Trip and Turbine Trip ONLY B. 1. Tavg is HIGHER reactor power is LOWER

2. Reactor Trip and closure of the MSIVs C. 1. Tavg is LOWER reactor power is HIGHER

2. Reactor Trip and Turbine Trip ONLY D. 1. Tavg is LOWER reactor power is HIGHER

2. Reactor Trip and closure of the MSIVs Pagel6oflOO

McGu ire Nuclear Station Question:

17 2U11 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

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 MSIVs for the given conditions?

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 Main Steam Isolation signal ONLY.

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

Page 17 of 100

McGuire Nuclear Station Question: 18 2OJIMNS SRO NRCExaminátion 1 point)

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

A. Both Feedwater Pumps trip CF to CA Nozzle Isolations close B. Both Feedwater Pumps trip Main Steam Isolation Valves close C. Both Feedwater Pumps speed goes to 2800 RPM CF to CA Nozzle Isolations close D. Both Feedwater Pumps speed goes to 2800 RPM Main Steam Isolation Valves close Pagel8oflOO

McGuire Nuclear Station Question:

19 2U1JMNSSRONRCExam1natIon 1 point)

Given the following on Unit 1:

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.AandBSGs

2. 450 GPM B. 1. CandDSGs

2. 450 GPM C. 1.AandBSGs

2. 900 GPM D. 1. CandDSGs

2. 900 GPM Pagel9oflOO

McGuire Nuclear Station Question:

20 2011 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

Bus 1EVDA has lost power The breakerfor 1A NV pump Which ONE (1) of the following completes the statement above?

A. automatically trips open but can be closed locally after manually charging the springs B. can only be operated locally but the springs must be manually charged prior to each opening and closing operation C. can only be operated locally but only closing operations require manually charging the springs D. can be opened one time from the control room but all further operations must be performed locally Page 20 of 100

McGu ire Nuclear Station Question: 21 2011 MNS SRO NRC Examination I point)

Given the following plant conditions:

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 current source(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

McGuire Nuclear Station Question:

22 2011 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

. 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 no longer 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 Question:

23 2(111 MNS SRO NRC Examination 1 point)

Given the following conditions on Unit 1:

. 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 0

0 O.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 Crankcase Vacuum is...

A. 14:10 B. 14:15 C. 14:20 D. 14:25 Page 23 of 100

McGuire Nuclear Station Question: 24 2OI1MNSSRONRCExam1naIion I point)

Given the following conditions on Unit 1:

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 the DG Cooling Water (KD) System?

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

A. 1. VG After Coolers

2. 11.5 inches B. 1. Air Intake System Intercooler

2. 25 inches C. 1. Air Intake System Intercooler

2. 11.5 inches D. 1. VG After Coolers

2. 25 inches Page 24 of 100

McGu ire Nuclear Station Question: 25 2011 MNS SRO NRC Exambwtion 1 point)

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 the monitor 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

McGuire Nuclear Station Question: 26 2011 MNS SRO NRC Examination 1 point)

Given the following conditions on Unit 2:

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 being supplied 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 SRUNKC Examination Question: 27 I point)

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 MANUALLY initiates 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 Station Question: 28 2O1IMNSSRUNRCExamInatIon M point)

In accordance with FR-Z.1 (Response to High Containment Pressure), to place ND Spray 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)

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

A. 1. 3PSIG

2. 50 minutes B. 1. 1OPSIG

2. 50 minutes C. 1. 3PSIG

2. 60 minutes D. 1. 1OPSIG

2. 60 minutes Page 28 of 100

McGu ire Nuclear Station Question: 29 2011 MNS SRO NRC Examination 1 point)

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 the inherent response to the transient described above?

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

B. PZRoutsurge Some of the liquid in the PZR would flash to steam which will limit the pressure decrease.

C. PZRinsurge Some of the steam volume in the PZR would condense limiting the pressure increase.

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

Page 29 of 100

McGuire Nuclear Station Question: 30 2011 MNS SRO NRUffxainiithtion 1 point)

Given the following conditions on Unit 1:

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 Heaters operated from the ASP would trip on a (2)

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

A. 1.A&BONLY

2. Safety Injection B. 1.A&BONLY

2. Pressurizer Low-Low Level C. &D 1

1.A,B

2. Safety Injection D. 1.A,B,&D

2. Pressurizer Low-Low Level Page 30 of 100

McGuire Nuclear Station Question:

31 2011 MNS SRU NRC ExaminatIon 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 the statements above?

A. 1. Non-Urgent

2. will B. 1. Non-Urgent

2. will not C. 1.Urgent

2. will D. 1.Urgent

2. will not Page3l of 100

McGuire Nuclear Station Question: 32 2011 MNS SRONRCExan,inatkin 1 point)

Given the following conditions on Unit 1:

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 Tave detector failures required to cause a DCS Alternate Action on the Tavg Reactor Temperature 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 to shifting CRD Select from MANUAL to AUTO?

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

A. 1.2

2. +/-1°F B. 1.2

2. +/-3°F C. 1.3

2. +/-1°F D. 1.3

2. +/-3°F Page 32 of 100

McGu ire Nuclear Station Question: 33 2011 MNS SRO NRC Examthation 1 point)

Given the following on Unit 1:

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 the alarms 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 Station Question: 34 2U11 MNS SRO NRCExmilnation 1 point)

Given the following conditions on Unit 2:

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 leak on the Spent Fuel Pool Cooling system?

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

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 Station Question: 35 2tfll MNS SRU NRC Examination 1 point)

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

ND Additional Requirements A. One Loop OPERABLE and Reactor subcritical for a minimum of in operation 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months B. Two Loops OPERABLE and A minimum of one door in each airlock ONE in operation closed C. One Loop OPERABLE and Equipment Hatch held in place by a in operation minimum of four bolts D. Two Loops OPERABLE and One OPERABLE VCJYC train ONE in operation Page 35 of 100

McGuire Nuclear Station Question: 36 21111 MNS SRONRCExamination 1 point)

Given the following conditions on Unit 2:

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 target temperature 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 Question: 37 2011 MNS SRONRC Examination 1 point)

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

2 concentration is 4.4%

H 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 H 2 concentration is less than a MAXIMUM of (2)

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

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 Question:

38 2011 MNS SRONRCExamintion 1 point)

Given the following plant conditions:

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 VI pressure reached a MINIMUM of (1) AND currently is (2) ?

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

A. 1. 85P51G

2. open B. 1. 85PSIG

2. closed C. 1. 9OPSIG

2. open D. 1. 9OPSIG

2. closed Page 38 of 100

McGuire Nuclear Station Question: 39 2tfli MNS SRO NRC Examimaion I point)

Given the following conditions on Unit 1:

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 be manually tripped, the crew will NEXT attempt to (2)

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

A. 1. throttle

2. close the MSIVs AND MSIV bypasses B. 1. governor

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

2. place the turbine in MANUAL AND close the governor valves D. 1. governor

2. close the MSIV5 AND MSIV bypasses Page 39 of 100

McGuire Nuclear Station Question: 40 2OlIMNSSRONRCExainination I point)

Given the following conditions on Unit 1:

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 indicate a 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 Question:

41 2011 MNS SRUNRCExamintion 1 point)

Given the following conditions on Unit 1:

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

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

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 to E-1 (Loss of Reactor or Secondary Coolant)?

SG5 NC Pumps Required for Running? Heat Removal?

A. YES YES B. YES NO C. NO YES D. NO NO Page 41 of 100

McGuire Nuclear Station Question: 42 2011 MNS SRO NRC Examination 1 point)

Given the following conditions on Unit 2:

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-tubes where the steam condenses and re-enters the core area via the SIG (2)

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

Lfl A. 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 Question:

43 2IlIlMNSSRONRCExaminatin 1 point)

Given the following initial conditions on Unit 2:

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 BREAKERS indicate (1)

The Unit 2 Reactor Trip breakers (2)

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

A. 1. CLOSED

2. are TRIPPED B. 1. CLOSED

2. remain CLOSED C. 1. OPEN

2. are TRIPPED D. 1. OPEN

2. remain CLOSED Page 43 of 100

McGuire Nuclear Station Question: 44 2011 MNS SRUNRUExamination 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 flow would be (1)

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

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

A. 1. 26GPM

2. SSF B. 1. 26GPM

2. AuxSID Panel C. I. 35GPM

2. SSF D. 1. 35GPM

2. Aux S/D Panel Page 44 of 100

McGu ire Nuclear Station Question: 45 21111 MNS SRO NRC Examination 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 is determined by monitoring (2)

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

A. 1. cavitating

2. Core ExitT/Cs B. 1. cavitating

2. Wide Range T-hot C. 1. operating at a runout condition

2. Core ExitT/Cs D. 1. operating at a run out condition

2. WideRangeT-hot Page 45 of 100

McGuire Nuclear Station Question: 46 2011 MNSSRONRCExamination 1 point)

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

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-energize the 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 Question: 47 2Ol1MNSSROiTRCExamintion I point)

Given the following conditions on Unit 1:

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 MAXIMUM pressure of (1) PSIG.

