Text

301 Final SRO As-Given Written Exam
	ML15141A173
Person / Time
Site:	Mcguire, McGuire
Issue date:	05/20/2015
From:	; NRC/RGN-II
To:	; Duke Energy Carolinas
References
	Download: ML15141A173 (166)
	v • d • e

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

Date: Facility/Unit: MCGUIRE Region: I II III IV Reactor Type: W 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 Results RO/SRO-Only/Total Examination Values / / Points Applicants Score / / Points Applicants Grade / / Percent

McGuire Nuclear Station Question: 1 ILT-31 MNS SRO NRC Examination (1 point)

Regarding the NC pump motor stator coolers,

1) cooling water is supplied from the system.

2) upon initiation of an signal, the cooling water supply will be isolated.

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

A. 1. RN

2. Ss (Safety Injection)

B. 1. RN

2. Sp (Phase B)

C. 1. KC

2. Ss (Safety Injection)

D. 1. KC

2. Sp (Phase B)

Page 1 of 100

McGuire Nuclear Station Question: 2 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

NCS Tavg is 215°F

NCS pressure is 250 PSIG

VCT pressure is 28 PSIG

The 1A NC pump is to be started for a unit heatup Subsequently:

The 1A2 Oil Lift pump is started

Oil Lift pressure is 580 PSIG In accordance with OP/1/A/6150/002A (REACTOR COOLANT PUMP OPERATION)

Enclosure 4.1 (STARTUP AND OPERATION), the MINIMUM required #1 Seal differential pressure for starting the NC pump (1) met.

Based on the conditions above, if the 1A NC PUMP SAFETY BKR "START" pushbutton is depressed, the pump (2) start.

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

A. 1. is

2. will B. 1. is

2. will NOT C. 1. is NOT

2. will D. 1. is NOT

2. will NOT Page 2 of 100

McGuire Nuclear Station Question: 3 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is in solid operations while cooling down

Both trains of ND are in service

A NV pump is in service

Letdown is through 2NV-121 Subsequently:

2A1 KC pump trips Per OP/2/A/6100/SD-8 (WATER SOLID OPERATIONS) which ONE (1) of the following describes operator actions necessary to respond to the failure?

COMPONENT LEGEND:

2NV-121 (ND LETDOWN CONTROL) 2NV-241 (SEAL INJECTION FLOW CONTROL)

A. Throttle CLOSED 2NV-241 OR Throttle OPEN 2NV-121 B. Throttle OPEN 2NV-241 OR Throttle OPEN 2NV-121 C. Throttle CLOSED 2NV-241 OR Throttle CLOSED 2NV-121 D. Throttle OPEN 2NV-241 OR Throttle CLOSED 2NV-121 Page 3 of 100

McGuire Nuclear Station Question: 4 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

The unit is in MODE 5

NC temperature is 112°F

LTOP is in service

1B NV pump is running

1A NV pump is currently tagged out for maintenance

Both NI pumps are tagged out Subsequently:

Maintenance on the 1A NV pump is complete

A 30 minute run of the 1A NV pump must be performed for post-maintenance testing (PMT)

Per Technical Specification 3.4.12 (LTOP SYSTEM),

1) to meet the LCO requirements of the Tech Spec during 1A NV pump PMT, the crew must .

2) LCO relief valve requirements can be met by having two PORVs with a lift setting of less than or equal to PSIG.

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

A. 1. stop the 1B NV pump ONLY

2. 450 B. 1. stop the 1B NV pump and rack out its breaker

2. 450 C. 1. stop the 1B NV pump ONLY

2. 385 D. 1. stop the 1B NV pump and rack out its breaker

2. 385 Page 4 of 100

McGuire Nuclear Station Question: 5 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is in Mode 4 with a cooldown in progress

Both ND trains are in operation

ND pump suction pressure is 30 PSIG

ND pump suction header temperature is 226°F Subsequently:

NC system and ND system temperatures begin to increase due to a reduction of KC flow to the ND heat exchangers Based on the indications above, ND pump cavitation will occur if ND pump suction temperature increases by a MINIMUM of (1) °F.

One indication that the ND pump is cavitating would be that motor amps are (2) .

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

REFERENCE PROVIDED A. 1. 25

2. high B. 1. 48

2. high C. 1. 25

2. fluctuating D. 1. 48

2. fluctuating Page 5 of 100

McGuire Nuclear Station Question: 6 ILT-31 MNS SRO NRC Examination (1 point)

Given the following:

At time 08:10:00, an inadvertent Reactor Trip/Safety Injection occurs due to IAE testing At time 08:10:30, SI (1) be reset.

When the Safety Injection RESET pushbutton is depressed after the required time delay, any subsequent AUTOMATIC actuation signals (2) start safeguards equipment.

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

A. 1. can NOT

2. will B. 1. can NOT

2. will NOT C. 1. can

2. will D. 1. can

2. will NOT Page 6 of 100

McGuire Nuclear Station Question: 7 ILT-31 MNS SRO NRC Examination (1 point)

The purpose of the LTOP System is to prevent a (1) concern.

In accordance with Tech Spec 3.4.12 (LTOP System), the LCO is applicable in MODE 4 if NC system temperature is less than (2) .

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

A. 1. Cold Overpressure

2. 300°F B. 1. Pressurized Thermal Shock (PTS)

2. 300°F C. 1. Cold Overpressure

2. 320°F D. 1. Pressurized Thermal Shock (PTS)

2. 320°F Page 7 of 100

McGuire Nuclear Station Question: 8 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The unit is at 100% RTP

1NC-34A (U1 PZR PORV) has been identified as leaking

Over the last 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , PRT level has increased from 8% to 26%

PRT pressure is 65 PSIG Based on the conditions above, the Tech Spec 3.4.13 (RCS OPERATIONAL LEAKAGE) limit for IDENTIFIED leakage (1) been exceeded.

The PRT rupture disc will rupture if PRT pressure increases to a MINUMUM pressure of (2) .

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

REFERENCE PROVIDED A. 1. has

2. 85 PSIG B. 1. has

2. 100 PSIG C. 1. has NOT

2. 85 PSIG D. 1. has NOT

2. 100 PSIG Page 8 of 100

McGuire Nuclear Station Question: 9 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The unit is in HOT SHUTDOWN on ND Cooling (Both Train A and B)

B Train KC is aligned to supply Reactor and Aux Bldg Non-Essential Headers with both 1B1 and 1B2 pumps in operation

A Train KC is aligned to supply the A ND HX Header with both 1A1 and 1A2 pumps in operation

The 1A1 KC pump has just tripped In accordance with the Limits and Precautions of OP/1/A/6400/005 (Component Cooling Water System), KC flow through the 1A ND Heat Exchanger shall be throttled to less than a MAXIMUM of .

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

A. 2000 GPM B. 4000 GPM C. 5000 GPM D. 6000 GPM Page 9 of 100

McGuire Nuclear Station Question: 10 ILT-31 MNS SRO NRC Examination (1 point)

The Pressurizer Pressure Master Controller soft controls indicate as follows:

With the current ERROR signal, "C" Pzr heaters will be energized (1) of the time.

The Pzr Backup heaters energize at a PRESSURE ERROR of (2) PSIG.

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

A. 1. 17%

2. (-) 25 B. 1. 17%

2. (-) 17 C. 1. 83%

2. (-) 25 D. 1. 83%

2. (-) 17 Page 10 of 100

McGuire Nuclear Station Question: 11 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is at 100% RTP

The top row of SSPS bistable status lights simultaneously illuminate

Power Range Channel N-41 indication is lost

Intermediate Range Channel N-35 indication is lost Which ONE (1) of the following describes the failure that has occurred, AND the response of the Reactor Protection System?

A. Loss of 120 VAC Bus 2EKVB; Train A SSPS General Warning Alarm is received.

B. Loss of 120 VAC Bus 2EKVA; Train A SSPS General Warning Alarm is received.

C. Loss of 120 VAC Bus 2EKVD; Train B SSPS General Warning Alarm is received.

D. Loss of 120 VAC Bus 2EKVC; Train B SSPS General Warning Alarm is received.

Page 11 of 100

McGuire Nuclear Station Question: 12 ILT-31 MNS SRO NRC Examination (1 point)

Given the following:

Unit 1 is at 100% power

A LBLOCA inside containment occurs

Containment pressure is 3.7 PSIG and stable

Safety Injection Train "B" fails to actuate Based on the conditions above AND PRIOR TO any operator actions,

1) Phase A, Train "B" containment isolation valves automatically CLOSE.

2) Phase B containment isolation valves on will automatically CLOSE.

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

A. 1. will

2. Train A ONLY B. 1. will

2. both Trains C. 1. will NOT

2. Train A ONLY D. 1. will NOT

2. both Trains Page 12 of 100

McGuire Nuclear Station Question: 13 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

A LOCA has occurred

Containment pressure peaked at 2.9 PSIG and now is 2.6 PSIG and stable Which ONE (1) of the following describes the operation of the Containment Cooling system based on these conditions?

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

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

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

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

Page 13 of 100

McGuire Nuclear Station Question: 14 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

Unit 1 is at 100% RTP

The Control Room has received Floor Cooling Glycol alarms on annunciator panel 1AD-9

An operator has been dispatched per the annunciator response procedure

Floor Cooling Glycol Temperature is 22°F

Ice bed temperature is 27°F Based on the conditions above,

1) the required actions per the annunciator response for Floor Cooling Glycol temperature, is to .

2) increased ice bed sublimation a concern.

A. 1. stop one Floor Cooling pump if both are running

2. is B. 1. stop one Floor Cooling pump if both are running

2. is NOT C. 1. start an additional Floor Cooling pump if available

2. is D. 1. start an additional Floor Cooling pump if available

2. is NOT Page 14 of 100

McGuire Nuclear Station Question: 15 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

A Large Break LOCA has occurred

"B" train of NS has been aligned per ES-1.3 (TRANSFER TO COLD LEG RECIRC)

2NI-185A (2A ND PUMP SUCTION FROM CONT SUMP ISOL) failed to OPEN from the control room

The crew is aligning ND aux spray Based on the conditions above,

1) 2NS-43A (2A ND HX OUTLET TO NS CONT OUTSIDE ISOL) OPEN from the control room.

2) if 2NS-43A is OPEN, when containment pressure decreases to less than 0.35 PSIG, 2NS-43A .

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

A. 1. will

2. will CLOSE automatically B. 1. will

2. must be CLOSED manually C. 1. will NOT

2. will CLOSE automatically D. 1. will NOT

2. must be CLOSED manually Page 15 of 100

McGuire Nuclear Station Question: 16 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

Unit is at 75% RTP Subsequently,

A small steam leak develops on 1A S/G

NC system pressure is 2210 PSIG and STABLE Based on the conditions above,

1) a Main Steam Isolation will occur if the 1A S/G pressure decreases to less than a MINIMUM of PSIG.

2) if a Main Steam Isolation occurs, the SM PORVs .close.

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

A. 1. 775

2. will NOT B. 1. 875

2. will NOT C. 1. 775

2. will D. 1. 875

2. will Page 16 of 100

McGuire Nuclear Station Question: 17 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

A loss of all feedwater has occurred

2A S/G is faulted inside Containment

Containment pressure has peaked at 2.8 PSIG and is stable

NO CA flow is available

FR-H.1 (RESPONSE TO LOSS OF SECONDARY HEAT SINK) has been implemented Given the following parameters:

TIME SG WR Level 1515 1530 2A S/G 10% 10%

2B S/G 37% 25%

2C S/G 33% 21%

2D S/G 35% 23%

Per FR-H.1 foldout page, the EARLIEST time that the crew is required to implement NC System Feed and Bleed is (1) .

Per FR-H.1, a MINIMUM of (2) PZR PORV(s) must be opened to establish NC system Feed and Bleed.

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

A. 1. 1515

2. ONE B. 1. 1515

2. TWO C. 1. 1530

2. ONE D. 1. 1530

2. TWO Page 17 of 100

McGuire Nuclear Station Question: 18 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

The unit is at 98% RTP

In preparation for a Unit 1 TDCA pump performance test the following flow control valves are positioned as follows:

1CA-64AB (TD CA PUMP TO 1A S/G) -- CLOSED 1CA-52AB (TD CA PUMP TO 1B S/G) -- CLOSED 1CA-48AB (TD CA PUMP TO 1C S/G) -- OPEN 1CA-36AB (TD CA PUMP TO 1D S/G) -- OPEN Subsequently,

An IAE technician inadvertently generates a U1 TDCA pump auto-start signal After the inadvertent auto-start signal is initiated, (1) of the U1 TDCA Flow Control valves will be OPEN.

In accordance with the Control Room Crew Expectations Manual, the crew will CLOSE any OPEN U1 TDCA Flow Control valves (2) .

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

A. 1. all four

2. as soon as practical B. 1. all four

2. when directed by OP/1/A/6250/002 (AUXILIARY FEEDWATER SYSTEM)

C. 1. only two

2. as soon as practical D. 1. only two

2. when directed by OP/1/A/6250/002 (AUXILIARY FEEDWATER SYSTEM)

Page 18 of 100

McGuire Nuclear Station Question: 19 ILT-31 MNS SRO NRC Examination (1 point)

Given the following on Unit 2:

Unit 2 is at 100% RTP

2A D/G has been started per PT/2/A/4350/002 A (DIESEL GENERATOR 2A OPERABILITY TEST)

2A D/G has been running idle for 45 minutes Based on the conditions above, the 2A D/G should be loaded to a MINIMUM of 3000 kW AND run for one hour to ensure (1) .