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

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

A. 1. 2500

2. Main Steam Safety Valves B. 1. 2735

2. Main Steam Safety Valves C. 1. 2500

2. Condenser Steam Dumps D. 1. 2735

2. Condenser Steam Dumps Page 47 of 100

McGuire Nuclear Station Question: 48 2011 MNSSRO NRCExaminatiim 1 point)

Given the following conditions on Unit 2:

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 Pressurizer Sprays are ineffective at reducing NC system pressure, the crew shall FIRST attempt to depressurize the NC system using (2)

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

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 Question: 49 2OITMNS SRONRUExamintion 1 point)

Given the following:

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

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

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

A. 1. 1 second

2. MSIV bypass valves ONLY B. 1. 1 second

2. MSIV bypass valves AND the S/G PORVs C. 1. 2seconds

2. MSlVbypass valves ONLY D. 1. 2seconds

2. MSIV bypass valves AND the S/G PORVs Page 49 of 100

McGuire Nuclear Station Question: 50 2011 MNS SROJVRC Examination I point)

Given the following conditions on Unit 1:

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)

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

A. 1. 0200

2. 0205 B. 1. 0205

2. 0205 C. 1. 0200

2. 0215 D. 1. 0205

2. 0215 Page 50 of 100

McGuire Nuclear Station Question: 51 2011 i1INS SRO NRCExamintion I point)

Given the following initial conditions:

Unit I and 2 have been in a Loss of Offsite Power condition for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months

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 Thts are STABLE 0

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 current state 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 board AND 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 PORV AND increase CA flow by notifying NLO to throttle CA valves locally.

Page 51 of 100

McGuire Nuclear Station Question:

52 2UHMNSSRONRCEaminatin 1 point)

Given the following initial conditions on Unit 1:

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 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months 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 required actions 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 4 AND Declare the IA NV pump INOPERABLE by 0400 on February 2.

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

Page 52 of 100

McGuire Nuclear Station Question: 53 2OIIMNSSRONRCExámination 1 point)

Given the following initial conditions on Unit 1:

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 a MAXIMUM of (1) minutes to prevent (2)

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

A. 1. 40 minutes

2. failure of components supplied by the battery B. 1. 60 minutes

2. failure of components supplied by the battery C. 1. 40 minutes

2. complete loss of the vital battery D. 1. 60 minutes

2. complete loss of the vital battery Page 53 of 100

McGuire Nuclear Station Question: 54 2011 MNSRO NRC Examination 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 Question: 55 2011 MNS SRO CExaiintioi I point)

Given the following conditions on Unit 1:

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 the reason for those actions?

A. The RB Train is aligned to the SNSWP To prevent pump runout due to high flow conditions B. The B Train is aligned to the SNSWP To prevent cavitation due having all 4 RN pumps suction aligned to LLI C. The A Train is aligned to the SNSWP To prevent pump runout due to high flow conditions D. The A Train is aligned to the SNSWP To prevent cavitation due having all 4 RN pumps suction aligned to LLI Page 55 of 100

McGuire Nuclear Station Question: 56 2011 MNS SRONRCExaminatlon I point)

Given the following condition associated with Unit 1:

SIGs 1A & lB are faulted and indicating <5% WR Level

SIGs 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 describes the NEXT procedure to be implemented and action(s) required?

A. Go to EP-E2 (Faulted SIG Isolation)

Isolate C & D S/Gs 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/Cs D. Go to EP-E2 (Faulted S/C Isolation)

Close all MSIVs and MSIV bypasses Page 56 of 100

McGuire Nuclear Station Question: 57 ZO1IMNS SEQ NRCExaminatlon 1 point)

Given the following conditions on Unit 1:

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°F

2. decrease B. 1. 1.5°F

2. decrease C. 1. 1°F

2. increase D. 1. 1.5°F

2. increase Page 57 of 100

McGuire Nuclear Station Question: 58 2011 MNS SRO NRC Examination 1 point)

If an Emergency Boration flowpath via I NV-265B (Boric Acid to NV Pumps) can NOT be established:

AP-38 directs the Operators to (1)

FR-S.1 directs (2)

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

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 (Boric Acid 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 the FWST.

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

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

Page 58 of 100

McGuire Nuclear Station Question:

59 2OHMNSSRONRCExamIntIbn 1 point)

Given the following plant conditions:

A Ventilation Unit Condensate Drain Tank (VUCDT) release to the RC Discharge 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 RC CNTRL) close.

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

Page 59 of 100

McGuire Nuclear Station Question: 60 2011 MNS SRO NRC Examination point)

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 for that action?

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

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

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

D. Place cation bed demineralizer in service The cation bed will remove lithium causing a pH change that prevents further fuel degradation.

Page 60 of 100

McGu ire Nuclear Station Question: 61 2OIIMNSSRONRCExáminatiwi 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

NCPs B and C are running Which ONE (1) of the following RVLIS DIP conditions would indicate Degraded Core Cooling conditions exist?

REFERENCE PROVIDED A TRAINA=18%

TRAIN B = 27%

B TRAINA=25%

TRAIN B = 27%

C. TRAINA=23%

TRAIN B = Failed D. TRAIN A = Failed TRAIN B = 21%

Page 61 of 100

McGuire Nuclear Station Question: 62 2011 MNS SRO NRCExaiiiihation 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 or Safety Injection).

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

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

A. 1. less

2. initiate an NC system cooldown B. 1. more

2. initiate an NC system cooldown C. 1. less

2. placing Auxiliary Spray in service D. 1. more

2. placing Auxiliary Spray in service Page 62 of 100

McGu ire Nuclear Station Question: 63 2OlIMNSSROiVRCExamination 1 point)

Unit 1 was in MODE 3 when a steam generator over pressure event occurred. Given the 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 taken per 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 reduce pressure in the lB SIC C. Manually isolate feedwater to the 1 B SIG to prevent additional feedwater from further pressurizing the lB SIG D. Dump steam from the 1A, 1C and 1D SIGs to reduce NC system temperature and reduce pressure in the lB SIC Page 63 of 100

McGuire Nuclear Station Question: 64 2011 MNSSRO NRC Examinaiidn 1 point)

Given the following conditions on Unit 2:

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 operations of the NS system?

A. 2B NS Pump can NOT be started.

B. ONLY 2NS-IB (Train B Containment Sump Isolation Valve) can be opened.

C. ONLY 2Nl-1 84B (Train B Containment Sump Isolation Valve) can be opened open.

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

Page 64 of 100

McGuire Nuclear Station Question:

65 2011 MNS SRO NRCExaminatibn 1 point)

Given the following:

A Unit 1 RO is performing PT111A146001003 A (Semi-Daily Surveillance Items Checklist)

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

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

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

REFERENCE PROVIDED A. 1. M0deIONLY

2. 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 NC system leak within one hour.

B. 1. M0deIONLY

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

C. 1. Modes 1-4

2. 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 NC system leak within one hour.

D. 1. Modes 1-4

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

Page 65 of 100

McGuire Nuclear Station Question: 66 2011 MNSSRONRCExamination i point)

Given the following on Unit 2:

Maintenance is performing PLANNED work on equipment in the Unit 2 Turbine Building

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

The alarm was not specifically discussed prior to the work starting and this is the 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 the REQUIRED 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 the C/R Supervisor.

The alarm response is REQUIRED to be reviewed.

D. Announce the noun name of the alarm which shall be repeated back by the C/R Supervisor.

The alarm response is NOT REQUIRED to be reviewed.

Page 66 of 100

McGuire Nuclear Station Question: 67 2(JII MNS SR 0 Ni? C Exminatibn 1 point)

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

1) Less than or equal to pCi/gm DEl

2) Less than or equal to pCi/gm GSA Which ONE (1) of the following completes the statements above?

A. 1. 0.1

2. 100/E-bar B. 1. 100/E-bar

2. 0.1 C. 1. 1.0

2. 100/E-bar D. 1. 100/E-bar

2. 1.0 Page 67 of 100

McGuire Nuclear Station Question: 68 2011 MNS SRO NRC Examination 1 point)

Given the following conditions on Unit 1:

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, with no concurrent responsibilities.

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

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

A. 1. one active licensed RO

2. allowed B. 1. one active licensed RO

2. NOT allowed C. 1. two active licensed ROs

2. allowed D. 1. twoactive licensed ROs

2. NOT allowed Page 68 of 100

McGuire Nuclear Station Question:

69 2011 MITS SRO NRCExamination 1 point)

In accordance with NSD 700 (Verification Techniques):

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

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

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

A. 1. no Red or White tags ONLY

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

2. remove locking device and physically verify valve position then replace the locking device C. 1. no Red or White tags ONLY

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

2. ensure the locking device is properly installed and locked ONLY Page 69 of 100

McGuire Nuclear Station Question:

70 2011 MNS SRU NRC Examination 1 point)

Given the following:

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) within aMAXIMUM0f (2)

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

A. 1. 1A Main Transformer

2. 8min45sec B. 1. lAMainTransformer

2. 28min45sec C. 1. 2A Main Transformer

2. 8min45sec D. 1. 2A Main Transformer

2. 28min45sec Page 70 of 100

McGu ire Nuclear Station Question:

71 2UlIMNSSRONRCExamination 1 point)

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)

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

A. 1. not clear until the ED is reset

2. clear when the dose rate drops below the alarm setpoint B. 1. not clear until the ED is reset

2. not clear until the ED is reset C. 1. automatically clear after 10 seconds

2. clear when the dose rate drops below the alarm setpoint D. 1. automatically clear after 10 seconds

2. not clear until the ED is reset Page 71 of 100

McGu ire Nuclear Station Question:

72 2OliMNSSRONRCExáinination 1 point)

Given the following conditions:

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

The Radiation Work Permit (RWP) specifies that the individual workers must not exceed their annual ALERT exposure limits Based on the conditions above, the individual workers exposure must remain LESS THAN a MAXIMUM of (1) of the annual (2)

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

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 Question:

73 2OlIMNSSRONRCExamination 1 point)

Given the following conditions on Unit 1:

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 which would require entry into the RED path procedure FR-C.1 (Response to Inadequate Core 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

McGuire Nuclear Station Question:

74 2011 MNS SRO NRC Examination 1 point)

The following conditions on Unit 1:

Due to a fire in the Auxiliary Building control of the plant has been transferred to the 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 should be 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 fill

2. NC Loop WIR Pressure and Steam Generator Pressure B. 1. Allow the PZR to fill

2. Core Exit Thermocouples and NC Loop WIR Pressure C. 1. Open the Rx Head vents to lower PZR level

2. NC Loop WIR Pressure and Steam Generator Pressure D. 1. Open the Rx Head vents to lower PZR level

2. Core Exit Thermocouples and NC Loop WIR Pressure Page 74 of 100

McGuire Nuclear Station Question:

75 2011 MNSSRO NRCExainination 1 point)

Given the following conditions on Unit 1:

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 into AP-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 Question: 76 21111 MNS SROITRC Examination I point)

Given the following conditions on Unit 1:

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 are allowed to depressurize to less than 190 PSIG?

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

A. 1. Nitrogen injection from the CLAs

2. 5 minutes B. 1. Inadvertent criticality

2. 5 minutes C. I. Nitrogen injection from the CLAs

2. 10 minutes D. 1. Inadvertent criticality

2. 10 minutes Page 76 of 100

McGu ire Nuclear Station Question:

77 2011 MNS SRO NRCExamination 1 point)

Given the following conditions:

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 EAL Number 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 Question:

78 2011 MNS SRO NRC ExamInatIon 1 point)

Given the following initial conditions on Unit 1:

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)

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

A. 1.4

2. 24%

B. 1.8

2. 24%

C. 1.4

2. 36%

D. 1.8

2. 36%

Page 78 of 100

McGuire Nuclear Station Question: 79 2U11MNSgRONRCExainihatioi I point)

Given the following conditions:

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 DG AND SI loading response if an automatic SI actuation occurs before any corrective actions are taken?

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

B. The I B DG is OPERABLE AND Train A SI loads will start automatically, but Train B SI loads must be started manually.

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

D. The I B DG is INOPERABLE AND Train A SI loads will start automatically, but Train B SI loads must be started manually.

Page 79 of 100

McGuire Nuclear Station Question: 80 2011 MNS SRO NRC Examination 1 point)

Given the following conditions:

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

Shortly after the release was initiated, I EMF-44 (Ventilation Unit Condensate Drain 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 describes additional actions required for this LWR (Liquid Waste Release) in accordance with 0P111A165001001 A (Ventilation Unit Condensate Drain Tank Operation)?

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

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

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

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

Page 80 of 100

McGuire Nuclear Station Question: 81 2011 MNS SRO NRC Examination point)

Given the following conditions on Unit 1:

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 to mitigate the loss of level in the Spent Fuel Pool in accordance with (AP-40 Loss of Refueling 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 (Fuel Transfer Tube Block valve).

Page 81 of 100

McGuire Nuclear Station Question:

82 2011 MNS SRO NRC Examination 1 point)

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

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

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

A. 1. 35R/hr

2. Start the Containment Auxiliary Charcoal Filter Unit.

B. 1. 15R/hr

2. Start the Containment Auxiliary Charcoal Filter Unit.

C. 1. 35RIhr

2. Ensure the VE system is in service and purge containment to the annulus D. 1. l5RIhr

2. Ensure the VE system is in service and purge containment to the annulus Page 82 of 100

McGu ire Nuclear Station Question: 83 2011 MNS SRO NRC Examination I point)

Given the following:

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 affected AND what MINIMUM actions are required per SLC 16.9.3 (Halon Systems)?

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

B. 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

McGuire Nuclear Station Question: 84 2011 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

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 for those procedure actions?

A. 1. 1TA and 1TC auto-swap.

2. Restart 1C NC pump within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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

McGuire Nuclear Station Question: 85 2011 MNS SRO NRC Examination 1 point)

Given the following plant conditions:

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 the LIMITING 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 Steam Line Break.

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

Page 85 of 100

McGuire Nuclear Station Question:

87 2OllMNSSRONRCExamination 1 point)

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 is required on CA flow rates?

A. SIGC ONLY No limitation on CA flow rate B. SIGsAORS/GC No limitation on CA flow rate C. SIGC ONLY 100 gpm limitation on CA flow rate D. S/GsCANDS/GD 100 gpm limitation on CA flow rate Page 87 of 100

McGuire Nuclear Station Question:

88 2O1XMNSSRONRCExaminaün 1 point)

Given the following conditions on Unit 1:

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

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

McGuire Nuclear Station Question:

89 2011 MNS SRO NRC Examination (1 point)

Given the following plant conditions:

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 the ruptured S/C in this situation AND the procedural guidance regarding WHEN the ruptured S/G will be isolated?

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

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

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

Page 89 of 100

McGuire Nuclear Station Question: 90 2011 iWNS SEQ NRC Examination 1 point)

Given the following:

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 reduced to less than a MAXIMUM of (1) power within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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

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

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 Question:

91 2011 MNS SRO NRC Examination 1 point)

Given the following conditions on Unit 1:

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 of PT/i /A/4600/003 C (TR 16.9.14.3) is (2)

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

REFERENCE PROVIDED A. 1. 0.95

2. 8.7%

B. 1. 0.95

2. 13.6%

C. 1. 0.98

2. 8.7%

D. 1. 0.98

2. 13.6%

Page 91 of 100

McGuire Nuclear Station Question: 92 2OlJMNSSRONRCExamination 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 conditions specified is (1)

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

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

A. 1. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

2. will B. 1. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

2. will not C. 1. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

2. will D. 1. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

2. will not Page 92 of 100

McGu ire Nuclear Station Question: 93 21111 MNS SRO NRC Examination 1 point)

Given the following plant conditions:

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 on both units In accordance with the AP-45 Background document, these valves must be closed within (1) minutes of thestart of the fire.

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

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

A. 1. 10

2. INOPERABLE B. 1.30

2. INOPERABLE C. 1. 10

2. OPERABLE D. 1.30

2. OPERABLE Page 93 of 100

McGu ire Nuclear Station Question:

94 2011 MNS SRO NRC Examination (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 and ES-0.1 is coordinated?

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

B. Suspend actions in ES-0.1 until AP-lO is complete; THEN Return 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 Question:

95 2011 MNSSRO NRC Exüminatton 1 point)

Given the following conditions on Unit 1:

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 is allowed to approve bypassing the interlock?

1. Fuel Handling SRO

2. Reactor Engineering

3. Operations Shift Manager A. IONLY B. 1 AND 2 ONLY C. lAND 3ONLY D. 1,2,AND3 Page 95 of 100

McGuire Nuclear Station Question: 96 2011 MNS SRO NRC Examination (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 emergent maintenance activities is performed ONLY by the (1)

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

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

A. 1.WCCSRO

2. are B. 1. CRS

2. are C. 1.WCCSRO

2. are NOT D. 1. CRS

2. are NOT Page 96 of 100

McGuire Nuclear Station Question:

97 2011 MNS SRONRCExamination (1 point)

Given the following conditions on Unit 2:

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

The Risk Management Plan does NOT meet the exemption requirements of NSD 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 risk condition where a Risk Management Plan must be generated is (1)

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

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

A. 1. RED

2. Site Vice-President B. 1. ORANGE

2. Site Vice-President C. 1. RED

2. Plant Operations Review Committee D. 1. ORANGE

2. Plant Operations Review Committee Page 97 of 100

McGuire Nuclear Station Question: 98 2011 MNS SRO NRC Examination 1 point)

Given the following conditions:

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 for Emergency Exposure) the emergency exposure is required to be APPROVED by the (2)

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

A. 1.30

2. Radiation Protection Manager B. 1.30

2. Emergency Coordinator C. 1.75

2. Radiation Protection Manager D. 1.75

2. Emergency Coordinator Page 98 of 100

McGuire Nuclear Station Question:

99 2011 MNS MW NRCExamination (1 point)

Given the following conditions on Unit 1:

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 the Emergency Procedures after the reactor trip?