The 2A D/G load limit for CONTINUOUS OPERATION is (2) kW.

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

A. 1. injector tips are clean

2. 4000 B. 1. burnout of excess fuel in cylinders

2. 4000 C. 1. injector tips are clean

2. 4400 D. 1. burnout of excess fuel in cylinders

2. 4400 Page 19 of 100

McGuire Nuclear Station Question: 20 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is in MODE 3

All NCPs are running and powered from their normal sources

A Generator-Switchyard protective lockout occurs on 2A Bus Line

AUTO / MAN mode selector switches are in AUTO Based on the conditions above,

1) NCP 2A .

2) Bus 2TA automatically transfers to its alternate power supply.

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

A. 1. trips

2. slow B. 1. trips

2. fast C. 1. continues to run

2. slow D. 1. continues to run

2. fast Page 20 of 100

McGuire Nuclear Station Question: 21 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

A BLACKOUT has occurred on 2ETB

D/G '2B' failed to start due to an 86N relay actuation

2AD-11 / F4 (BATT EVCD UNDERVOLTAGE) is in alarm

EVDD bus voltage is 113 VDC and lowering slowly In accordance with AP-07 (LOSS OF ELECTRICAL POWER), the action required to restore bus EVDD is to .

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

A. cross tie Bus EVDB to Bus EVDD B. align Battery Charger EVCS to Bus EVDD C. swap Battery Charger EVCD power supply to 1EMXB D. swap Battery Charger EVCD power supply to 2EMXH Page 21 of 100

McGuire Nuclear Station Question: 22 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2:

A loss of voltage has occurred on 2ETA

Blackout loading is in progress Subsequently:

A Safety Injection signal is received before Blackout loading is completed on 2ETA Based on the conditions above, the Blackout load sequence (1) , 2ETA is cleared of (2) loads, and the SI load sequence is actuated.

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

A. 1. stops

2. all B. 1. is completed

2. all C. 1. stops

2. all non-SI D. 1. is completed

2. all non-SI Page 22 of 100

McGuire Nuclear Station Question: 23 ILT-31 MNS SRO NRC Examination (1 point)

Regarding Diesel Generator Auxiliaries:

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

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

A. 1. VG After Coolers

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

2. 11.5 inches C. 1. VG After Coolers

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

2. 25 inches Page 23 of 100

McGuire Nuclear Station Question: 24 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

Unit is at 100% RTP

A leak on the RC piping in the turbine building basement has occurred

All TB Sump pumps are in Manual and ON, maintaining sump level stable Subsequently:

A detector failure occurs due to a failed power supply on 1EMF-31 (TURBINE BUILDING SUMP MONITOR)

Based on the conditions above,

1) the Unit 1TB Sump pumps trip automatically.

2) to continue with the leak mitigation, the crew will be required to .

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

A. 1. will NOT

2. open 1WP-35 (WMT & VUCDT TO RC CONTROL)

B. 1. will

2. clear the EMF-31 Trip 2 signal and restart the TB sump pumps C. 1. will NOT

2. open 1WP-6 (U1 TB SUMP PUMPS DISCH TO WC ISOL)

D. 1. will

2. place the HI RAD INH/BYP switch in BYP and restart the TB sump pumps Page 24 of 100

McGuire Nuclear Station Question: 25 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

I&E has requested that the 1A Nuclear Service Water Pump breaker be racked out for lubrication To which ONE (1) of the following locations should an Operator be dispatched to rack out the breaker?

A. 1TA B. 1TD C. 1ETA D. 1ETB Page 25 of 100

McGuire Nuclear Station Question: 26 ILT-31 MNS SRO NRC Examination (1 point)

Concerning the VI System, 1VI-1812 (VI AIR DRYER BYPASS FILTER ISOL) solenoid will vent the actuator when VI system pressure decreases to less than a MAXIMUM of (1) PSIG.

RESET must be depressed on the local VI (2) to CLOSE 1VI-1812.

Which ONE of the following completes the statements above?

A. 1. 90

2. Sequencer Control Panel B. 1. 90

2. Reflash Panel C. 1. 85

2. Sequencer Control Panel D. 1. 85

2. Reflash Panel Page 26 of 100

McGuire Nuclear Station Question: 27 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions:

Both Units are at 100% RTP

VI (INSTRUMENT AIR) compressors are in a D, E, F alignment

VS (STATION AIR) is in a normal alignmment Subsequently:

A VI header rupture occurs

Annunciators 1AD-12 / C1 (VI/VS LO PRESSURE) AND 1AD-12 / D1 (VI/VS LO LO PRESSURE) are in alarm Based on the conditions above, 1VI-820 (VI TO VS CONTROL VALVE) will auto-close when VI system pressure decreases to less than a MAXIMUM of (1) PSIG AND the VS compressor (2) auto-start.

A. 1. 85

2. will B. 1. 90

2. will C. 1. 85

2. will NOT D. 1. 90

2. will NOT Page 27 of 100

McGuire Nuclear Station Question: 28 ILT-31 MNS SRO NRC Examination (1 point)

Regarding Containment isolation signals,

1) the S/G CF Containment Isolation valves (CF-35, 30, 28, & 26) will close if Containment pressure increases to a MINIMUM of PSIG.

2) a Containment Phase A isolation will occur if NC system pressure decreases to less than a MAXIMUM of PSIG.

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

A. 1. 1.0

2. 1845 B. 1. 1.0

2. 1945 C. 1. 3.0

2. 1845 D. 1. 3.0

2. 1945 Page 28 of 100

McGuire Nuclear Station Question: 29 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is at 100% RTP

Annunciator 1AD-2 / A10 (ROD CONTROL URGENT FAILURE) is LIT

I&E had determined that the alarm is the result of an urgent failure in the Logic Cabinet Based on the conditions above,

1) one of the possible causes of the Logic Cabinet urgent failure is a .

2) the Logic Cabinet Urgent Failure blocks rod motion in .

A. 1. Slave Cycler failure

2. AUTO ONLY B. 1. Slave Cycler failure

2. AUTO AND MANUAL C. 1. Phase failure

2. AUTO ONLY D. 1. Phase failure

2. AUTO AND MANUAL Page 29 of 100

McGuire Nuclear Station Question: 30 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

Unit is at 25% RTP

Power ascension to 50% RTP is in progress

"A" Pzr heaters are ON Subsequently,

A DCS malfunction occurs in the Pzr Level Median Select for Selected Pzr Level 1

Selected Pzr Level 1 fails at its current output

No operator action is taken Which ONE (1) of the following statements describes the plant response as the power ascension continues?

A. Charging flow decreases Letdown isolates B. Charging flow increases Pzr backup heaters energize C. Charging flow decreases Letdown will NOT isolate D. Charging flow increases Pzr backup heaters will NOT energize Page 30 of 100

McGuire Nuclear Station Question: 31 ILT-31 MNS SRO NRC Examination (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)

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

A. D ONLY B. C ONLY C. A and B ONLY D. C and D ONLY Page 31 of 100

McGuire Nuclear Station Question: 32 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The unit is at 100% RTP

A malfunction occurs which results in a DATA A Failure for Control Rod H-6 on DRPI

At the time of the failure, the highest DRPI coil penetrated by Control Rod H-6 was a "B" Coil

The actual position of all control rods has remained the same After the failure, the DRPI indication for Control Rod H-6 .

In addition to the DATA A Failure alarm, a DRPI alarm will also be received.

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

A. 1. remains the same

2. Non-Urgent Failure B. 1. decreases

2. Non-Urgent Failure C. 1. remains the same

2. Urgent Failure D. 1. decreases

2. Urgent Failure Page 32 of 100

McGuire Nuclear Station Question: 33 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

A Large Break LOCA has occurred

Subcooling based on the 5 HI T/C AVG indicates negative (-) 4°F on the Inadequate Core Cooling Monitor (ICCM)

Based on the conditions above, ICCM indication of Subcooling based on the 5 HI T/C AVG (1) be displayed in reverse video.

The Core Exit Thermocouples (CETs) will indicate a MAXIMUM temperature of (2) .

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

A. 1. will

2. 1200°F B. 1. will NOT

2. 1200°F C. 1. will

2. 2300°F D. 1. will NOT

2. 2300°F Page 33 of 100

McGuire Nuclear Station Question: 34 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

PT/1/A/4450/003 A (ANNULUS VENTILATION SYSTEM TRAIN A OPERABILITY TEST) is being performed

During the test, a VE filter pre-heater malfunctions When the 1A VE Filter temperature rises to a MINIMUM of (1) °F, annunciator 0AD-12 / F2 (1A VE FILTER HI TEMP) will alarm.

As a result of this alarm, the (2) .

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

A. 1. 220

2. 1A VE pre-heaters trip ONLY B. 1. 220

2. 1A VE pre-heaters and 1A VE Fan trip C. 1. 325

2. 1A VE pre-heaters trip ONLY D. 1. 325

2. 1A VE pre-heaters and 1A VE Fan trip Page 34 of 100

McGuire Nuclear Station Question: 35 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2:

Unit is in Mode 6

VP (CONTAINMENT PURGE SYSTEM) is in service and refueling is in progress Subsequently:

A Trip 2 alarm on 2EMF-38(L) (CONTAINMENT PARTICULATE) is received Based on the conditions above,

1) the VP Supply and Exhaust .

2) to regain control of VP components, 2EMF-38 must be reset and the Containment Ventilation (SH) Reset push button located on must be depressed.

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

A. 1. fans will be "OFF" ONLY

2. 2MC-11 B. 1. fans will be "OFF" ONLY

2. Unit 2 HVAC panel C. 1. fans will be "OFF" AND dampers will be CLOSED

2. 2MC-11 D. 1. fans will be "OFF" AND dampers will be CLOSED

2. Unit 2 HVAC panel Page 35 of 100

McGuire Nuclear Station Question: 36 ILT-31 MNS SRO NRC Examination (1 point)

Given the following on Unit 2:

The unit is 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) If a Trip 2 signal is received on 2EMF-42, what AUTOMATIC actions occur?

CONSIDER EACH QUESTION SEPARATELY.

COMPONENT LEGEND:

2EMF-4 (SPENT FUEL BLDG REFUEL BRDG) 2EMF-42 (UNIT 2 FUEL BUILDING VENTILATION)

A. 1. 2EMF-4

2. The VF Supply and Exhaust Fans will stop B. 1. 2EMF-42

2. The VF Supply and Exhaust Fans will stop C. 1. 2EMF-4

2. The Exhaust Filter Bypass Damper (D-5) will close D. 1. 2EMF-42

2. The Exhaust Filter Bypass Damper (D-5) will close Page 36 of 100

McGuire Nuclear Station Question: 37 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The unit is at 47% RTP for repairs to the 1B CF Pump

The 1A CF Pump is in AUTOMATIC S/G NR Level (%)

1A 45 1B 51 1C 50 1D 47 Based on current conditions, a S/G LEVEL DEVIATION annunciator is LIT for S/G(s)

(1) .

In accordance with the Annunciator Response Procedure for S/G LEVEL DEVIATION, the crew will take manual control of the (2) to restore S/G levels to program.

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

A. 1. 1A ONLY

2. CF Control or Bypass Valves B. 1. 1A AND 1D

2. CF Control or Bypass Valves C. 1. 1A ONLY

2. 1A CF Pump D. 1. 1A AND 1D

2. 1A CF Pump Page 37 of 100

McGuire Nuclear Station Question: 38 ILT-31 MNS SRO NRC Examination (1 point)

Regarding the CF&E (CONTAINMENT FLOOR AND EQUIPMENT) Sumps,

1) one input into the CF&E sumps is .

2) the CF&E sumps discharge is aligned to the .

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

A. 1. VU AHU drains

2. FDT (Floor Drain Tank)

B. 1. VU AHU drains

2. WMT (Waste Monitoring Tank)

C. 1. Ice Condenser Drains

2. FDT (Floor Drain Tank)

D. 1. Ice Condenser Drains

2. WMT (Waste Monitoring Tank)

Page 38 of 100

McGuire Nuclear Station Question: 39 ILT-31 MNS SRO NRC Examination (1 point)

Given the following on Unit 1:

A Loss of Offsite power has occurred

1A and 1B D/Gs are supplying the 4160V busses

The crew has completed E-0 (REACTOR TRIP OR SAFETY INJECTION) and entered ES-0.1 (REACTOR TRIP RESPONSE)

NCS Tavg is 552°F and slowly lowering Based on the conditions above,

1) all Feedwater Isolation status lights on 1SI-4 be lit.

2) ES-0.1 will check NC system stable or trending to 557°F.

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

A. 1. will

2. Tavg B. 1. will NOT

2. Tavg C. 1. will

2. Tcolds D. 1. will NOT

2. Tcolds Page 39 of 100

McGuire Nuclear Station Question: 40 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is operating at 100% RTP

One PZR PORV is leaking past its seat

Pressurizer pressure is 2235 PSIG

Pressurizer Steam Space temperature is 653°F

PRT pressure is 15 PSIG Which ONE (1) of the following is the approximate expected temperature downstream of the leaking PZR PORV?

REFERENCE PROVIDED A. 220°F B. 240°F C. 250°F D. 300°F Page 40 of 100

McGuire Nuclear Station Question: 41 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

Unit is at 100% RTP

2B KC surge tank level is slowly trending up

2EMF-46B (COMPONENT COOLING TRAIN B) is in a Trip 2 condition The indications above can be caused by a leak on the (1) heat exchanger.

When 2EMF-46B Trip 2 clears, 2KC-122 (KC SURGE TANK VENT VALVE)

(2) .automatically re-OPEN.

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

A. 1. Letdown

2. will NOT B. 1. Letdown

2. will C. 1. Seal Water Return

2. will NOT D. 1. Seal Water Return

2. will Page 41 of 100

McGuire Nuclear Station Question: 42 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

Unit was operating at 90% RTP

1NV-7B (L/D CONT OUTSIDE ISOL) failed CLOSED

The operators entered AP-12 (LOSS OF LETDOWN, CHARGING OR SEAL INJECTION)

Excess L/D has been placed in service 1NV-24B (C NC LOOP TO EXS L/D HX ISOL) and 1NV-25B (C NC LOOP TO EXS L/D HX ISOL) can be controlled from the C/R AND the (1) .

Per AP-12, when placing Excess L/D in service, 1NV-26B (U1 EXCESS L/D HX OUTLET CNTRL) is cycled OPEN for two minutes and then CLOSED to (2) .

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

A. 1. SSF

2. minimize possible reactivity excursions B. 1. ASP

2. minimize possible reactivity excursions C. 1. SSF

2. reduce the possibility of water hammers D. 1. ASP

2. reduce the possibility of water hammers Page 42 of 100

McGuire Nuclear Station Question: 43 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2 :

Both trains of Residual Heat Removal (ND) are in service

ND heat exchanger outlet valves (2ND-14 & 2ND-29) are throttled to 2000 GPM each

The ND heat exchanger bypass valve (2ND-34) is throttled to 50% OPEN Subsequently:

A loss of Instrument Air (VI) occurs Based on the conditions above,

1) 2ND-14 and 2ND-29 fail .

2) 2ND-34 fails .

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

A. 1. open

2. closed B. 1. open

2. open C. 1. closed

2. closed D. 1. closed

2. open Page 43 of 100

McGuire Nuclear Station Question: 44 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The Pressurizer Pressure Calculated Error observed on the Ovation Soft Panel has failed to +100 PSIG

1AD-2 / E8 (DCS TROUBLE) is in alarm

Actual Pressurizer Pressure is 2200 PSIG and decreasing Based on the conditions above, which ONE (1) of the following indicates the status of the Pressurizer Pressure Control system?

A. PORV 1NC-34A is OPEN PZR Spray Valves are OPEN B. PORV 1NC-34A is OPEN PZR Spray Valves are CLOSED C. PORV 1NC-34A is CLOSED PZR Spray Valves are OPEN D. PORV 1NC-34A is CLOSED PZR Spray Valves are CLOSED Page 44 of 100

McGuire Nuclear Station Question: 45 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

A valid First Out annunciator is LIT on Annunciator Panel 2FO-1

Both Reactor Trip breakers indicate OPEN

IR SUR meters indicate +0.5 DPM

All WR Neutron flux indications are stable

The STA is manually evaluating the Critical Safety Function Status Trees (CSFSTs) because SPDS in not working Based on the conditions above, the CSFSTs require implementation of FR-S.1 (RESPONSE TO NUCLEAR GENERATION / ATWS) if indicated power is greater than a MINIMUM of (1) OR if WR Neutron Flux indicates greater than a MINIMUM of (2) .

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

A. 1. 5 %

2. 10-3 %

B. 1. 5 %

2. 10-5 %

C. 1. 10 %

2. 10-3 %

D. 1. 10 %

2. 10-5 %

Page 45 of 100

McGuire Nuclear Station Question: 46 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The 1D S/G is ruptured

E-3 (STEAM GENERATOR TUBE RUPTURE) has been implemented

The first NC system depressurization was stopped with the following indications:

NC system pressure 1250 PSIG 1D S/G pressure 1050 PSIG

SI has been terminated and the crew is preparing for a second NC system depressurization In accordance with the SGTR mitigating strategy,

1) the second NC system depressurization is performed to .

2) the reason for initially establishing a minimum level in the ruptured S/G is to prevent .

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

A. 1. prevent S/G over-fill

2. ruptured S/G depressurization B. 1. establish indicated level in the PZR

2. ruptured S/G depressurization C. 1. prevent S/G over-fill

2. additional damage of ruptured S/G tubes D. 1. establish indicated level in the PZR

2. additional damage of ruptured S/G tubes Page 46 of 100

McGuire Nuclear Station Question: 47 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

A Steam line break has occurred

NC pressure is 1700 PSIG and slowly lowering

Tavg is 518°F and slowly lowering

Containment pressure is 2.9 PSIG and stable on channels 1, 2 and 4 and 3.1 PSIG and stable on channel 3

S/G 1A pressure is 755 PSIG on channels 1 and 2 and 765 PSIG on channel 4 and slowly lowering on all channels

NO operator actions have been taken Based on the conditions above,

1) MSIV and MSIV Bypass valves are .

2) The purpose of CLOSING MSIVs and MSIV Bypass valves in E-2 (FAULTED STEAM GENERATOR ISOLATION) is to .

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

A. 1. OPEN

2. terminate an uncontrolled cooldown B. 1. OPEN

2. isolate a faulted S/G from the non-faulted S/Gs C. 1. CLOSED

2. terminate an uncontrolled cooldown D. 1. CLOSED

2. isolate a faulted S/G from the non-faulted S/Gs Page 47 of 100

McGuire Nuclear Station Question: 48 ILT-31 MNS SRO NRC Examination (1 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 NCS Press (PSIG) 1965 1960 1976 1991 NR SG A (%) 19 18 19 19 NR SG B (%) 20 18 17 16 NR SG C (%) 20 19 18 16 NR SG D (%) 18 16 18 19 Based on the conditions above,

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

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

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 48 of 100

McGuire Nuclear Station Question: 49 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions:

MNS has experienced a Station Blackout

The 1A D/G fails to start

The 1B D/G started and tripped on low lube oil pressure Based on the conditions above,

1) the vital DC batteries that should be monitored for decaying voltage are .

2) the battery discharge rate will on the vital DC batteries with decaying voltage, until the design battery capacity is exhausted.

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

A. 1. EVCA and EVCB

2. remain constant B. 1. EVCA and EVCB

2. increase steadily C. 1. EVCC and EVCD

2. remain constant D. 1. EVCC and EVCD

2. increase steadily Page 49 of 100

McGuire Nuclear Station Question: 50 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

A Loss of Offsite Power has occurred

The crew is verifying natural circulation flow per EP/1A/5000/ G-1 (Generic Enclosures) Enclosure 33 (NATURAL CIRCULATION PARAMETERS)

Given the following parameters:

1. NC system Subcooling > 0°F

2. NC system hot leg temperatures at saturation temperature for S/G pressure

3. NC system cold leg temperatures going up slowly

4. NC system hot leg temperatures going down

5. S/G pressure stable

6. NC system cold leg temperatures at saturation temperature for S/G pressure

7. NC system pressure stable

8. Core Exit T/C's stable Which ONE (1) of the following sets of conditions confirm that Natural Circulation exists and is effective in cooling the core in accordance with G-1, Enclosure 33?

A. 1, 3, 4, 5, 7 B. 2, 3, 5, 7, 8 C. 1, 4, 5, 6, 8 D. 1, 2, 5, 7, 8 Page 50 of 100

McGuire Nuclear Station Question: 51 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2:

Both CFPT's tripped causing a Rx Trip 15 minutes ago

The crew has completed E-0 (REACTOR TRIP OR SAFETY INJECTION) and transitioned to ES-0.1 (REACTOR TRIP RESPONSE)

Subsequently:

EVDA output breaker to 2EVIA inverter trips OPEN Based on the conditions above,

1) which ONE (1) of the following indicates the impact on the CA system flow instrumentation?

2) what alternate indication can be used to determine the status of CA flow to the affected S/G?

A. 1. 2A S/G CA flow fails low

2. 2A CA Pump amps and breaker indicating lights B. 1. 2B S/G CA flow fails low

2. 2B CA Pump amps and breaker indicating lights C. 1. 2C S/G CA flow fails low

2. 2A CA Pump amps and breaker indicating lights D. 1. 2D S/G CA flow fails low

2. 2B CA Pump amps and breaker indicating lights Page 51 of 100

McGuire Nuclear Station Question: 52 ILT-31 MNS SRO NRC Examination (1 point)

Given the following on Unit 1:

Unit is at 100% RTP

125VDC Battery CXB is aligned for an "equalizing charge" Subsequently:

A fault on bus DCB causes the CXB Battery charger output breaker and the DCA -DCB cross tie breakers to OPEN Based on the conditions above and per AP-15 (LOSS OF VITAL OR AUX CONTROL POWER),

1) switch indication on any component powered from 6.9 kV switchgear will be DARK.

2) breakers powered from the affected 6.9 kV switchgear be remotely operated.

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

A. 1. 1TB ONLY

2. can B. 1. 1TB AND 1TD

2. can C. 1. 1TB ONLY

2. can NOT D. 1. 1TB AND 1TD

2. can NOT Page 52 of 100

McGuire Nuclear Station Question: 53 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

Unit power ascension is in progress

Unit is at 660 MW Subsequently:

MVARs are (-) 500 and slowly decreasing

The operating crew enters AP-05 (GENERATOR VOLTAGE AND ELECTRICAL GRID DISTURBANCES)

1) If the voltage regulator is placed in MANUAL, the voltage regulator under excitation limiter function to reduce leading MVARs.

2) Based on the conditions above and per AP-05, if unable to maintain MVARs within limits of the generator capability curve, then the crew will .

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

PROCEDURE LEGEND:

AP-02 (TURBINE TRIP)

E-0 (REACTOR TRIP OR SAFETY INJECTION)

A. 1. will

2. trip Unit 1 Turbine AND GO TO AP-02 B. 1. will

2. trip Unit 1 Reactor AND GO TO E-0 C. 1. will NOT

2. trip Unit 1 Turbine AND GO TO AP-02 D. 1. will NOT

2. trip Unit 1 Reactor AND GO TO E-0 Page 53 of 100

McGuire Nuclear Station Question: 54 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

ECA-1.2 (LOCA OUTSIDE CONTAINMENT) has been implemented

NC System pressure is 1700 psig and stable In accordance with ECA-1.2,

1) the crew will FIRST stop and isolate the pumps from the FWST.

2) the overall mitigating strategy includes cooldown and depressurization of the NCS to allow the .

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

A. 1. ND

2. Cold Leg Accumulators to inject B. 1. NI

2. Cold Leg Accumulators to inject C. 1. ND

2. ND isolation valves (1NI-173A and 1NI-178B) to close D. 1. NI

2. ND isolation valves (1NI-173A and 1NI-178B) to close Page 54 of 100

McGuire Nuclear Station Question: 55 ILT-31 MNS SRO NRC Examination (1 point)

Given the following condition on Unit 1:

S/G's 1C & 1D are indicating 30% Narrow Range Level

S/G's 1A & 1B are faulted and indicating <5% Wide Range Level

All CA is unavailable

Containment Pressure is 3.5 PSIG

E-0 (REACTOR TRIP OR SAFETY INJECTION) has been completed Based on the conditions above, which ONE (1) of the following indicates the NEXT procedure to be implemented AND the action(s) required?

PROCEDURE LEGEND:

E-2 (FAULTED S/G ISOLATION)

FR-H.1 (RESPONSE TO LOSS OF SECONDARY HEAT SINK)

A. Go to E-2; Isolate C & D S/G's B. Go to FR-H.1; Commence NCS feed and bleed C. Go to E-2; Close all MSIVs and MSIV bypasses D. Go to FR-H.1; Restore feed water flow to C & D S/G's Page 55 of 100

McGuire Nuclear Station Question: 56 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

A Large Break LOCA has occurred inside Containment

A and B ND pumps are not available

The Control room crew has implemented ECA-1.1 (LOSS OF EMERGENCY COOLANT RECIRC)

Containment pressure is 8 PSIG and slowly rising

FWST level is 105 inches and lowering When the FWST Level LO setpoint is reached, 1NI-184B (1B ND PUMP SUCTION FROM CONT SUMP ISOL) AND 1NI-185A (1A ND PUMP SUCTIONFROM CONT SUMP ISOL) (1) automatically OPEN.

Per ECA-1.1 Foldout Page, when FWST level decreases to less than a MAXIMUM of (2) inches ALL ECCS pumps must be secured.

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

A. 1. will

2. 95 B. 1. will NOT

2. 95 C. 1. will

2. 20 D. 1. will NOT

2. 20 Page 56 of 100

McGuire Nuclear Station Question: 57 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

A Unit runback has occurred

Rods are inserting in AUTO

Rod M-12 in Control Bank D is NOT inserting with its bank Subsequently:

A Rod Control Urgent Failure alarm is received

Unit1 Control Rod Bank Select is taken to MANUAL

I&E determines source of alarm is Power Cabinet 1BD Due to the Power Cabinet Rod Control Urgent Failure alarm, Group 1 rods in Control Banks B and D will NOT move in (1) .

When Unit 1 OATC attempts to continue control rod insertion, rods in all groups of the other banks (2) insert.

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

A. 1. AUTO ONLY

2. will B. 1. AUTO OR MANUAL

2. will C. 1. AUTO ONLY

2. will NOT D. 1. AUTO OR MANUAL

2. will NOT Page 57 of 100

McGuire Nuclear Station Question: 58 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2:

A loss of offsite power has occurred

The crew has commenced a cooldown and depressurization in accordance with ES-0.2 (NATURAL CIRCULATION COOLDOWN)

The following conditions are observed:

0220 0230 NC Pressure 1685 1635 (PSIG)

T-Colds (°F) 602 598 CETs (°F) 612 610

At time 0230 Pressurizer level begins increasing rapidly In accordance with ES-0.2, the MAXIMUM allowable cooldown rate is (1) .