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

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

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

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

Page 99 of 100

McGu ire Nuclear Station Question: 100 2U11 MNS SRO NRC ExaMination (1 point)

Given the following conditions on Unit 1:

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

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

1) In accordance with the Emergency Response Guidelines (ERG5), which ONE (1) of the following describes the basis for why ECA-1.1 has priority over FR-Z.1 regarding 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, no actions in any other procedures should be performed until the ECA is complete.

2. Containment pressure exceeds a MINIMUM of 10 PSIG.

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

2. Containment pressure exceeds a MINIMUM of 10 PSIG.

C. 1. Per ERG rules of usage, once an ECA procedure is implemented, no actions in any other procedures should be performed until the ECA is complete.

2. Containment pressure exceeds a MINIMUM of 15 PSIG.

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

2. Containment pressure exceeds a MINIMUM of 15 PSIG.

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 of 1 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 MI4S t.OSS OF ALL AG POWER RECOVERY WITHOUT I1 PAGE NO:*

REQUIRED EPIIIAI5000IECA-O.1 Enclosure 4 Page 1 of 1 I 31 of 34 i

UNIT 1 Natural Circulation Parameters Rev. 10 -

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

NC subcooling - GREATER THAN 0°F a 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 to establish Natural Circulation fl-ow.

1200 1100 S

G 1000 p

R 900 E

S 800 S

U R 700 F

p 600 S

I 500 G

400 300 200 1 QO 0

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

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

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

b. Two diesel generators (DGs) capable of supplying the Onsite Essential Auxiliary Power Systems; AND The automatic load sequencers for Train A and Train B shall be OPERABLE.

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

ACTIONS NOTE 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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. OPERABLE offsite circuit.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no available inoperable when offsite power to one its redundant required train concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

AND (continued)

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

AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore offsite circuit to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND 6 days from discovery of failure to meet LCO B. One DG inoperable. B.1 Perform SR 3.8.1.1 for the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months offsite circuit(s).

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

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

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

AND (continued)

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

AC SourcesOperating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.4 Restore DG to OPERABLE 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

status.

AND 6daysfrom discovery of failure to meet LCO

C. Two offsite circuits C.1 Declare required feature(s) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from inoperable, inoperable when its discovery of redundant required Condition C feature(s) is inoperable, concurrent with inoperability of redundant required feature(s)

AND C.2 Restore one offsite circuit 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to OPERABLE status.

(continued)

For Unit I only, the Completion Time that the IA EDG can be inoperable as specified by Required Action B.4 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 6 days from discovery of failure to meet the LCO up to a total of 10 days as part of the IA EDG Jacket/Intercooler Water Pump Motor repair. Upon completion of the repair and restoration, this footnote is no longer applicable and will expire at 1741 hours0.0202 days 0.484 hours 0.00288 weeks 6.624505e-4 months on June 15, 2007.

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

- A Sourcss - Operat 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Oneoffsitecircuit -NOTE-inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, AND Distribution Systems Operating, when Condition D is One DG inoperable, entered with no AC power source to any train.

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

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

E. Two DGs inoperable. E.1 Restore one DG to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months OPERABLE status.

F. One automatic load F.1 Restore automatic load 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sequencer inoperable, sequencer to OPERABLE status.

G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, AND C, D, E, or F not met.

G.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months H. Three or more AC H.1 Enter LCO 3.0.3. Immediately sources inoperable.

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

- AG Sources Operating 3.8.1 SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power 7 days availability for each offsite circuit.

SR 3.8.1.2 NOTES

1. Performance of SR 3.8.1.7 satisfies this SR.

2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.

3. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer.

When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.

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

(continued)

McGuire Units I and 2 3.8.1-5 Amendment Nos. 184/166

- AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.3 NOTES

1. DG loadings may include gradua[ loading as recommended by the manufacturer.

2. Momentary transients outside the load range do not invalidate this test.

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

4. This SR shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.7.

Verify each DG is synchronized and loaded and operates 31 days for 60 minutes at a load 3600 kW and 4000 kW.

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 days tank.

SR 3.8.1.6 Verify the fuel oil transfer system operates to 31 days automatically transfer fuel oil from storage tank to the day tank.

(continued)

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

AC Sources Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.7 NOTES All DG starts may be preceded by an engine prelube period.

Verify each DG starts from standby condition and 184 days achieves in 11 seconds voltage of 3740 V and frequency of 57 Hz and maintains steady state voltage 3740 V and 4580 V, and frequency 58.8 Hz and 61.2 Hz.

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

Verify automatic and manual transfer of AC power 18 months sources from the normal offsite circuit to each alternate offsite circuit.

(continued)

McGuire Units I and 2 3.8.1-7 Amendment Nos. 184/166

AC Seurce = Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.9 Verify each DG, when connected to its bus in parallel with 18 months offsite power and operating with maximum kVAR loading that offsite power conditions permit, rejects a load greater than 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, the voltage is 3740 V and 4580 V; and

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

SR 3.8.1.10 Verify each DG does not trip and voltage is maintained 18 months 4784 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 Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.11 NOTES

1. All DG starts may be preceded by an engine prelube period.

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

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

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 in 11 seconds,

2. energizes auto-connected blackout loads through automatic load sequencer,

3. maintains steady state voltage 3740 V and 4580 V,

4. maintains steady state frequency 58.8 Hz and 61.2 Hz, and

5. supplies auto-connected blackout loads for 5 minutes.

(continued)

McGuire Units I and 2 3.8.1-9 Amendment Nos. 184/166

ACSouroes Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR3.8.1.12 . .

All DG starts may be preceded by prelube period.

Verify on an actual or simulated Engineered Safety 18 months Feature (ESF) actuation signal each DG auto-starts from standby condition and:

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

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

c. Operates for 5 minutes; and

d. The emergency bus remains energized from the offsite power system.

(continued)

McGuire Units I and 2 3.8.1-10 Amendment Nos. 184/166

AC Sources Operating-3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.13 Verify each DGs non-emergency automatic trips are 18 months bypassed on actual or simulated loss of voltage signal on the emergency bus concurrent with an actual or simulated ESF actuation signal.

SR 3.8.1.14 NOTES

1. Momentary transients outside the load range do not invalidate this test.

2. DG loadings may include gradual loading as recommended by the manufacturer.

Verify each DG, when connected to its bus in parallel with 18 months offsite power and operating with maximum kVAR loading that offsite power conditions permit, operates for 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

a. For 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded 4200 kW and 4400 kW; and

b. For the remaining hours of the test loaded 3600 kW and 4000 kW.

(continued)

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

AC Sources Opersting 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.15 NOTES

1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded 3600 kW and 4000 kW.

Momentary transients outside of load range do not invalidate this test.

2. All DG starts may be preceded by an engine prelube period.

Verify each DG starts and achieves, in 11 seconds, 18 months voltage 3740 V, and frequency 57 Hz and maintains steady state voltage 3740 V and 4580 V and frequency 58.8 Hz and 61.2 Hz.

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

Verify each DG: 18 months

a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;

b. Transfers loads to offsite power source; and

c. Returns to standby operation.

(continued)

McGuire Units I and 2 3.8.1-12 Amendment Nos. 184/166

AC Sourcs=Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.17 NOTES This Surveillance shall not be performed in MODE 1, 2, 3, or4.

Verify, with a DG operating in test mode and connected 18 months to its bus, an actual or simulated ESF actuation signal overrides the test mode by:

a. Returning DG to standby operation; and

b. Automatically energizing the emergency load from offsite power.

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

(continued)

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

AC.Srces Operatin 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.19 -NOTES

1. All DG starts may be preceded by an engine prelube period.

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

Verify on an actual or simulated loss of offsite power 18 months signal in conjunction with an actual or simulated ESF actuation 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 11 seconds,

2. energizes auto-connected emergency loads through load sequencer,

3. achieves steady state voltage 3740 V and 4580 V,

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

5. supplies auto-connected emergency loads for 5 minutes.

(continued)

McGuire Units I and 2 3.8.1-14 Amendment Nos. 184/1 66

- AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.20 NOTES All DG starts may be preceded by an engine prelube period.

Verify when started simultaneously from standby 10 years condition, each DG achieves, in 11 seconds, voltage of 3740 V and frequency of 57 Hz and maintains steady state voltage 3740 V and 4580 V, and frequency 58.8 Hz and 61.2 Hz.

McGuire Units I and 2 3.8.1-15 Amendment Nos. 184/1 66

MNS RESPONSE TO DEGRADED CORE COOLING PAGE NO.