The cause of the Pressurizer level increase is (2) .

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

A. 1. 50°F / Hr

2. voiding in the Reactor Vessel head B. 1. 50°F / Hr

2. an increase in flow from the NI pumps C. 1. 100°F / Hr

2. voiding in the Reactor Vessel head D. 1. 100°F / Hr

2. an increase in flow from the NI pumps Page 58 of 100

McGuire Nuclear Station Question: 59 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

Shutdown Bank Rods are withdrawn in preparation for a Reactor startup

The startup is on hold while I&E completes an inspection of the Channel 1 Nuclear Instrument cabinet PROCEDURE LEGEND:

Tech Spec 3.3.1 (RTS INSTRUMENTATION)

In accordance with Tech Spec 3.3.1,

1) based on the conditions above, channel(s) of Source Range Nuclear Instrumentation is/are currently required to be OPERABLE.

2) if the I&E technician caused a a loss of power to Source Range Channel N31, the crew required to IMMEDIATELY open the Reactor Trip Breakers (RTBs).

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

REFERENCE PROVIDED A. 1. ONE

2. is B. 1. ONE

2. is NOT C. 1. BOTH

2. is D. 1. BOTH

2. is NOT Page 59 of 100

McGuire Nuclear Station Question: 60 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2:

A reactor startup is being performed per OP/2/A/6100/003 (CONTROLLING PROCEDURE FOR UNIT OPERATION)

Reactor power increase to allow taking critical rod height data is in progress

Reactor power is 7X10-6 % (IR)

Subsequently:

The IR Signal Processor for detector channel N36 fails Based on the conditions above,

1) per Tech Spec 3.3.1 (RTS INSTRUMENTATION), the power increase

2) Reactor power indication on has been lost.

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

A. 1. can continue

2. N36 ONLY B. 1. must be suspended

2. N36 ONLY C. 1. can continue

2. N32 AND N36 D. 1. must be suspended

2. N32 AND N36 Page 60 of 100

McGuire Nuclear Station Question: 61 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

2EMF-59 (EQUIPMENT STAGING BUILDING VENTILATION MONITOR) is in Trip 2 alarm due to a release in the building

The VK (EQUIPMENT STAGING BUILDING VENT) system selector switch is in the "ON" position Which ONE (1) of the following describes the actions, if any, that will occur as a result of the Trip 2 alarm on 2EMF-59?

A. The VK Supply fans ONLY will trip B. The VK Exhaust AND Supply fans will trip C. The VK exhaust filter bypass damper will CLOSE D. NO automatic actions will occur Page 61 of 100

McGuire Nuclear Station Question: 62 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

Unit 1 is at 100% RTP

1AD-13 / E3 (FIRE DET SYS ALERT) is in alarm

An electrical fire inside the auxiliary building cable spreading room has been reported

AP-45 (PLANT FIRE) has been implemented Fire suppression for the affected area will be accomplished by .

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

A. automatic halon actuation B. automatic sprinkler actuation C. an AO dispatched to open a MANUAL deluge valve D. an AO dispatched to actuate a manual Cardox system Page 62 of 100

McGuire Nuclear Station Question: 63 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is at 100% RTP

2EMF-6 (REACTOR COOLANT FILTER 2B) is in Trip 2 alarm

2EMF-48 (REACTOR COOLANT HI RAD) is also in Trip 2 alarm

The crew has implemented AP-18 (HIGH ACTIVITY IN REACTOR COOLANT)

Chemistry has reported that the cause of the high activity is due to FAILED FUEL In accordance with the mitigating strategy for AP-18, the crew will (1) .

The reason for performing this action is because it (2) .

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

A. 1. ensure that a mixed bed demineralizer is in service

2. causes a pH change that prevents further fuel degradation B. 1. ensure that a mixed bed demineralizer is in service

2. facilitates the removal of fission products resulting from the failed fuel C. 1. increase letdown to 120 GPM

2. increases the effectiveness of the fission product gas removal by the VCT D. 1. increase letdown to 120 GPM

2. increases the removal rate of fission products by the demineralizer Page 63 of 100

McGuire Nuclear Station Question: 64 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The operating crew initiated a manual SI due to a small break LOCA

Equipment failures resulted in a RED condition on the Integrity CSF Status Tree

NC Cooldown rate was approximately 220°F/hr

NC System temperature is currently 240°F

The crew is performing a soak in accordance with FR-P.1 (RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION)

Based on the conditions given, which ONE (1) of the following actions is permitted by FR-P.1 during the soak?

A. Energize PZR heaters B. Start an additional NV Pump C. Place Auxiliary Spray in service D. Increase CA flow to recover S/G NR level Page 64 of 100

McGuire Nuclear Station Question: 65 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

Unit tripped from 100% RTP

A Loss of Off-Site Power has occurred

ES-0.2 (NATURAL CIRCULATION COOLDOWN) has been implemented Blocking automatic Safety Injection (1) required to be performed in ES-0.2 prior to implementing ES-0.3 (NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL).

The FIRST Major Action associated with ES-0.3 is (2) .

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

A. 1. is

2. initiate an NCS Cooldown B. 1. is

2. try to start an NC pump C. 1. is NOT

2. initiate an NCS Cooldown D. 1. is NOT

2. try to start an NC pump Page 65 of 100

McGuire Nuclear Station Question: 66 ILT-31 MNS SRO NRC Examination (1 point)

Concerning AD-OP-ALL-0203 (REACTIVITY MANAGEMENT) during abnormal operating conditions,

1) a reactor trip should be initiated if the cause of a power change is not understood and reactor power level exceeds the pre-transient power level by greater than a MINIMUM of , or is not controllable.

2) the ROs shall inform the CRS of .

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

A. 1. 5%

2. all MANUAL control rod withdrawals B. 1. 5%

2. the first MANUAL control rod withdrawal ONLY C. 1. 10%

2. all MANUAL control rod withdrawals D. 1. 10%

2. the first MANUAL control rod withdrawal ONLY Page 66 of 100

McGuire Nuclear Station Question: 67 ILT-31 MNS SRO NRC Examination (1 point)

When performing a normal fuel reload from the spent fuel pool to a core location, PT/0/A/4150/033 (TOTAL CORE RELOADING) requires the Reactor Building Crane operator to obtain permission from the to PLACE a fuel assembly into a core location.

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

A. Fuel Handling SRO B. Site Refueling Supervisor C. Fuel Handling Reactor Engineer D. Refueling Booth Support Reactor Operator Page 67 of 100

McGuire Nuclear Station Question: 68 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2:

Unit is at 4% RTP conducting a plant startup Subsequently:

One control bank A rod drops fully into the core

NCS temperature decreases to 550°F Based on the conditions above, the MOST limiting Tech Spec required action is to (1) within (2) .

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

A. 1. restore rod to within alignment limits

2. 30 minutes B. 1. be in MODE 2 with Keff less than 1.0

2. 30 minutes C. 1. restore rod to within alignment limits

2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. 1. be in MODE 2 with Keff less than 1.0

2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Page 68 of 100

McGuire Nuclear Station Question: 69 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial plant conditions:

Both units are operating at 100% RTP

The following alarms are lit on each unit:

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

Subsequently:

An AO reports a loss of BOTH Cooling Groups has occurred on each transformer To prevent a turbine runback to <56% RTP, cooling to the (1) Main Transformer must be restored within a MAXIMUM of (2) .

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

A. 1. 1A

2. 8 minutes, 45 seconds B. 1. 2A

2. 8 minutes, 45 seconds C. 1. 1A

2. 28 minutes, 45 seconds D. 1. 2A

2. 28 minutes, 45 seconds Page 69 of 100

McGuire Nuclear Station Question: 70 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

You are to perform a valve lineup in the Letdown Heat Exchanger Room

The dose rate in the room is 3000 mREM/HR

Your total exposure for the year is 1000 mREM In accordance with PD-RP-ALL-0001 (RADIATION PROTECTION):

The Letdown Heat Exchanger Room must be posted as a (1) Area.

The MAXIMUM amount of time you can spend in the room before reaching your EXCLUDE exposure limit is (2) minutes.

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

A. 1. Locked High Radiation

2. 12 B. 1. Locked High Radiation

2. 16 C. 1. Very High Radiation

2. 12 D. 1. Very High Radiation

2. 16 Page 70 of 100

McGuire Nuclear Station Question: 71 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions:

An RO is performing a valve lineup in the Unit 1 Auxiliary Building pipe chase

The RO receives a Dose Rate alarm on his Electronic Dosimeter (ED)

The possibility of Dose Rate alarms was discussed during the RP brief In accordance with PD-RP-ALL-0001 (RADIATION WORKER RESPONSIBILITIES),

1) the Dose Rate alarm will .

2) based on the conditions above, the RO .

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

A. 1. NOT clear until the ED is reset

2. must stop work and exit the area B. 1. NOT clear until the ED is reset

2. will reset the dose rate alarm and continue to work until two additional dose rate alarms are received C. 1. automatically clear when dose rate drops below 80% of alarm setpoint

2. must stop work and exit the area D. 1. automatically clear when dose rate drops below 80% of alarm setpoint

2. may continue to work until two additional dose rate alarms are received Page 71 of 100

McGuire Nuclear Station Question: 72 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

Mode 4 valve checklist PT is being performed

The PT calls for independent verification of a single valve located in a room with a general dose rate of 110 mREM/hr

Estimated time to independently verify the valve's position is 5 minutes In accordance with NSD-700 (VERIFICATION TECHNIQUES), independent verification of the valve above (1) be waived because (2) .

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

A. 1. can

2. the general area dose rate is greater than 100 mREM/hr B. 1. can NOT

2. the general area dose rate is less than 500 mREM/hr C. 1. can

2. the radiation exposure for a single verification exceeds the allowable limit D. 1. can NOT

2. the radiation exposure for a single verification is within the allowable limit Page 72 of 100

McGuire Nuclear Station Question: 73 ILT-31 MNS SRO NRC Examination (1 point)

Related to Emergency Operating Procedures (EOPs) rules of usage,

1) the step below is a (an) action step.

2) steps which may be performed in any order are designated by .

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

A. 1. immediate

2. asterisks B. 1. immediate

2. bullets C. 1. continuous

2. asterisks D. 1. continuous

2. bullets Page 73 of 100

McGuire Nuclear Station Question: 74 ILT-31 MNS SRO NRC Examination (1 point)

Given the following plant conditions:

A fire has been reported on a small oil cooled transformer

The transformer may be energized Which ONE (1) of the following indicates the fire class ratings of the portable fire extinguishers that must be used in this situation?

A. A and B B. B and C C. C and D D. A and D Page 74 of 100

McGuire Nuclear Station Question: 75 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

A Site Area Emergency has been declared

A Site Assembly is being conducted in accordance with RP/0/A/5700/011 (CONDUCTING A SITE ASSEMBLY, SITE EVACUATION, OR CONTAINMENT EVACUATION)

Per Enclosure 4.3 (OSM ACTIONS FOR SITE ASSEMBLY),

1) the announcement for the Site Assembly shall be repeated every

.minutes until notification that the Site Assembly has been completed.

2) the Site Assembly shall be completed within a MAXIMUM of minutes.

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

A. 1. 10

2. 30 B. 1. 20

2. 30 C. 1. 10

2. 60 D. 1. 20

2. 60 Page 75 of 100

McGuire Nuclear Station Question: 76 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 2:

The unit was at 100% RTP when a LOCA occurs

Safety Injection has been initiated

The Reactor failed to trip automatically

Manual attempts to trip the reactor were unsuccessful Current conditions:

FR-S.1 (RESPONSE TO NUCLEAR GENERATION / ATWS) has been implemented and the crew is at the step to check CA pump status

Containment pressure is 3.1 PSIG and INCREASING

NC system pressure is 1460 PSIG and DECREASING

NC subcooling is (-)1°F

Reactor power is 9% and DECREASING Based on the conditions above, the crew (1) trip the NC pumps because (2) .

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

A. 1. should

2. SI has been initiated and subcooling has been lost B. 1. should NOT

2. immediate actions of FR-S.1 have not been completed C. 1. should NOT

2. reduced NC system heat removal could challenge fuel integrity D. 1. should

2. the normal support systems for running NC pumps are not satisfied Page 76 of 100

McGuire Nuclear Station Question: 77 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

A reactor trip and Safety Injection have occurred

The crew completed E-0 (REACTOR TRIP OR SAFETY INJECTION)

Pzr level is off-scale low

NC System pressure is 1700 PSIG and lowering slowly

Containment pressure is 1.7 PSIG and rising slowly

NC pumps have been secured

SG pressures are 1050 PSIG and stable

CA flow is 600 GPM

FWST level is 135 inches and lowering at 1.0 inches per minute

The operators have just transitioned to E-1 (LOSS OF REACTOR OR SECONDARY COOLANT)

PROCEDURE LEGEND:

ES-1.2 (POST LOCA COOLDOWN AND DEPRESSURIZATION)

ES-1.3 (TRANSFER TO COLD LEG RECIRCULATION)

The basis for stopping NC pumps in E-0 is to (1) .

Based on the conditions above, the next procedure transition will be to (2) .