EPI1IAI5000IFR-C.2 10 of 44 Rev. 7 UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

7. Check Reactor Vessel level as follows:

a. Determine required reactor vessel delta P 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 DELTA P. 1) if REACTOR VESSEL DIP going up, THEN observe Note prior to Step 1 and RETURN ] Step 1.

_2) TQStep8.

c. RETURN ] procedure and step in effect.

8. Check if one NC pump should be stopped:

a. All NC pumps ON.

- a. ]Q.Step 10.

b. Stop lB NC pump.

c. Place 1 NC-29C (B NC Loop PZR Spray Control) 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 be OPERABLE:

a. The containment floor and equipment sump level monitors and the incore 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 containment atmosphere particulate radioactivity monitor.

ACTIONS NOTE Separate Condition entry is allowed for each leakage detection instrument.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or both containment A.1 floor and equipment - NOTE -

sump level monitor(s) Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable, after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.2 Restore inoperable 30 days containment floor and equipment sump level monitor(s) to OPERABLE status.

(continued)

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

RCS Leakage Detection instrumentation 3t5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Containment B.1 atmosphere particulate - NOTE -

radioactivity monitor Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable, after establishment of steady state operation Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR B.2 Analyze grab samples of Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the containment atmosphere.

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

tank level monitor Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable, after establishment of steady state operation Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR C. 1.2 Analyze grab samples of Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

the containment atmosphere.

OR C.1.3 Perform SR 3.4.15.1. Once per8 hours.

AND C.2 During Modes 2, 3, and 4, 30 days restore inoperable containment ventilation unit condensate drain tank level monitor to OPERABLE status.

(continued)

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

RCS Leakage Detection instrumentation 3.4:15 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Containment D.1 Restore containment 30 days atmosphere particulate atmosphere particulate radioactivity monitor radioactivity monitor to inoperable in MODE 1. OPERABLE status.

AND OR Containment ventilation D.2 Restore containment unit condensate drain ventilation unit condensate 30 days tank level monitor drain tank level monitor to inoperable in MODE 1. OPERABLE status.

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

Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND F.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months G. All required monitors G.1 Enter LCO 3.0.3. Immediately inoperable.

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

RCS Leakage Detection instrumentation

- 3.4t5 SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the containment 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months atmosphere particulate radioactivity monitor.

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

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

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

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

SR 3.4.15.6 Perform CHANNEL CALIBRATION of the incore 18 months instrument 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.

MC0048M3 Reference Use PERFORMANCE UNCONTROLLED FORPRINT (ISSUED) PDF Format

RP/OIA/5700J00U Page 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 degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

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

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

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

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

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

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

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

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

RP/OIAJ57001IQ0 Page 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 affect fission product barriers, THEN review the listing of enclosures to determine if the event is applicable to one of the categories shown.

2.4 Compare actual plant conditions to the Emergency Action Levels evaluated in 2.2 and/or 2.3, 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 classification and continue with subsequent steps of this procedure.

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

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

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

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

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

En .re 4.1 RP/O/A/57U. O Fission Product Barrier Matrix Page 1 of5 Use EALs to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all 3 barriers. Classif,r according to the table on page 2 of 5 of this enclosure.

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

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

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

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

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

Elk re4.1 i/O!A/57. 0 Fission Product Barrier Matrix Page 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 is included in the matrix as a place holder only, and has no point value assigned.

Barrier Points (1-5) Potential Loss (X) Loss (X) Total Points Classification Containment 1 3 1 3 Unusual Event NCS 4 5 46 Alert Fuel Clad 4 5 710 SiteArea Emergency Total Points 11 - 13 General Emergency

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

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

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

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

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

6. In the table on page 1 of this enclosure, under one of the four classification columns, select the event (e.g. 4.l.A.l for Loss of Nuclear Coolant System) 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 for Emergency Coordinator/EOF Director approval and transmittal.

EL Are 4.1 1u!O/A157u..

Fission Product Barrier Matrix Page3of5 4.1.C CONTAINMENT BARRIER 4.1.N NCS BARRIER 4.1.F FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

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

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

Containment-RED. Not applicable.

NCS Integrity-

Not applicable.

Core Cooling-

Core Cooling-RED.

RED. ORANGE.

Core Cooling -

RED Path is

Heat Sink-RED.

Heat Sink-RED.

indicated for >15 minutes.

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

Rapid unexplained

Unisolable leak

GREATER THAN

Not applicable.

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

H2 concentration initial increase. the normal of NCS subcooling. (DEl) 1-13 1.

>9%. charging mode

Containment with letdown

Containment pressure or sump isolated.

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

and a VX-CARF operating.

CONTINUED CONTINUED CONTINUED

En .re 4.1 RP/OIA/57(. .jO Fission Product Barrier Matrix Page 4 of 5 4.1.C CONTAINMENT BARRIER 4.1.N NCS BARRIER 4.1.F FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

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

3. Containment Isolation Valves Status After 3. SG Tube Rupture 3. Containment Radiation Monitoring Containment Isolation Actuation Not applicable. Containnient

Primary-to-

Indication that a Not applicable.

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

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

4. SG Secondary Side Release With Primary-to- 4. Containment Radiation Monitoriiw 4. Emcrency Coordinator/EOF Director Secondary Leakage Judement

- Not applicable. Release of Not applicable.

Not applicable.

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

GREATER THAN Tech Spec allowable. END CONTINUED CONTINUED

En re4.1 RP/O/A/57U O Fission Product Barrier Matrix Page 5 of5 4.1.C CONTAINMENT BARRIER 4.1.N NCS BARRIER 4.1.F FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

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

5. Significant Radioactive Inventory In Emergency CoordinatorlEOF Director Containment Judement Containment Rad.

Not applicable. Any condition, including inability to monitor Monitor EMF5 1A the barrier, that in the opinion of the or 51B Emergency Coordinator/EOF Director Reading @ time indicates LOSS or POTENTIAL LOSS of since shutdown: the NCS barrier.

>39ORJhr@

o 0.5 hr

> l70RJhr(

0.5 2 hr

- END

> 125 RJhr @

2 4 hr

> 9ORJhr @

4 8 hr

> 53 RJhr @

> 8 hr.

6. Emergency Coordinator /EOF Director Judgement

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

END

Enclosure 4.2 RPIO/A!57001000 System Malfunctions Page 1 of2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.2.U.1 Inability to Reach Required 4.2.A.1 Unplanned Loss of Most or All 4.2.S.1 Inability to Monitor a Shutdown Within Technical Safety System Annunciation or Significant Transient in Specification Limits. Indication in Control Room Progress.

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

Compensatory Non-Alarming OPERATING MODE: 1,2,3,4 4.2.U.1-1 Plant is brought to required Indicators Unavailable.

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

Time.

4.2.A.1-1 The following conditions exist: Loss of most (>50%)

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

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

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

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

Director, the loss of the annunciators A significant plant transient is or indicators requires additional in progress.

subcriticality personnel (beyond normal shift

core cooling compliment) to safely operate the unit. OR

heat sink

containment.

CONTINUED

Loss of the OAC.

END END

Enclosure 4.2

/OIAI57oo/ooo System 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 the Technical Specification allowable limit. (*mode 3 with Tavg 5 00°F) 4.2.U.4 Reactor Coolant System (NCS)

Leakage.

OPERATING MODE: 1,2,3,4 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 or Offsite Communications.

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

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

END

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

(procedure calculations) cannot be period indicated in the EAL OPERATING MODE: ALL completed within this time period, AND the required assessments OPERATING MODE: ALL declaration must be made based on the (procedure calculations) cannot Note 1: These EMF readings are valid radiation monitor reading. be completed within this time calculated based on Note 1: These EMF reading are period, declaration must be made average annual calculated based on 4.3.U.1-1 A valid indication on radiation monitor based on the valid radiation meteorology, site average annual EMF- 49L, EMF.44L or EMF-3 1 monitor reading. boundary dose rate, and meteorology, site (when aligned to RC) of design unit vent flow rate. boundary dose rate, find 5.45E+06 cpm for 60 minutes or will 4.3.A.1-1 A valid indication on Calculations by the dose design unit vent flow likely continue for 60 minutes, which radiation monitor EMF- 49H assessment team use rate. Calculations by the indicates that the release may have of1. 56E+03 cpmfor actual meteorology, dose assessment teaiji use exceeded the initiating condition and 15 minutes or will likely release duration, and unit actual meteorology, indicates the need to assess the release continue for 15 minutes, which vent flow rate. Therefore, release duration, and unit with procedure HP/0/B/1 009/010, indicates that the release may these EMF readings vent flow rate.

HP/0/B/1 009/029, or SH/0/B/2005/00 1. have exceeded the initiating should not be used if dose Therefore, these EMIF condition and indicates the need assessment team readings should not be to assess the release with calculations are available. used if dose assessment (Continued) procedure HP/0/B/1 009/010, team calculations are HP/0/B/1009/029, or (Continued) available.

SHJO/B/2005/00 1.

(Continued)

Enclosure 4.3 Ip/O!A!57oo/ooo Abnormal Rad Levels/Radiological Effluent Page 2 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.U.1-2 A valid indication on radiation monitor 4.3.A.1-2 A valid indication on Note 2: If dose assessment team Note 2: If dose assessment tam EMF- 36L of 2.05E+04 cpm for radiation monitor EMF- 36L calculations cannot be calculations cannot le 60 minutes or will likely continue for of 2.05E+06 cpm for completed in 15 minutes, completed in 15 minutes, 60 minutes, which indicates that the > 15 minutes or will likely then valid monitor reading then valid monitor release may have exceeded the initiating continue for 15 minutes, should be used for reading should be used condition and indicates the need to assess which indicates that the release emergency classification. for emergency the release with procedure may have exceeded the initiating classification.

HP/0/B/1009/0 10, HP/0/B/l009/029, or condition and indicates the need 4.3.S.1-1 A valid indication SH/0/B/2005/00 1. to assess the release with on radiation monitor 4.3.G.1-1 A valid indication procedure HP/0/B/1 009/010, EMF-36H of on radiation monitor 4.3.U.l-3 A valid indication on radiation monitor HP/0/B/1009/029, or 3.4E+03cpm EMF-36H of EMF-3 1 (when aligned to WC or SHJO/B/2005/00 1. sustained for 3.4 E + 04 cpm WWCB) of 9.174 E+03 cpm for > 15 minutes. sustained for 60 minutes or will likely continue for 4.3.A.1-3 Gaseous effluent being released >15 minutes.

60 minutes which indicates that the exceeds 200 times the level of 4.3.S.1-2 Dose assessment team release may have exceeded the initiating SLC l6.ll-6for>l5minutesas calculations indicate 4.3.G.1-2 Dose assessment condition and indicates the need to assess determined by Radiation dose consequences team calculations the release with procedure Protection (RP) procedure. greater than 100 indicate dose HP/0/B/1 009/010, HP/0/B/1 009/029, or mRem TEDE or 500 consequences SH/0/B/2005/00 1. 4.3.A.1-4 Liquid effluent being released mRem CDE Adult greater than 1000 exceeds 200 times the level of Thyroid at the site mRem TEDE or 4.3.U.1-4 Gaseous effluent being released exceeds SLC 16.11-1 for> 15 minutes as boundary. 5000 mRem CDE two times SLC 16.1 1-6 for determined by Radiation Adult Thyroid at the 60 minutes as determined by Radiation Protection (RP) procedure. 4.3.S.1-3 Analysis of field site boundary.

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

END

Enclosure 4.3 R1:/O/A/57oo/ooo Abnormal Rad Levels/Radiological Effluent Page3of5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.U.2 Unexpected Increase in Plant 4.3.A.2 Major Damage to Radiation or Airborne Concentration. Irradiated Fuel or Loss of Water Level that Has or OPERATING MODE: ALL Will Result in the Uncovering of Irradiated 4.3 .U .2-1 Indication of uncontrolled water level Fuel Outside the Reactor decrease of greater than 6 inches in the Vessel.

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

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

4.3.U.2-3 Unplanned valid area EMF reading exceeds the levels shown in Enclosure Spent Fuel Building 4.10. Refueling Bridge 1EMF-17 2EMF-4 END Spent Fuel Pool Ventilation 1 EMF42 2EMF-42 Reactor Building Refueling Bridge lEMFl6*

2EMF-3

Containment Noble Gas 1EMF39*

2EMF39*

Applies to Mode 6 and No Mode Only.

(Continued)

Enclosure 4.3 RP/O/A/5700/000 Abnormal Rad Levels/Radiological Effluent Page 4 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.2-2 Plant personnel report that water level drop in reactor refueling cavity, spent fuel pool, or fuel transfer canal has or will exceed makeup capacity such that any irradiated fuel will become uncovered.

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

<358 inches after initiation of NC system make-up.

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

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

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

(Continued)

Enclosure 4.3 /O/A/s7oo/ooo Abnormal Rad Levels/Radiological Effluent Page 5 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.

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

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

4.3.A.33 Valid area EMF reading exceeds the levels shown in Enclosure 4.10.

END

Enclosure 4.4 Iu!O/A/5700/000 Loss of Shutdown Functions Page 1 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY END 4.4.A.1 Failure of Reactor 4.4.S.1 Failure of Reactor 4.4.G.1 Failure of the Reactor Protection System Protection System Protection System to Instrumentation to Complete Instrumentation to Complete Complete an Automatic Trip or Initiate an Automatic or Initiate an Automatic and Manual Trip {j Reactor Trip Once a Reactor Trip Once a Successful and There is Reactor Protection System Reactor Protection System Indication of an Extreme Setpoint Has Been Exceeded Setpoint Has Been Exceeded Challenge to the Ability to and Manual Trip and Manual Trip WAS NOT Cool the Core.

Successful. Successful.

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

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

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

was not successful. was not successful.

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

decreasing.

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

Core Cooling CSF-RED

Heat Sink CSF-RED.

END

Enclosure 4.4 RP/O/A/5700/000 Loss of Shutdown Functions Page 2 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.A.2 Inability to Maintain Plant 4.4.S.2 Complete Loss of Function in Cold Shutdown. Needed to Achieve or Maintain Hot Shutdown.

OPERATING MODE: 5,6 OPERATING MODE: 1,2,3,4 4.4.A.21 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 the Reactor Vessel That Has or

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

below 200°F OPERATING 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 Vessel Level Indication System (RVLIS) decreasing after initiation of NC system makeup.

(Continued)

Enclosure 4.4 1u/O/A!5700/000 Loss of Shutdown Functions Page 3 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.S.3-2 Failure of heat sink causes loss of cold shutdown conditions.

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

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

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

END

Enclosure 4.5 ip/O/A/57oo/ooo Loss of Power Page 1 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.5.U.1 Loss of All Offsite 4.5.A.1 Loss of All Offsite 4.5.S.1 Loss of All Offsite 4.5.G.1 Prolonged Loss of All Power to Essential Power and Loss of All Power and Loss of All (Offsite and Onsite) AC Busses for Greater Than Onsite AC Power to Onsite AC Power to Power.

15 Minutes. Essential Busses During Essential Busses.

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

indicated by:

Loss of offsite power to 4.5.A.1-1 Loss of all offsite and Loss of power on essential essential buses ETA and onsite AC power as Loss of power on essential buses ETA and ETB for ETB for greater than indicated by: buses ETA and ETB. greater than 15 minutes.

15 minutes.

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

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

within 15 minutes. (Continued)

AND (Continued)

(Continued)

Enclosure 4.5 1u/O/A/5700/000 Loss of Power Page 2 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY OPERATING MODE: 5,6, No 4.5.A.2 AC Power to Essential 4.5.S.2 Loss of All Vital DC At least one of the Mode Busses Reduced to a Power. following conditions Single Power Source for exist:

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

Restoration of at least Loss of offsite power to Additional Single 4.5.S.2-1 The following conditions one essential bus essential buses ETA and Failure Could Result in exist: within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is ETB for greater than Station Blackout. j likely 15 minutes. Loss of both unit related OPERATING MODE: 1,2,3,4 EVDA and EVDD busses

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

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

of:

END SATA SATB ATC ATD DIG A DIG B.

END

Enclosure 4.5 RP/O/A/5700/000 Loss of Power Page 3 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.5.U.2 Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes.

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

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

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

END

Enclosure 4.6 I.P!O/A/57oo/ooo Fire/Explosion and Security Events Page 1 of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.6.U.1 Fire Within Protected Area 4.6.A.1 Fire or Explosion Affecting 4.6.S.1 HOSTILE ACTION within 4.6.G.1 HOSTILE ACTION Boundary NOT the Operability of Plant the PROTECTED AREA Resulting in Loss of Physical Extinguished Within Safety Systems Required to Control of the Facility.

15 Minutes of Detection OR Establish or Maintain Safe OPERATING MODE: ALL Explosion Within the Shutdown. OPERATING MODE: ALL Protected Area Boundary. 4.6.5.1-I A HOSTILE ACTION is OPERATING MODE: 1,2,3, 4, 5, 6 occurring or has occurred within the 4.6.G.1-1 A HOSTILE ACTION has OPERATING MODE: ALL PROTECTED AREA as reported by the occurred such that plant 4.6.A.1-1 The following conditions exist: MiSTS Security Shift Supervision. personnel are unable to operate 4.6.U.1-1 Fire in any of the following (includes non-security events) equipment required to areas NOT extinguished Fire or explosion in any of the maintain safety functions.

within 15 minutes of control following areas: END room notification or

Reactor Building 4.6.G.1-2 A HOSTILE ACTION has verification of a control room

Auxiliary Building caused failure of Spent Fuel fire alarm.

Diesel Generator Rooms Cooling Systems and

Control Room IMMINENT fuel damage is

Reactor Building

Standby Shutdown Facility likely for a freshly off-loaded

Auxiliary Building

CAS reactor core in pool.