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

A. 1. minimize heat input into the NC system

2. ES-1.2 B. 1. minimize mass loss from the NC system

2. ES-1.2 C. 1. minimize heat input into the NC system

2. ES-1.3 D. 1. minimize mass loss from the NC system

2. ES-1.3 Page 77 of 100

McGuire Nuclear Station Question: 78 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The unit has experienced a Large Break LOCA

Containment pressure is currently 4.5 PSIG Based on the conditions above, entry into FR-C.1 (RESPONSE TO INADEQUATE CORE COOLING) per the CSF status trees is required if CETs are greater than 700°F and RVLIS LR level is less than or equal to a MINIMUM of (1) .

After S/Gs have been depressurized to atmospheric in FR-C.1, one of the MINIMUM conditions which must be met to allow FR-C.1 to be exited to E-1 (LOSS OF REACTOR OR SECONDARY COOLANT) is (2) .

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

A. 1. 39%

2. NC T-Colds less than 350°F B. 1. 60%

2. NC T-Colds less than 350°F C. 1. 39%

2. CET's less than 1200°F D. 1. 60%

2. CETs less than 1200°F Page 78 of 100

McGuire Nuclear Station Question: 79 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions on Unit 1:

Unit is at 100% RTP

Both operating KC pumps have tripped

AP-21 (LOSS OF KC OR KC SYSTEM LEAKAGE) has been implemented

All NC Pump Motor Bearing LO KC Flow annunciators are LIT

All NC Pump Motor Bearing temperatures are increasing In accordance with the AP-21 Background Document, the NC pumps can be run for a MINUMUM of (1) minutes without KC cooling before any NCP trip criteria will be met.

If any NCP trip criteria are met, AP-08 (MALFUNCTION OF NC PUMP) will direct the crew to trip the reactor and then stop the affected NC pump(s) after reactor power has decreased to less than (2) .

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

A. 1. 10

2. 5%

B. 1. 10

2. 10%

C. 1. 33.5

2. 5%

D. 1. 33.5

2. 10%

Page 79 of 100

McGuire Nuclear Station Question: 80 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial conditions:

A Blackout has occurred on Unit 2

AP-07 (LOSS OF ELECTRICAL POWER) Case I (LOSS OF NORMAL POWER TO BOTH 2ETA OR 2ETB) has been implemented

The 2A D/G is running with the Emergency breaker closed

2B D/G is tripped and SATB is faulted

Pressurizer level is 50%

In accordance with Tech Spec 3.4.9 (PRESSURIZER) basis, Backup Heater Group

'2A' is (1) .

In accordance with Tech Spec 3.4.9, the Pressurizer is (2) .

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

A. 1. OPERABLE

2. OPERABLE B. 1. INOPERABLE

2. OPERABLE C. 1. OPERABLE

2. INOPERABLE D. 1. INOPERABLE

2. INOPERABLE Page 80 of 100

McGuire Nuclear Station Question: 81 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial plant conditions:

A seismic event caused a loss of both switchyards

All four D/Gs started and loaded normally Subsequently:

Flooding occurs in the 1B D/G Room due to a rupture of the RN piping in the room

The rupture is on the KD Heat Exchanger Inlet piping

D/G room sump pumps are running and sump level is RISING

1B D/G Lube Oil Temperature is 180°F and RISING

ES-0.1 (REACTOR TRIP RESPONSE) has been implemented on both units

AP-44 (PLANT FLOODING) is being performed concurrently on Unit 1 PROCEDURE LEGEND:

AP-44 (PLANT FLOODING), Enclosure 7 (AUXILIARY BLDG FLOODING)

OP/1/A/6350/002 (Diesel Generator), Enclosure 4.4 (1B D/G Shutdown)

In accordance with AP-44, Enclosure 7, the 1B D/G will be shut down (1) .

If 1B D/G Lube Oil Temperature exceeds the trip setpoint before it is manually secured by the crew, the diesel (2) trip.

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

A. 1. using the Emergency Stop pushbutton

2. WILL B. 1. in accordance with OP/1/A/6350/002, Enclosure 4.4

2. WILL C. 1. using the Emergency Stop pushbutton

2. WILL NOT D. 1. in accordance with OP/1/A/6350/002, Enclosure 4.4

2. WILL NOT Page 81 of 100

McGuire Nuclear Station Question: 82 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is operating at 100% RTP

Rod control is in AUTOMATIC

Annunciator 2AD-2 / D9 (RPI AT BOTTOM ROD DROP) alarms

DRPI indicates that rod M14 (adjacent to Power Range N-44) has dropped Based on the conditions above, over the next several hours the overall core QPTR will (1) .

To comply with the requirements of Tech Spec 3.1.4 (ROD GROUP ALIGNMENT LIMITS), power must be reduced to less than a MAXIMUM of (2) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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

A. 1. increase

2. 75%

B. 1. increase

2. 50%

C. 1. decrease

2. 75%

D. 1. decrease

2. 50%

Page 82 of 100

McGuire Nuclear Station Question: 83 ILT-31 MNS SRO NRC Examination (1 point)

Loading of a Dry Storage Cask is complete, the cask lid has been welded shut, and the crew is preparing to move the cask to the Vertical Concrete Cask (VCC).

Per Tech Spec 3.7.12 (FUEL HANDLING VENTILATION EXHAUST SYSTEM (FHVES)), the VF System (1) required to be OPERABLE and in operation in the Filter Mode when the cask is moved using the 125 -Ton Overhead Crane with the rollup door closed?

Per Tech Spec 3.7.12 Bases, total system failure could result in the atmospheric release from the fuel handling building exceeding the (2) limits at the site exclusion area boundary in the event of a fuel handling accident.

PROCEDURE LEGEND:

10 CFR PART 20 (STANDARDS FOR PROTECTION AGAINST RADIATION) 10 CFR PART 100 (REACTOR SITE CRITERIA)

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

A. 1. is

2. 10 CFR 20 B. 1. is

2. 10 CFR 100 C. 1. is NOT

2. 10 CFR 20 D. 1. is NOT

2. 10 CFR 100 Page 83 of 100

McGuire Nuclear Station Question: 84 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

A Reactor trip and Safety Injection have occurred due to a SGTR on the 2B S/G

E-3 (STEAM GENERATOR TUBE RUPTURE) has been implemented In accordance with E-3, transition to a contingency procedure (ECA) is required if ruptured S/G pressure cannot be maintained above a MINUMUM of (1) PSIG.

In accordance with the E-3 Background Document, if Steam Generator overfill becomes a concern, the contingency procedure which will be most effective at preventing an overfill condition is (2) .

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

PROCEDURE LEGEND:

ECA-3.1 (SGTR WITH LOSS OF REACTOR COOLANT - SUBCOOLED RECOVERY DESIRED)

ECA-3.2 (SGTR WITH LOSS OF REACTOR COOLANT - SATURATED RECOVERY DESIRED)

A. 1. 280

2. ECA-3.2 B. 1. 300

2. ECA-3.2 C. 1. 280

2. ECA-3.1 D. 1. 300

2. ECA-3.1 Page 84 of 100

McGuire Nuclear Station Question: 85 ILT-31 MNS SRO NRC Examination (1 point)

Given the following plant conditions:

Both units are in MODE 1 at 100% RTP

Welding activities in the Control Room ceiling has dropped hot slag onto the control boards and started a fire (Active)

The fire has generated toxic gas, rendering the Control Room uninhabitable

AP-45 (PLANT FIRE) has been implemented

A plant cooldown is required due to the extent of the fire damage AP-45 is designed to protect the (1) for the pending shutdown/cooldown of the plant.

The next procedure the CRS will implement as directed by AP-45 is (2) .

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

PROCEDURE LEGEND:

AP-17 (LOSS OF CONTROL ROOM)

AP-24 (LOSS OF PLANT CONTROL DUE TO FIRE OR SABOTAGE)

A. 1. TD CA pump

2. AP-17 B. 1. Standby Makeup pump

2. AP-17 C. 1. TD CA pump

2. AP-24 D. 1. Standby Makeup pump

2. AP-24 Page 85 of 100

McGuire Nuclear Station Question: 86 ILT-31 MNS SRO NRC Examination (1 point)

Given the following sequence of events on Unit 2:

04/05/15 @ 0100 - Reactor trip breaker opened per shutdown procedure in preparation for Refueling Outage

04/08/15 @ 0100 - Unit enters MODE 5

04/13/15 @ 0100 - NC system is open to atmosphere Current conditions 04/14/15 @ 0115:

AP-19 (LOSS OF ND OR ND SYSTEM LEAKAGE) implemented after running ND pump (2B) trips on overcurrent

NC system T-colds = 182°F

NC system temperature is increasing at 1°F/min

2A NI pump is available to start

Attempts to start the 2A ND pump have been unsuccessful Based on the conditions above, the crew (1) required to immediately initiate NC system feed and bleed.

If NC system feed and bleed is initiated, the MINIMUM amount of required makeup flow is (2) GPM. (Assume NC system conditions are approaching saturation)

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

REFERENCE PROVIDED A. 1. is

2. 600 B. 1. is

2. 660 C. 1. is NOT

2. 600 D. 1. is NOT

2. 660 Page 86 of 100

McGuire Nuclear Station Question: 87 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

A LOCA has occurred inside Containment

E-1 (LOSS OR REACTOR OR SECONDARY COOLANT) has been implemented

H2 Analyzers are in service

The Hydrogen Igniters are NOT in service

The NF AHUs (ICE CONDENSER AIR HANDLING UNITS) are OFF

Containment H2 concentration is 4.5%

In accordance with the E-1 Background Document, the reason for stopping the NF AHUs is because (1) .

Based on the conditions above, the crew (2) place the H2 Igniters in service.

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

A. 1. operation of the NF AHUs results in erroneous H2 Analyzer indication

2. will B. 1. operation of the NF AHUs was not included in the Containment H2 combustion analysis

2. will C. 1. operation of the NF AHUs results in erroneous H2 Analyzer indication

2. will NOT D. 1. operation of the NF AHUs was not included in the Containment H2 combustion analysis

2. will NOT Page 87 of 100

McGuire Nuclear Station Question: 88 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit was manually tripped from 100% RTP due to a loss of both CF pumps

At Step 7 of E-0 (REACTOR TRIP OR SAFETY INJECTION) the OATC reports that the Feedwater Isolation status lights on 1SI-4 are NOT LIT

Narrow Range (NR) levels in ALL S/Gs are 78% and INCREASING Based on the conditions above, a YELLOW PATH on Heat Sink will occur if NR level in the intact S/Gs reaches a MINIMUM of (1) .

In accordance with the Background Document for FR-H.3 (RESPONSE TO STEAM GENERATOR HIGH LEVEL), the reason for preventing S/G levels from going above the NR level span is to (2) ..

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

A. 1. 83%

2. prevent an NC system overcooling event B. 1. 83%

2. ensure each S/G remains effective for secondary heat removal C. 1. 92%

2. prevent an NC system overcooling event D. 1. 92%

2. ensure each S/G remains effective for secondary heat removal Page 88 of 100

McGuire Nuclear Station Question: 89 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

Unit is at 100% RTP

OAC Alarm M1A1590 (1A1 VG HEADER PRESS LO-LO) in alarm

Annunciator 1AD-11 / A6 (DG A PANEL TROUBLE) in alarm An AO dispatched to the 1A DG Room reports the following:

Annunciator 1AD-19 / C5 (STARTING AIR PRESS LOW) is in alarm

No other annunciators are lit at the local panel

VG Starting Air Tank 1A1 is 208 PSIG and going down slowly

VG Starting Air Tank 1A2 is 220 PSIG and stable

VG compressor 1A1 cannot be started

1A1 VG Starting Air Tank has been isolated Based on the conditions above, the 1A Emergency Diesel Generator is (1) .

Per Tech Spec 3.8.3 (DIESEL FUEL OIL AND STARTING AIR) Bases, for the Starting Air system to be considered OPERABLE it must be capable of a MINIMUM of (2) .start attempts without recharging the air start receivers.

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

1. OPERABLE A.

2. three

1. INOPERABLE B.

2. three

1. OPERABLE C.

2. five

1. INOPERABLE D.

2. five Page 89 of 100

McGuire Nuclear Station Question: 90 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is at 100% RTP

A failure of the power supply for 2EMF-38L (CONTAINMENT PARTICULATE MONITOR) occurs Based on the conditions above, a Trip 2 alarm (1) be LIT on 2EMF-38L.

In accordance with Tech Spec 3.4.15 (RCS LEAKAGE DETECTION INSTRUMENTATION) Basis, (2) is capable of identifying a 1 GPM Reactor Coolant system leak in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or less after leakage has reached the sumps.

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

A. 1. will

2. Incore Instrument sump level B. 1. will NOT

2. Incore Instrument sump level C. 1. will

2. Containment Floor and Equipment (CFAE) sump level D. 1. will NOT

2. Containment Floor and Equipment (CFAE) sump level Page 90 of 100

McGuire Nuclear Station Question: 91 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The unit is in MODE 4 following a Refueling Outage

OP/1/A/6100/SU-13 (HEATUP TO 350°F) is in effect

Two CRD Vent (VR) fans are currently in service Subsequently:

One VR fans trips In accordance with OP/1/A/6100/SU-15 (MODE 3 CHECKLIST), the crew must start (1) of the remaining available VR fans to enter MODE 3.

In accordance with the design basis, the VR system is designed to maintain temperature exiting the CRDM shroud equal to or less than a MAXIMUM of (2) .

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

A. 1. one

2. 100°F B. 1. one

2. 150°F C. 1. both

2. 100°F D. 1. both

2. 150°F Page 91 of 100

McGuire Nuclear Station Question: 92 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

The unit has experienced a Safety Injection

All S/G pressures are 1000 PSIG and STABLE

NC system pressure is 1550 PSIG and STABLE

Containment pressure is 1.3 PSIG and LOWERING SLOWLY

The crew has implemented E-3 (STEAM GENERATOR TUBE RUPTURE) due to 1EMF-24 (STEAM LINE 1A MONITOR) Trip 2 light being LIT

The BOP informs the CRS that 1RAD-3 / F5 (CABINET TROUBLE) is LIT

The CRS believes the 1EMF-24 detector may have failed

1. In accordance with E-3, to verify the validity of the EMF indication, the crew will

2. Once the indication is determined to be false, the NEXT procedure to which the crew will transition is .

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

PROCEDURE LEGEND:

E-0 (REACTOR TRIP OR SAFETY INJECTION)

E-1 (LOSS OR REACTOR OR SECONDARY COOLANT)

A. 1. verify the Trip 2 alarm on 1EMF-71 (STEAM LINE 1A N16 MONITOR) is DARK

2. E-0 B. 1. verify the Trip 2 alarm on 1EMF-71 (STEAM LINE 1A N16 MONITOR) is DARK

2. E-1 C. 1. request that RP frisk cation columns

2. E-0 D. 1. request that RP frisk cation columns

2. E-1 Page 92 of 100

McGuire Nuclear Station Question: 93 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is at 100% RTP

AOs are performing the SSF D/G Operability Test in accordance with PT/0/A/4200/002 (STANDBY SHUTDOWN FACILITY OPERABILITY TEST)

The following sequence of events occurs:

0200 - A fuel oil leak occurs on the SSF D/G 0203 - A fire breaks out on the SSF D/G 0205 - The SSF D/G is secured 0220 - The fire brigade extinguished the fire 0222 - It is determined that damage to the SSF D/G has occurred because the SSF Sprinkler system failed to actuate rendering the SSF D/G non-functional Based on the conditions above,

1) the Emergency Classification for this event would be an .

2) in accordance with SLC 16.9.7 STANDBY SHUTDOWN SYSTEM), the crew required to immediately notify Security that the SSF D/G is non-functional.

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

REFERENCE PROVIDED A. 1. Unusual Event

2. is B. 1. Alert

2. is C. 1. Unusual Event

2. is NOT D. 1. Alert

2. is NOT Page 93 of 100

McGuire Nuclear Station Question: 94 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is at 100% RTP

Engineering has requested access to the Containment Annulus area for an inspection In accordance with MSD-585 (REACTOR BUILDING PERSONNEL ACCESS AND MATERIAL CONTROL),

1) the MINIMUM level of approval for the Annulus entry is the .

2) the use of the "buddy system" .required for entry into the Annulus.

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

A. 1. WCC SRO

2. is NOT B. 1. WCC SRO

2. is C. 1. Shift Manager

2. is NOT D. 1. Shift Manager

2. is Page 94 of 100

McGuire Nuclear Station Question: 95 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is in MODE 6 for a Refueling Outage In accordance with AD-NS-ALL-1001 (CONDUCT OF REFUELING),

1) bypassing fuel handling interlocks not specified in approved procedures requires permission of the (1) and concurrence of the Shift Manager.

2) the Refueling SRO (2) required to be present inside Containment for control rod latching.

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

A. 1. Refueling SRO

2. is B. 1. Refueling SRO

2. is NOT C. 1. Duty Reactor Engineer

2. is D. 1. Duty Reactor Engineer

2. is NOT Page 95 of 100

McGuire Nuclear Station Question: 96 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 2:

The unit is in MODE 6 with core RELOAD in progress

NC system boron concentration is 2705 PPM

Boric Acid Tank level is 10%

The following surveillances are being performed:

o PT/2/A/4600/100 (SURVEILLANCE REQUIREMENTS FOR SHUTDOWN CONDITIONS) o PT/2/A/4600/003 C (WEEKLY SURVEILLANCE ITEMS CHECKLIST)

The surveillance for NC system boron concentration performed during PT/2/A/4600/100 (SR 3.9.1.1) ensures that keff during MODE 6 remains less than or equal to a MAXIMUM of (1) .

Based on the conditions above the MINIMUM required Boric Acid Tank level (2) met.

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

REFERENCE PROVIDED A. 1. 0.95

2. is B. 1. 0.95

2. is NOT C. 1. 0.98

2. is D. 1. 0.98

2. is NOT Page 96 of 100

McGuire Nuclear Station Question: 97 ILT-31 MNS SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

A SGTR has occurred

E-3 (STEAM GENERATOR TUBE RUPTURE) is in progress

After terminating Safety Injection, the following indications are observed:

o NC system pressure is DECREASING o Pressurizer level is 10% and DECREASING o Containment temperature, pressure, and radiation levels are INCREASING PROCEDURE LEGEND:

ES-0.0 (RE-DIAGNOSIS)

E-0 (REACTOR TRIP OR SAFETY INJECTION)

E-1 (LOSS OR REACTOR OR SECONDARY COOLANT)

ECA-3.1 (SGTR WITH LOSS OF REACTOR COOLANT - SUBCOOLED RECOVERY DESIRED)

Based on the conditions above, the CRS will direct the crew to:

A. initiate Safety Injection and go to E-0.

B. initiate Safety Injection and go to ES-0.0.

C. manually start ECCS pumps as necessary and transition to E-1.

D. manually start ECCS pumps as necessary and transition to ECA-3.1.

Page 97 of 100

McGuire Nuclear Station Question: 98 ILT-31 MNS SRO NRC Examination (1 point)

Given the following plant conditions:

A release of Waste Monitor Tank (WMT) A has been planned

A radioactive liquid release permit has been prepared with the following data:

RC PUMP DATA ==========================================

RC pumps running................................................................... 3.00 RC pumps assigned to RELEASE........................................... 3.00 Total RC pumps required (all concurrent releases)................. 4.00

RECOMMENDED RELEASE RATE ============================

Allowable release rate (gpm)................................................. 1.61E+01 Recommended release rate (gpm)........................................ 1.20E+02

SETPOINT DATA =========================================

EMF49L in Service ................ Yes Monitor Background (cpm)...... 4.49E+03 Cs-137 Equivalence (uCi/ML)... 7.23E-06 Expected CPM.................. 4.50E+03 Trip 1 setpoint (cpm)............... 8.97E+03 Trip 2 setpoint (cpm)............... 1.42E+04 Which ONE (1) of the following indicates the actions related to the approval of this release permit?

A. The release may NOT be approved due to inadequate number of RC pumps.

Recommended Release Rate information is correct.

B. The release may NOT be approved due to incorrect Recommended Release Rate. RC Pump Data information is adequate.

C. The release may NOT be approved due to inadequate number of RC Pumps AND incorrect Recommended Release Rate.

D. The release MAY be approved as presented.

Page 98 of 100

Question 99 (page 99 of 100) deleted and intentionally left blank McGuire Nuclear Station Question: 100 ILT-31 MNS SRO NRC Examination (1 point)

Given the following initial plant conditions:

Both units are at 100% RTP Subsequently, the following are observed:

The operating crew feels a tremor in the Control Room

1AD-13 / E7 (O.B.E EXCEEDED) comes into alarm

The crew has implemented RP/0/A/5700/007 (EARTHQUAKE)

2EMF-3 (CONTAINMENT REFUELING BRIDGE) is in Trip 2

1EMF-17 (UNIT 1 SPENT FUEL REFUELING BRIDGE) is in Trip 1 indicating 200 mr/hr Based on the conditions above,

1) in accordance with RP-07, the crew required to swap both trains of RN to the SNSWP using AP-20 (LOSS OF RN) Case II (LOSS OF LOW LEVEL OR RC CROSSOVER).

2) in accordance with RP-000 (EMERGENCY CLASSIFICATION), the classification for this event is a/an .

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

REFERENCE PROVIDED A. 1. is

2. Notification of Unusual Event B. 1. is NOT

2. Notification of Unusual Event C. 1. is

2. Alert D. 1. is NOT

2. Alert Page 100 of 100

Reference List for: ILT-31 MNS SRO NRC Examination Steam Tables Unit 1 Plant Data Book, Curve 7.2 Tech Spec 3.3.1 (RTS Instrumentation) Pages 1-15 Data Book Table 2.10.4 Unit 2 COLR - Borated Water Sources RP-000, Classification of Emergency Printed 4/24/2015 12:09:41 PM

UNIT 1 OP/1/A/6100/22 ENCLOSURE 4.3 CURVE 7.2 PRESSURIZER RELIEF TANK (VOLUME vs. LEVEL) 14000 13000 100% LEVEL = 13,172 GALLONS 12000 11000 10000 9000 VOLUME (GALLONS) 8000 7000 6000 TOTAL VOLUME OF TANK = 13,488 GALLONS 5000 4000 3000 2000 1000 0% LEVEL = 317 GALLONS 0

0 10 20 30 40 50 60 70 80 90 100 110 PERCENT LEVEL INSTRUMENTATION INDICATION This data is also available on the OAC. UNIT 1

RTS Instrumentation 3.3.1 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.3.1 The RTS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1-1.

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Enter the Condition Immediately with one or more referenced in Table 3.3.1-1 required channels for the channel(s).

inoperable.

B. One Manual Reactor B.1 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Trip channel inoperable. OPERABLE status.

OR B.2 Be in MODE 3. 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months C. One channel or train C.1 Restore channel or train to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months inoperable. OPERABLE status.

OR C.2 Open reactor trip breakers 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months (RTBs).

(continued)

McGuire Units 1 and 2 3.3.1-1 Amendment Nos. 184/166

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One channel inoperable. ------------------NOTE-------------------

One channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing and setpoint adjustment.

D.1.1 ------------NOTE---------------

Only required to be performed when the Power Range Neutron Flux input to QPTR is inoperable Perform SR 3.2.4.2 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from discovery of THERMAL POWER

> 75% RTP AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND D.1.2 Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR D.2 Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months (continued)

McGuire Units 1 and 2 3.3.1-2 Amendment Nos. 248/228

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One channel inoperable. ------------------NOTE-------------------

One channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

E.1 Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR E.2 Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months F. THERMAL POWER F.1 Reduce THERMAL 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

> P-6 and < P-10, one POWER to < P-6.

Intermediate Range Neutron Flux channel OR inoperable.

F.2 Increase THERMAL 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months POWER to > P-10.

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

Limited boron concentration changes associated with RCS inventory control or limited plant temperature changes are allowed.

G. THERMAL POWER G.1 Suspend operations Immediately

> P-6 and < P-10, two involving positive reactivity Intermediate Range additions.

Neutron Flux channels inoperable. AND G.2 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to < P-6.

H. THERMAL POWER H.1 Restore channel(s) to Prior to increasing

< P-6, one or two OPERABLE status. THERMAL POWER Intermediate Range to > P-6 Neutron Flux channels inoperable.

(continued)

McGuire Units 1 and 2 3.3.1-3 Amendment Nos. 248/228

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

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

Limited boron concentration changes associated with RCS inventory control or limited plant temperature changes are allowed.

I. One Source Range I.1 Suspend operations Immediately Neutron Flux channel involving positive reactivity inoperable. additions.

J. Two Source Range J.1 Open RTBs. Immediately Neutron Flux channels inoperable.

K. One Source Range K.1 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Neutron Flux channel OPERABLE status.

inoperable.

OR K.2 Open RTBs. 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months

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

Plant temperature changes are allowed provided that SDM is maintained and Keff remains <

0.99.

L. Required Source Range L.1 Suspend operations Immediately Neutron Flux channel involving positive reactivity inoperable. additions.

AND L.2 Close unborated water 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months source isolation valves.

AND L.3 Perform SR 3.1.1.1. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter (continued)

McGuire Units 1 and 2 3.3.1-4 Amendment Nos. 216 / 197

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME M. One channel inoperable. ------------------NOTE-------------------

One channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

M.1 Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR M.2 Reduce THERMAL 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months POWER to < P-7.

N. One Reactor Coolant -----------------NOTE--------------------

Flow - Low (Single One channel may be bypassed for Loop) channel up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance inoperable. testing.

N.1 Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR N.2 Reduce THERMAL 76 hours8.796296e-4 days 0.0211 hours 1.256614e-4 weeks 2.8918e-5 months POWER to < P-8.

(continued)

McGuire Units 1 and 2 3.3.1-5 Amendment Nos. 250/230

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME O. One Turbine Trip - Low ------------------NOTE-------------------

Fluid Oil Pressure One channel may be bypassed for channel inoperable. up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

O.1 Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR O.2 Reduce THERMAL 76 hours8.796296e-4 days 0.0211 hours 1.256614e-4 weeks 2.8918e-5 months POWER to < P-8.

P. One or more Turbine P.1 Place channel(s) in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Trip - Turbine Stop Valve Closure channels OR inoperable.

P.2 Reduce THERMAL 76 hours8.796296e-4 days 0.0211 hours 1.256614e-4 weeks 2.8918e-5 months POWER to < P-8.

Q. One train inoperable. ------------------NOTE-------------------

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing provided the other train is OPERABLE.

Q.1 Restore train to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

OR Q.2 Be in MODE 3. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months (continued)

McGuire Units 1 and 2 3.3.1-6 Amendment Nos. 248/228

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME R. One RTB train ------------------NOTE------------------

inoperable. One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing, provided the other train is OPERABLE.