Diesel Generator Rooms

SAS Control Room

FWST END

Standby Shutdown Facility

Doghouses (Applies in

CAS Mode 1,2,3,4 only).

SAS Doghouses AND FWST Turbine Building

Service Building (Continued)

Interim Radwaste Building Equipment Staging Building

ISFSL (Continued)

Enclosure 4.6 RP/O/A/5700/000 Fire/Explosion and Security Events Page2of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY One of the following:

4.6.U.l-2 Report by plant personnel of an unanticipated explosion Note: Only one train of a system within the protected area needs to be affected or boundary resulting in visible damaged in order to satisfy damage to permanent this condition.

structures or equipment or a loaded cask in the ISFSI.

Affected safety system parameter indications show 4.6.U.2 Confirmed SECURITY degraded performance CONDITION or Threat

Plant personnel report Which Indicates a Potential visible damage to Degradation in the Level of permanent structures or Safety of the Plant. equipment within the specified area.

OPERATING MODE: All 4.6.A.2 Fire or Explosion Affecting 4.6.U.2-1 A SECURITY CONDITION the Operability of Plant that does NOT involve a Safety Systems Required to HOSTILE ACTION as Establish or Maintain Safe reported by the MNS Security Shutdown.

Shift Supervision.

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

(includes non-security events) 4.6.U.2-3 A validated notification from Fire or explosion in any of the NRC providing information of following areas:

an aircraft threat.

Spent Fuel Pool

Auxiliary Building.

END AND (Continued)

Enclosure 4.6 RP/O!A!5700/000 Fire/Explosion and Security Events Page 3 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY One of the following:

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

Spent Fuel Pool level and/or temperature show degraded performance

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

4.6.A.3 HOSTILE ACTION Within the OWNER CONTROLLED AREA or Airborne Attack Threat.

OPERATING MODE: ALL 4.6.A.3-1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the MNS Security Shift Supervision.

(Continued)

Enclosure 4.6 RP,/O/A/5700/000 Fire/Explosion and Security Events Page 4 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.6.A.3-2 A validated notification from NRC of an airliner attack threat within 30 minutes of the site.

END

Enclosure 4.7 1P/O/A/57oo/ooo Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 1 of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.U.1 Natural and Destructive 4.7.A.1 Natural and Destructive 4.7.S.1 Control Room Evacuation 4.7.G.1 Other Conditions Existing Phenomena Affecting the Phenomena Affecting the Has Been Initiated and Plant Which in the Judgement of Protected Area. Plant Vital Area. Control Cannot Be the Emergency Established. Coordinator/EOF Director OPERATING MODE: ALL OPERATING MODE: ALL Warrant Declaration of OPERATING MODE: ALL General Emergency.

4.7.U.1-1 Tremor felt and valid alarm on 4.7.A.1-1 Valid OBE Exceeded Alarm the Syscom Seismic on 1AD-l3, E-7 4.7.S.1.-1 The following conditions OPERATING MODE: ALL Monitoring System (OAC exist:

M1D2422). 4.7.A.1-2 Tornado or high winds: 4.7.G.1-1 Other conditions exist which Control Room evacuation has in the Judgement of the 4.7.U.1-2 Report by plant personnel of Tornado striking plant been initiated per Emergency tornado striking within structures within the vital AP/l (2)/A!5500!O 17, or Coordinator/EOF Director protected area area: AP/l (2)!A!5500/024. {3] indicate:

boundary/ISFSI.

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

Auxiliary Building substantial core degradation structures or systems within

FWST Control of the plant carmot be with potential for loss of protected area

Diesel Generator Rooms established from the Auxiliary containnent, boundary/ISFSI.

Control Room Shutdown Panel or the

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

resulting in casing penetration

Doghouses (2) potential for or damage to turbine or

CAS uncontrolled radionuclide generator seals.

SAS. (Continued) releases. These releases can reasonably be expected to (Continued) OR exceed Environmental Protection Agency Sustained winds 74 mph for Protective Action Guideline

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

(Continued)

END

Enclosure 4.7 1P/O/A/57oo/ooo Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page2of4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.U.15 Independent Spent Fuel Cask 4.7.A.1-3 Visible structural damage 4.7.S.2 Other Conditions Existing tipped over or dropped greater caused by either: Which in the Judgement of than 12 inches.

Vehicle crashes, OR the Emergency

Turbine failure generated Coordinator/EOF Director 4.7.U.1-6 Uncontrolled flooding in the missiles, OR Warrant Declaration of Site ISFSI area.

Other catastrophic events Area Emergency.

4.7.U.l-7 Tornado generated missile(s) on any of the following plant OPERATING MODE: ALL impacting the ISFSI. structures:

4.7.S.2-1 Other conditions exist which

Reactor Building in the Judgement of the 4.7.U.2 Release of Toxic or

Auxiliary Building Emergency Coordinator/EOF Flammable Gases Deemed

FWST Director indicate actual or Detrimental to Safe

Diesel Generator Rooms likely major failures of plant Operation of the Plant.

Control Room functions needed for

Standby Shutdown protection of the public.

OPERATING MODE: ALL Facility END

Doghouses 4.7.U.2-1 Report or detection of toxic or

CAS flammable gases that could

SAS enter within the site boundary

Ultimate heat sink in amounts that can affect safe (Standby Nuclear Service operation of the plant. Water Pond Dam and Dikes).

4.7.U.2-2 Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event.

(Continued)

Enclosure 4.7 RP/O/A/5700/000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 3 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.U.3 Other Conditions Existing 4.7.A.2 Release of Toxic or Which in the Judgement of Flammable Gases Within a the Emergency Facility Structure Which Coordinator/EOF Director Jeopardizes Operation of Warrant Declaration of an Systems Required to Unusual Event. Maintain Safe Operations or to Establish or Maintain OPERATING MODE: ALL Cold Shutdown.

4.7.U.3-1 Other conditions exist which OPERATING MODE: ALL in the judgernent of the Emergency Coordinator/EOF Note: Structures for the below EALs:

Director indicate a potential

Reactor Building degradation of the level of

Auxiliary Building safety of the plant.

Diesel Generator Rooms

Control Room END

Standby Shutdown Facility

Doghouses

CAS

SAS.

4.7.A.2-1 Report or detection of toxic gases within a Facility Structure in concentrations that will be life threatening to plant personnel.

4.7.A.2-2 Report or detection of flammable gases within a Facility Structure in concentra tions that will affect the safe operation of the plant.

(Continued)

Enclosure 4.7 RP/O/A/5700I000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 4 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.A.3 Control Room Evacuation Has Been Initiated.

OPERATING MODE: ALL 4.7.A.3-l Control Room evacuation has been initiated per AP/l (2)/A!5500/Ol 7, or AP/l (2)/A15500/024. {3}

4.7.A.4 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of an Alert.

OPERATING MODE: ALL 4.7.A.4-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.

END

Enc1oure4.8 RP/O/A/5700/000 Definitions/Acronyms Page 1 of 4 ALERT Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels ALL (As relates to Operating Mode Applicability)

- At all times.

BOMB Refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

CIVIL DISTURBANCE A group of persons violently protesting station operations or activities at the site.

CONFINEMENT BOUNDARY The barrier(s) between areas containing radioactive substances and the environment.

EXPLOSION A rapid, violent unconfined combustion, or a catastrophic failure of pressurizedlenergized equipment that imparts energy of sufficient force to potentially damage permanent structures, 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 or overheated electrical equipment do not constitute fires. Observation of flames is preferred but is NOT required if large quantities of smoke and heat are observed. An electrical breaker flash that creates high temperatures for a short duration and merely localized scorching to that breaker and its compartment should not be considered a fire.

GENERAL EMERGENCY Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile action that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guidelines exposure levels offsite for more than the immediate site area.

HOSTAGE A person(s) held as leverage against the station to ensure demands will be met by the station.

HOSTILE ACTION An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and / or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (i.e., this may include violent acts between individuals in the owner controlled area).

EneIosure4.8

/O/A/57oo/ooo Definitions/Acronyms Page 2 of 4 HOSTILE FORCE One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

IMMINENT Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates that the event or condition will occur. Where IMMINENT time frames are specified, they shall apply.

INABILITY TO DIRECTLY MONITOR Operational Aid Computer data points are unavailable or gauges/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 a specified 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 determined by the judgement of the Emergency Coordinator.

PROTECTED AREA Typically the site specific area which normally encompasses all controlled areas within the security PROTECTED AREA fence.

REACTOR COOLANT SYSTEM (RCS/NCS) LEAKAGE - RCS Operational Leakage as defined in the Technical Specification Basis B 3.4.13.

RUPTURED (As relates to Steam Generator) Existence of primary to secondary leakage of a magnitude sufficient to require or cause a reactor trip and safety injection.

SABOTAGE Deliberate damage, misalignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered with or damaged due to malicious mischief may not meet the definition of SABOTAGE until this determination is made by security supervision.