R.1 Restore train to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

OR R.2 Be in MODE 3. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months S. One or more channel(s) S.1 Verify interlock is in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. required state for existing unit conditions.

OR S.2 Be in MODE 3. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (continued)

McGuire Units 1 and 2 3.3.1-7 Amendment Nos. 248/228

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME T. One or more channel(s) T.1 Verify interlock is in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. required state for existing unit conditions.

OR T.2 Be in MODE 2. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months U. One trip mechanism U.1 Restore inoperable trip 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months inoperable for one RTB. mechanism to OPERABLE status.

OR U.2 Be in MODE 3. 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months V. Two RTS trains V.1 Enter LCO 3.0.3. Immediately inoperable.

McGuire Units 1 and 2 3.3.1-8 Amendment Nos. 184/166

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 ------------------------------NOTES----------------------------------

1. Adjust NIS channel if absolute difference is > 2%

RTP.

2. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is > 15% RTP.

Compare results of calorimetric heat balance calculation In accordance with to Nuclear Instrumentation System (NIS) channel output. the Surveillance Frequency Control Program SR 3.3.1.3 ------------------------------NOTES----------------------------------

1. Adjust NIS channel if absolute difference is > 3%

AFD.

2. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is > 15% RTP.

Compare results of the incore detector measurem In accordance with ents to NIS AFD. the Surveillance Frequency Control Program (continued)

McGuire Units 1 and 2 3.3.1-9 Amendment Nos. 261/241

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.4 ------------------------------NOTES----------------------------------

This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.1.5 Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.6 ------------------------------NOTES----------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is > 75% RTP.

Calibrate excore channels to agree with incore detector In accordance with measurements. the Surveillance Frequency Control Program SR 3.3.1.7 ------------------------------NOTES----------------------------------

Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3.

Perform COT. In accordance with the Surveillance Frequency Control Program (continued)

McGuire Units 1 and 2 3.3.1-10 Amendment Nos. 261/241

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.8 ------------------------------NOTES----------------------------------

This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions.

Perform COT. ---------NOTE-------

Only required when not performed within the Frequency specified in the Surveillance Frequency Control Program or previous 184 days Prior to reactor startup AND Four hours after reducing power below P-10 for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND In accordance with the Surveillance Frequency Control Program (continued)

McGuire Units 1 and 2 3.3.1-11 Amendment Nos. 261/241

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.9 ------------------------------NOTES----------------------------------

Verification of setpoint is not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.1.10 ------------------------------NOTES----------------------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.11 ------------------------------NOTES----------------------------------

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

2. Power Range Neutron Flux high voltage detector saturation curve verification is not required to be performed prior to entry into MODE 1 or 2.

3. Intermediate Range Neutron Flux detector plateau voltage verification is not required to be performed prior to entry into MODE 1 or 2.* In accordance with

the Surveillance Frequency Control Perform CHANNEL CALIBRATION. Program SR 3.3.1.12 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program (continued)

This note applies to the Westinghouse-supplied compensated ion chamber neutron detectors. The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors which do not require detector plateau voltage verification. Therefore, this note does not apply to the fission chamber neutron detectors.

McGuire Units 1 and 2 3.3.1-12 Amendment Nos. 261/241

RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.13 Perform COT. In accordance with the Surveillance Frequency Control Program SR 3.3.1.14 ------------------------------NOTES----------------------------------

Verification of setpoint is not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.1.15 ------------------------------NOTES---------------------------------- --------NOTE--------

Verification of setpoint is not required. Only required

when not performed within previous 31 days Perform TADOT. Prior to reactor startup SR 3.3.1.16 ------------------------------NOTES----------------------------------

Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits. In accordance with the Surveillance Frequency Control Program SR 3.3.1.17 Verify RTS RESPONSE TIME for RTDs is within limits. In accordance with the Surveillance Frequency Control Program McGuire Units 1 and 2 3.3.1-13 Amendment Nos. 261/241

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 7)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Manual Reactor Trip 1,2 2 B SR 3.3.1.14 NA NA 3(a), 4(a), 5(a) 2 C SR 3.3.1.14 NA NA

2. Power Range Neutron Flux

a. High 1,2 4 D SR 3.3.1.1 < 110% RTP 109% RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16

b. Low 1(b),2 4 E SR 3.3.1.1 < 26% RTP 25% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16

3. Power Range Neutron Flux Rate High Positive Rate 1,2 4 D SR 3.3.1.7 < 5.5% RTP 5% RTP SR 3.3.1.11 with time with time constant constant

> 2 sec > 2 sec

4. Intermediate Range 1(b), 2(c) 2 F,G SR 3.3.1.1 < 30% RTP* 25% RTP Neutron Flux SR 3.3.1.8(j)(k) < 38% RTP SR 3.3.1.11(j)(k) 2(d) 2 H SR 3.3.1.1 < 30% RTP*

25% RTP

< 38% RTP SR 3.3.1.8(j)(k)

SR 3.3.1.11(j)(k)

(continued)

The < 30% RTP Allowable Value applies to the Westinghouse-supplied compensated ion chamber Intermediate Range neutron detectors. The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors. The < 38% Allowable Value applies to the replacement fission chamber Intermediate Range neutron detectors.

(a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b) Below the P-10 (Power Range Neutron Flux) interlocks.

(c) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(j) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(k) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in the UFSAR.

McGuire Units 1 and 2 3.3.1-14 Amendment Nos. 257/237

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 7)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

5. Source Range 2(d) 2 I,J SR 3.3.1.1 < 1.3 E5 cps** 1.0 E5 cps Neutron Flux SR 3.3.1.8(j)(k) < 1.44 E5 cps SR 3.3.1.11(j)(k) 3(a), 4(a), 5(a) 2 J,K SR 3.3.1.1 < 1.3 E5 cps** 1.0 E5 cps SR 3.3.1.7(j)(k) < 1.44 E5 cps SR 3.3.1.11(j)(k) 3(e), 4(e), 5(e) 1 L SR 3.3.1.1 N/A N/A SR 3.3.1.11

6. Overtemperature T 1,2 4 E SR 3.3.1.1 Refer to Note 1 Refer to SR 3.3.1.3 (Page Note 1 (Page SR 3.3.1.6 3.3.1-18) 3.3.1-18)

SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.17

7. Overpower T 1,2 4 E SR 3.3.1.1 Refer to Note 2 Refer to SR 3.3.1.3 (Page Note 2 (Page SR 3.3.1.6 3.3.1-19) 3.3.1-19)

SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.17

8. Pressurizer Pressure

a. Low 1(f) 4 M SR 3.3.1.1 > 1935 psig 1945 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. High 1,2 4 E SR 3.3.1.1 < 2395 psig 2385 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 (continued)

- The < 1.3 E5 cps Allowable Value applies to the Westinghouse-supplied boron triflouride (BF3) Source Range neutron detectors. The BF3 neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors. The < 1.44 E5 cps Allowable Value applies to the replacement fission chamber Source Range neutron detectors.

(a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(e) With the RTBs open. In this condition, source range Function does not provide reactor trip but does provide indication.

(f) Above the P-7 (Low Power Reactor Trips Block) interlock.

(j) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(k) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in the UFSAR.

McGuire Units 1 and 2 3.3.1-15 Amendment Nos. 257/237

UNIT 2 Page 1 of 2 OP/2/A/6100/022 ENCLOSURE 4.3 SECTION 2.10.4 UNIT 2

UNIT 2 Page 2 of 2 OP/2/A/6100/022 ENCLOSURE 4.3 SECTION 2.10.4 Detailed Discussion of Development and Application of Section 2.10.4 Section 2.10.4 contains a graph of Core Flow Required to Prevent Boiling for Loss of Decay Heat Removal. The supporting calculations for this graph are in DPC-1552.08 0143, Upper Internals Spray Nozzle Flow Capability During A Loss of Decay Heat Removal Event. Detailed calculations to determine the point at which required core flow to prevent boiling can be met by the circulation available through the upper internal nozzles have not been developed.

Therefore, the thermal margin values in 2.10.2 are non-conservative when the upper internals are installed, and thus do not apply. For conservatism, anytime upper internals are installed, no credit can be taken for water level greater than 84, and the appropriate curves from Section 2.10.1 should be used.

The results presented in the thermal margin curves for section 2.10.1 are not affected by this analysis. However, the presence of the upper internals coupled with the heat loads at times near shutdown may lead to localized core voiding. If decay heat removal capability (forced flow) were lost, the limited flow past the upper internals nozzles (with the normal configuration of a vented RCS) could lead to rapid voiding in the top of the core even with water level above the upper internals. In addition, the core flow requirements for mitigation of core boiling are based on forced flow required to dissipate decay heat with no allowances for recovery of core volume. The flow requirement presented in the graph are valid for mitigation of boiling for a pre-refueling core. This information should be used in conjunction with 2.10.1 and 2.10.2 to ensure appropriate contingencies are planned to mitigate core boiling in the event of a loss of decay heat removal capability.

For AP/EP use:

1. This curve specifically addresses the condition of a vented reactor coolant system.

2. The makeup flow to the NC system is assumed to be at 140 °F.

3. To ensure adequate makeup flow for inlet temperatures approaching saturation, multiply the flow requirements of this graph by 1.10.

4. This curve assumes atmospheric pressure, any application for RCS pressures > 14.7 psig could result in a non-conservative flowrate.

5. SAMG Calculation Aids requirements for makeup flow may also be helpful.

References:

1. DPC-1552.08-00-0143, Upper Internals Spray Nozzle Flow Capability During A Loss of Decay Heat Removal Event

2. WOG DW-95-23 UNIT 2

Duke Energy Procedure No.

McGuire Nuclear Station 0 RP/ /A/5700/000 Classification Of Emergency Revision No.

021 Electronic Reference No.

MC0048M3 Reference Use PERFORMANCE

* * * * * * * *

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

(ISSUED) - PDF Format

RP/0/A/5700/000 Page 2 of 5 Revision History (significant issues, limited to one page)

Rev 021 (1/16/14) Enclosure 4.2 section 4.2.U.3-1 changed operating mode from 1,2,3 to "1,2,3,4,5" and deleted "Mode 3 with tavg greater than 500 deg F" to make EAL comply with original SER.

Rev 020 (3/5/13) added Note prior to 4.2.U.4-1 Interconnected system leakage (ie: letdown, RHR) that can be easily detected and readily isolated is not included in this IC. (Section D of the Emergency Plan).

Section 4.8 changed "takes HOSTAGES" to "take HOSTAGES".

RP/0/A/5700/000 Page 3 of 5 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 result in exposure levels which 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 Guideline exposure levels offsite for more than the immediate site area.

RP/0/A/5700/000 Page 4 of 5

2. Immediate Actions 2.1 Assessment, classification and declaration of any applicable emergency condition should be completed within 15 minutes after the availability to indications or information to cognizant facility staff that an EAL threshold has been exceeded. (Refer to enclosure 4.9, Emergency Declaration Guidelines, as needed.)

_____ 2.2 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.

_____ 2.3 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).

_____ 2.4 IF a General Emergency is NOT declared in Step 2.3, OR the condition does not affect fission product barriers, THEN review the listing of enclosures to determine if the event is applicable to one of the categories shown.

_____ 2.5 Compare actual plant conditions to the Emergency Action Levels evaluated in 2.3 and/or 2.4, then declare the appropriate Emergency Class as indicated.

2.5.1 Event Declaration time _______________________.

_____ 2.6 Implement the applicable Emergency Response Procedure (RP) for that classification and continue with subsequent steps of this procedure.

Notification of Unusual Event RP/0/A/5700/001 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.

RP/0/A/5700/000 Page 5 of 5 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

Enclosure 4.1 RP/0/A/5700/000 Fission Product Barrier Matrix Page 1 of 5 Use EALs to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all 3 barriers. Classify according to the table on page 2 of 5 of this enclosure.

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

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

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

Note 4: Critical Safety Function (CSF) indications are not meant to include transient alarm conditions which may appear during the start-up of engineered safeguards equipment. A CSF condition is satisfied when the alarmed state is valid and sustained. The STA should be consulted to affirm if any CSF has been validated prior to that CSF being used as the basis to classify an emergency. {1} Example: If ECA-0.0, 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-0.0.

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 4.1.S.1 Loss OR Potential Loss 4.1.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 4.1.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 4.1.A.3 Potential Loss of 4.1.S.3 Loss of Containment Containment AND AND Loss OR Potential Loss Loss OR Potential Loss of Any Other Barrier of Any Other Barrier

Enclosure 4.1 RP/0/A/5700/000 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 4-6 Alert Fuel Clad 4 5 7 - 10 Site Area Emergency Total Points 11 - 13 General Emergency

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

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

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

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

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

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

7. Using that EAL number (e.g. 4.1.A.1) select the preprinted notification form OR a blank form and complete the required information for Emergency Coordinator/EOF Director approval and transmittal.

Enclosure 4.1 RP/0/A/5700/000 Fission Product Barrier Matrix Page 3 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)

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 µCi/cc Dose pressure following charging pump in indicated by a loss Equivalent Iodine

H2 concentration initial increase. the normal of NCS subcooling. (DEI) I-131.

> 9%. charging mode

Containment with letdown

Containment pressure or sump isolated.

pressure greater than level response not 3 psig with either a consistent with failure of both trains LOCA conditions.

of NS OR failure of both trains of VX-CARF.

CONTINUED CONTINUED CONTINUED

Enclosure 4.1 RP/0/A/5700/000 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.

Containment

Primary-to-

Indication that a

Not applicable.

Containment isolation is Secondary leak SG is Ruptured and radiation monitor incomplete and a rate exceeds the has a Non-Isolable EMF 51 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 R/hr with letdown SG is ruptured and 0.5-2 hrs > 43 R/hr isolated. a prolonged release 2-4 hrs > 31 R/hr of contaminated 4-8 hrs > 22 R/hr secondary coolant >8 hrs > 13 R/hr is occurring from the affected SG to the environment.

4. SG Secondary Side Release With Primary-to- 4. Containment Radiation Monitoring 4. Emergency Coordinator/EOF Director Secondary Leakage Judgement

Not applicable.

Release of

Not applicable.

Not applicable.

Any condition, including inability to monitor secondary 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

Enclosure 4.1 RP/0/A/5700/000 Fission Product Barrier Matrix Page 5 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)

5. Significant Radioactive Inventory In 5. Emergency Coordinator/EOF Director Containment Judgement

Containment Rad.

Not applicable.

Any condition, including inability to monitor Monitor EMF51A 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.

> 390 R/hr @

0 - 0.5 hr

> 170 R/hr @

0.5 - 2 hr END

> 125 R/hr @

2 - 4 hr

> 90 R/hr @

4 - 8 hr

> 53 R/hr @

> 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 RP/0/A/5700/000 System Malfunctions Page 1 of 2 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 END Shutdown Within Technical Safety System Annunciation or Significant Transient in Specification Limits. Indication in Control Room Progress.

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

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

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

Time.

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

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

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

4.2.U.2-1 The following conditions exist: In the opinion of the Operations Shift 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 RP/0/A/5700/000 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, 5 4.2.U.3-1 Dose Equivalent I-131 greater than the Technical Specification allowable limit.

4.2.U.4 Reactor Coolant System (NCS)

Leakage.

OPERATING MODE: 1, 2, 3, 4 NOTE: Interconnected system leakage (ie:

letdown, RHR) that can be easily detected and readily isolated is not included in this IC. (Section D of the Emergency Plan).

4.2.U.4-1 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 RP/0/A/5700/000 Abnormal Rad Levels/Radiological Effluent Page 1 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.U.1 Any Unplanned Release of Gaseous or 4.3.A.1 Any Unplanned Release of 4.3.S.1 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 AND the required assessments reading is sustained for the time of the Release.

(procedure calculations) cannot be period indicated in the EAL OPERATING MODE: ALL completed within this time period, AND the required assessments OPERATING MODE: ALL declaration must be made based on the (procedure calculations) cannot Note 1: These EMF readings are valid radiation monitor reading. be completed within this time calculated based on Note 1: These EMF readings are period, declaration must be made average annual calculated based on 4.3.U.1-1 A valid indication on radiation monitor based on the valid radiation meteorology, site average annual EMF- 49L, EMF-44L or EMF-31 monitor reading. boundary dose rate, and meteorology, site (when aligned to RC) of design unit vent flow rate. boundary dose rate, and 5.45E+06 cpm for 60 minutes or will 4.3.A.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 of 1. 56 E + 03 cpm for actual meteorology, dose assessment team 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/1009/010, indicates that the release may these EMF readings vent flow rate.

HP/0/B/1009/029, or SH/0/B/2005/001. have exceeded the initiating should not be used if dose Therefore, these EMF condition and indicates the need assessment team readings should not be to assess the release with calculations are available. used if dose assessment (Continued) procedure HP/0/B/1009/010, team calculations are HP/0/B/1009/029, or (Continued) available.

SH/0/B/2005/001.

(Continued)

Enclosure 4.3 RP/0/A/5700/000 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 team EMF- 36L of 2.05E+04 cpm for radiation monitor EMF- 36L calculations cannot be calculations cannot be

> 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/1009/010, HP/0/B/1009/029, or condition and indicates the need 4.3.S.1-1 A valid indication SH/0/B/2005/001. to assess the release with on radiation monitor 4.3.G.1-1 A valid indication procedure HP/0/B/1009/010, EMF-36H of on radiation monitor 4.3.U.1-3 A valid indication on radiation monitor HP/0/B/1009/029, or >3.4 E + 03 cpm EMF-36H of EMF-31 (when aligned to WC or SH/0/B/2005/001. 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 16.11-6 for > 15 minutes as 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/1009/010, HP/0/B/1009/029, or mRem TEDE or 500 consequences SH/0/B/2005/001. 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 16.11-6 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 two times SLC 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 mRem CDE Adult Thyroid at the site boundary.

END

Enclosure 4.3 RP/0/A/5700/000 Abnormal Rad Levels/Radiological Effluent Page 3 of 5 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 1EMF-42 2EMF-42 Reactor Building Refueling Bridge 1EMF-16*

2EMF-3*

Containment Noble Gas 1EMF-39*

2EMF-39*

Applies to Mode 6 and No Mode Only.

(Continued)

Enclosure 4.3 RP/0/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 RP/0/A/5700/000 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.3-3 Valid area EMF reading exceeds the levels shown in Enclosure 4.10.

END

Enclosure 4.4 RP/0/A/5700/000 Loss of Shutdown Functions Page 1 of 3 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 WAS NOT Reactor Trip Once a Reactor Trip Once a Successful and There is Reactor Protection System Reactor Protection System Indication of an Extreme Setpoint Has Been Exceeded Setpoint Has Been Exceeded Challenge to the Ability to and Manual Trip WAS and Manual Trip WAS NOT Cool the Core.

Successful. Successful.

OPERATING MODE: 1 OPERATING MODE: 1, 2, 3 OPERATING MODE: 1 4.4.G.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/0/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.2-1 Total loss of ND and/or RN and/or KC. 4.4.S.2-1 Subcriticality CSF-RED.

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

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

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

below 200ºF 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

>180ºF.

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

(Continued)

Enclosure 4.4 RP/0/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 RP/0/A/5700/000 Loss of Power Page 1 of 3 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 RP/0/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. NOT 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 following condition voltage less than degradation of core One emergency diesel exists: 110 VDC. cooling based on generator is supplying Fission Product power to its respective AC power capability has AND Barrier monitoring.

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

of:

END SATA SATB ATC ATD D/G A D/G B.

END

Enclosure 4.5 RP/0/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 RP/0/A/5700/000 Fire/Explosion and Security Events Page 1 of 4 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.S.1-1 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: MNS 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 y Reactor Building 4.6.G.1-2 A HOSTILE ACTION has verification of a control room y Auxiliary Building caused failure of Spent Fuel fire alarm. y Diesel Generator Rooms Cooling Systems and y Control Room IMMINENT fuel damage is y Reactor Building y Standby Shutdown Facility likely for a freshly off-loaded y Auxiliary Building y CAS reactor core in pool.

y Diesel Generator Rooms y SAS y Control Room y FWST END y Standby Shutdown Facility y Doghouses (Applies in y CAS Mode 1, 2, 3, 4 only).

y SAS y Doghouses AND y FWST y Turbine Building y Service Building (Continued) y Interim Radwaste Building y Equipment Staging Building

ISFSI.

(Continued)

Enclosure 4.6 RP/0/A/5700/000 Fire/Explosion and Security Events Page 2 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY One of the following:

4.6.U.1-2 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. y Spent Fuel Pool y Auxiliary Building.

END AND (Continued)

Enclosure 4.6 RP/0/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/0/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 RP/0/A/5700/000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 1 of 4 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-13, 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/1(2)/A/5500/017, or Coordinator/EOF Director protected area area: AP/1(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 cannot be with potential for loss of protected area

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

Control Room Shutdown Panel or the

Standby Shutdown Standby Shutdown Facility OR 4.7.U.1-4 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 RP/0/A/5700/000 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 2 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.U.1-5 Independent Spent Fuel Cask 4.7.A.1-3 Visible structural damage 4.7.S.2 Other Conditions Existing tipped over or dropped greater caused by either: Which in the Judgement of than 12 inches.

Vehicle crashes, OR the Emergency

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

Other catastrophic events Area Emergency.

4.7.U.1-7 Tornado generated missile(s) on any of the following plant OPERATING MODE: ALL impacting the ISFSI. structures:

4.7.S.2-1 Other conditions exist which

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

Auxiliary Building Emergency Coordinator/EOF Flammable Gases Deemed

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/0/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 judgement of the Emergency Coordinator/EOF Note: Structures for the below EALs:

Director indicate a potential

Reactor Building degradation of the level of

Auxiliary Building safety of the plant.

Diesel Generator Rooms

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/0/A/5700/000 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.3-1 Control Room evacuation has been initiated per AP/1(2)/A/5500/017, or AP/1(2)/A/5500/024. {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

Enclosure 4.8 RP/0/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.

COGNIZANT FACILITY STAFF - any member of facility staff, who by virtue of training and experience is qualified to assess the indications or reports for validity and to compare the same to the EALs in the licensee's emergency classification scheme. (Does not include staff whose positions require they report, rather than assess, abnormal conditions to the facility.)

EXPLOSION - A rapid, violent unconfined combustion, or a catastrophic failure of pressurized/energized 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.

Enclosure 4.8 RP/0/A/5700/000 Definitions/Acronyms Page 2 of 4 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).

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.

Enclosure 4.8 RP/0/A/5700/000 Definitions/Acronyms Page 3 of 4 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: (l) 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 10%.

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

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: (l ) an instrument channel check, or (2) indications on related or redundant instrumentation, or (3) by direct observation by plant personnel such that doubt related to the instrument's operability, the condition's existence or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Enclosure 4.8 RP/0/A/5700/000 Definitions/Acronyms Page 4 of 4 VIOLENT - Force has been used in an attempt to injure site personnel or damage plant property.

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

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

VITAL AREA - Areas within the PROTECTED AREA that house equipment important for nuclear safety. Access to a VITAL AREA is allowed only if an individual has been authorized to be in that area per the Security plan, therefore VITAL AREA is a Security term.

Enclosure 4.9 RP/0/A/5700/000 Emergency Declaration Guidelines Page 1 of 2 THE FOLLOWING GUIDANCE IS TO BE USED BY THE EMERGENCY COORDINATOR IN ASSESSING EMERGENCY CONDITIONS.

Assessment, classification and declaration of any applicable emergency condition should be completed within 15 minutes after indication or information is available to cognizant facility staff that an EAL threshold has been exceeded.

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-M97-4638} If the initiating event puts more than one unit in the same Emergency Classification (example Alert), then the unit designation may be either the Affected Unit numbers or "All." If the initiating event drives one unit to a higher classification, then the unit with the higher classification should be listed as Affected Unit. {PIPs M-03-3294 and C-04-2586}

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

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

The fission product barrier matrix is applicable only to those events that occur at hot shutdown or higher.

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

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

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

Enclosure 4.9 RP/0/A/5700/000 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 10CFR50.72 may be required. Such a condition could occur, for example, if a follow-up evaluation of an abnormal condition uncovers evidence that the condition was more severe than earlier believed.

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

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

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

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

DETERMINATION OF "EVENT TIME" (TIME THE 15 MINUTE CLOCK STARTS)

1. 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/0/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' EE, FF-58 Unit 2 NM Sample Room 300 5000 2EMF-4 786' N/A Unit 2 Spent Fuel Pool Refueling Bridge 100 5000 2EMF-9 767' JJ-59 Aux. Bldg. Corridor 100 5000

Enclosure 4.11 RP/0/A/5700/000 Commitment Reference for Emergency Action Levels Page 1 of 1

{1} PIP-M-00-2138, CA # 18

{2} PIP-M-02-0187, CA # 6

{3} PIP-M-01-2860, CA #2

{4} PIP-M-03-4281, CA #3

{5} PIP-M-05-3403, CA#3, multiple changes in enclosure 4.6

Examination KEY for: ILT-31 MNS SRO NRC Examination Question Answer Number 1 B 2 B 3 A 4 D 5 D 6 B 7 A 8 D 9 B 10 C 11 B 12 D 13 B 14 C 15 D 16 C 17 D 18 A 19 A 20 D 21 C 22 C 23 B 24 D 25 C Printed 4/24/2015 1:47:10 PM Page 1 of 4

Examination KEY for: ILT-31 MNS SRO NRC Examination Question Answer Number 26 D 27 D 28 A 29 B 30 D 31 C 32 A 33 C 34 A 35 C 36 C 37 A 38 C 39 C 40 C 41 B 42 D 43 B 44 A 45 B 46 A 47 D 48 C 49 B 50 C Printed 4/24/2015 1:47:10 PM Page 2 of 4

Examination KEY for: ILT-31 MNS SRO NRC Examination Question Answer Number 51 A 52 D 53 B 54 C 55 D 56 C 57 B 58 A 59 D 60 D 61 D 62 C 63 B 64 C 65 B 66 A 67 D 68 B 69 B 70 B 71 D 72 D 73 B 74 B 75 A Printed 4/24/2015 1:47:10 PM Page 3 of 4

Examination KEY for: ILT-31 MNS SRO NRC Examination Question Answer Number 76 C 77 B 78 C 79 A 80 C 81 C 82 A 83 B 84 A 85 C 86 B 87 B 88 B 89 C 90 C 91 D 92 C 93 B 94 B 95 A 96 B 97 D 98 C 99 C 100 C Printed 4/24/2015 1:47:10 PM Page 4 of 4

ML15141A173

Contents

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RC PUMP DATA ==========================================

RECOMMENDED RELEASE RATE ============================

SETPOINT DATA =========================================

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ML15141A173

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RC PUMP DATA ==========================================

RECOMMENDED RELEASE RATE ============================

SETPOINT DATA =========================================

References:

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