SECURITY CONDITION Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the 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 electrical load; (3) reactor trip, (4) safety injection, (5) thermal power oscillations 1O%.

Enclosure 4.8 iwOiA/57oo/ooo Definitions/Acronyms Page 3 of 4 SITE AREA EMERGENCY Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile action that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

SITE BOUNDARY That area, including the protected area, in which Duke Power Company has the authority to control all activities, including exclusion or removal of personnel and property.

SLC Selected Licensee Commitments.

SUSTAINED A duration of time long enough to confirm that the CSF is valid (not momentary).

TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE) The sum of external dose exposure to a radioactive plume, to radionuclides deposited on the ground by the plume, and the internal exposure from inhaled radionuclides deposited in the body.

TOXIC GAS A gas that is dangerous to life or health by reason of inhalation or skin contact (e.g.

chlorine).

UNCONTROLLED Event is not the result of planned actions by the plant staff UNPLANNED An event or action is UNPLANNED if it is not the expected result of normal operations, testing, or maintenance. Events that result in corrective or mitigative actions being taken in accordance with abnormal or emergency procedures are UNPLANNED.

UNUSUAL EVENT Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

VALID An indication or report or condition is considered to be VALID when it is conclusively verified by: (1) an instrument channel check, or (2) indications on related or redundant instrumentation, or (3) by direct observation by plant personnel such that doubt related to the instruments operability, the conditions existence or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

VIOLENT Force has been used in an attempt to injure site personnel or damage plant property.

VISIBLE DAMAGE Damage to equipment or structure that is readily observable without measurements, testing, or analyses. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage:

deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering.

Enchure RP/O/A[5700/000 Definitions/Acronyms Page 4 of 4 VITAL AREA Areas within the PROTECTED AREA that house equipment important for nuclear safety. Access to a VITAL AREA is allowed only if an individual has been authorized to be in that area per the Security plan, therefore VITAL AREA is a Security term.

Eneloure

aP/O/Ai5700/000 Emergency Declaration Guidelines Page 1 of 2 THE FOLLOWING GUIDANCE IS TO BE USED BY THE EMERGENCY COORDINATOR iN ASSESSING EMERGENCY CONDITIONS.

The Emergency Coordinator shall review all applicable initiating events to ensure proper classification.

The BASIS Document (located in Section D of the McGuire Nuclear Site Emergency Plan) is available for review if any questions arise over proper classification.

If an event occurs on more than one unit concurrently, the event with the higher classification will be classified on the emergency notification form. Information relating to the problem on the other unit will be captured on the emergency notification form line 13 remarks section.

The Affected Unit(s) on Line 11 is tied to the EAL (IC) Number and EAL (IC) Description on Line

4. Certain events could occur at the plant site such that multiple units are affected. These may include Abnormal Rad Levels/Radiological Effluents; Fire/Explosion and Security Events; and Natural Disasters, Hazards, and Other Conditions Affecting Plant Safety. This shall be considered when evaluating the accuracy of the Unit designation. {PIP 0-M97-463 8 } If the initiating event puts more than one unit in the same Emergency Classification (example Alert), then the unit designation may be either the Affected Unit numbers or All. If the initiating event drives one unit to a higher classification, then the unit with the higher classification should be listed as Affected Unit. {PIPs M 03-3294 and C-04-2586}

The EAL (IC) Number and EAL (IC) Description provided on Line 4 of the emergency notification form should be based on the highest emergency classification that applies. Other classifiable events should be included on Line 13, Remarks, on the emergency notification form, but not given an EAL number. {PIPs M-03-3294 and C-04-2586}

If an event occurs, and a lower or higher plant operating mode is reached before the classification can be made, the classification shall be based on the mode that existed at the time the event occurred.

The fission product barrier matrix is applicable only to those events that occur at hot shutdown or higher. An event that is recognized at cold shutdown or lower shall not be classified using the fission product barrier matrix. Reference would be made to the additional enclosures that provide emergency action levels for specific events (e.g. severe weather, fire, security).

If a transient event should occur, the following guidance is provided.

1. Some emergency action levels specify a specific duration. For these EALs, the classification is made when the Emergency Coordinator assessment concludes that the specified duration is exceeded or will be exceeded (i.e. condition cannot be reasonably corrected before the duration elapses), whichever is sooner.

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 is exceeded, the event is NOT classified by that EAL. Lower Severity EALs, if any, shall be reviewed for possible applicability in these cases.

3. If a plant condition exceeding EAL criteria is not recognized at the time of occurrence, but is identified well after the condition has occurred (e.g. as a result of routine log or record review) and the condition no longer exists, an emergency shall NOT be declared. Reporting under 10CFR5O.72 may be required. Such a condition could occur, for example, if a follow-up evaluation of an abnormal condition uncovers evidence that the condition was more severe than earlier believed.

4. If an emergency classification was warranted, but the plant condition has been corrected prior to declaration and notification, the following are applicable: {2}

a. For UNUSUAL EVENT, the emergency shall be declared and the condition shall be reported. The event should be terminated in a follow-up notification as soon as time permits, but within one hour.

b. For ALERT, SITE AREA EMERGENCY, and GENERAL EMERGENCY, the emergency shall be declared and the Emergency Response Organization shall be activated. The TSC Emergency Coordinator shall be responsible for terminating the emergency as soon as time permits when appropriate.

c. The Control Room Emergency Coordinator (Operations Shift Manager) shall ensure that any required follow-up notifications are conducted as required prior to activation of the TSC.

DETERMINATION OF EVENT TIME (TIME THE 15 MINUTE 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 met momentarily, and in the judgment of the Emergency Coordinator are not likely to recur, the entry into the higher classification must be acknowledged. Acknowledgment is performed as follows:

If this condition occurs prior to the initial notification to the emergency response organization and off site agencies, the initial message should note that the site is currently in the lower classification, but had momentarily met the criteria for the higher classification. It should also be noted that plant conditions have improved and stabilized to the point that the criteria for the higher classification are not expected to be repeated.

Enclosure 4.10 RP/O/A/5700/000 Radiation Monitor Readings for Enclosure 4.3 EALs Page 1 of 1 Note: These values are not intended to apply to anticipated temporary increases due to planned events (e.g. incore detector movement, radwaste container movement, depleted resin transfers, etc.)

Detector Elevation Column Identifier Unusual Alert Event mR/hr mR/hr 1EMF-1 695 FF, GG-56 Aux. Bldg. Corridor 500 5000 1EMF-5 716 FF-54 Unit 1 NM Sample Room 600 5000 1EMF-8 733 HH-56 Aux. Bldg. Corridor 100 5000 1EMF-10 750 LL-56 Aux. Bldg. Corridor 100 5000 1EMF-13 775 QQ-56 Shift Lab/Count Room 100 5000 1EMF-17 786 N/A Unit 1 Spent Fuel Pool Refueling Bridge 100 5000 2EMF-1 716 BE, FF-58 Unit 2 NM Sample Room 300 5000 2EMF-4 786 N/A Unit 2 Spent Fuel Pool Refueling Bridge 100 5000 2EMF-9 767 JJ-59 Aux. Bldg. Corridor 100 5000

Enoi& 4.11 RP/O/A/570O/000 Commitment 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..232 PaeZ9of 32 Revision U MeGuire 1 Cyde 21 Crt Operating Limits Report 2.16 Borated Water Sources Shutdown (SLC 16.9.14) 2.1 .l VoJumc and boroe coacentzians for the &nc Acid Tank (BAT) .nd üie RefuclIng Water 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 gallons 13.6% L.evel

[Not; When cycle boruup is> 455 EFPD Figure 6 maybe used to Lh1 the rcquired BAT mInimum IeveL -

BAT miiiiimuxn boron concentration 7,000 ppm BAT minimum water votume reqiired to 2,G.gallons..

maintain 5DM al 7,0130 ppm RWST rninianum contained borated water 47,70 gallons volume 41 inches RWST minimum boron concentration 2,675 ppm RWST minimum water vourne required to 8200 gallons mthitair SDM al 2,675 ppm

MJE.4J4UU-23Z Pac3I af32 RevisioiO MCufre 1. Cyde 21 Core Operating Lhnit Report figure 6 rkirk Acid Serage Tank Indlcted Level Versus RCS Baron Concentration (Valid When Cycle Eurnup Is> 455 EFPD)

This igureincbdas dditiona[volunieollsted in SLC L69t4 and 16.9ll 4I RS Boron 5-o CQnceatrat1n BAT Luvel (ppi) 00 370 300<500 33.0 500<7GG 28.O 700 ID0 20

.1000<13001 16.

25.0 41

-J 1Jtab1ej Opvt4ion 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 Answer Number 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 of 4

Examination KEYfor: 2011 MNS SRO NRC Examination Question Answer Number 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

Examination KEYfor: 2011 MNS SRO NRC Examination Question Answer Number 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

Examination KEYfor: 2011 MNS SRO NRC Examination Question Answer Number 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 of 4

ML112280604

Text

Navigation menu

Search