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Catawba Initial Exam 2015-301 Final SRO Written Exam
	ML15181A232
Person / Time
Site:	Catawba
Issue date:	06/30/2015
From:	; Division of Nuclear Materials Safety II
To:	;
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	Download: ML15181A232 (144)
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Question:(1 point)1ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 entered EP/1/A/5000/ECA-1.1, (Loss of Emergency Coolant Recirculation) following a LOCA outside containment

Safety Injection Termination criteria is NOT met o The crew has been directed to determine minimum SI flow per Enclosure 4 (Minimum S/I Flowrate Versus Time After Trip)

Current conditions:

The Unit 1 Reactor was tripped 60 minutes ago

NCS pressure is 1000 psig.

1B NI pump is running with flow indicated at 380 gpm

1A NV pump is running with flow indicated at 400 gpm

The MINIMUM S/I Flowrate required to remove current reactor decay heat is _____(1)_____.

The _____(2)_____ pump is required to be secured.

REFERENCE PROVIDED A. 1. 340 gpm 2. 1A NV B. 1. 340 gpm 2. 1B NI C. 1. 360 gpm 2. 1A NV D. 1. 360 gpm 2. 1B NI

Page 1 of 100 Question:(1 point)2ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The crew has just completed the actions of EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) following a Safety Injection due to Hi Containment Pressure

NV pump flow to the NC system Cold Legs is 390 GPM

NC system pressure is 1350 PSIG and STABLE

SG pressures are 1092 PSIG and STABLE

NC system subcooling on the ICCM is 22°F and STABLE

Per the requirements of EP/1/A/5000/E-0, Enclosure 1 (Foldout Page), the crew _____(1)_____ be required to secure NC pumps.

Upon transition to EP/1/A/5000/E-1 (Loss of Reactor or Secondary Coolant), Steam Generators _____(2)_____ be required for heat removal.

A. 1. will 2. will B. 1. will 2. will NOT C. 1. will NOT 2. will D. 1. will NOT 2. will NOT

Page 2 of 100 Question:(1 point)3ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 3

Rod control is capable of rod withdrawal.

NC loops 1A, 1B, and 1D are in operation.

The crew has entered AP/1/A/5500/008 (Malfunction of Reactor Coolant Pump)

1A NC Pump lower bearing temperature is currently 190ºF and increasing 5ºF per minute (1) 1A NC Pump lower bearing temperature will reach trip setpoint in ________________ . (2) What is the specified Completion Time of the action required by TS 3.4.5 (RCS Loops - MODE 3) following the trip of 1A NCP?

A. 1. 7 minutes 2. immediately B. 1. 7 minutes 2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. 1. 1 minute 2. immediately D. 1. 1 minute 2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Page 3 of 100 Question:(1 point)4ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 refueling is in progress

NC level is at 98%

1B ND is in service

1AD-9 B/6 "ND Trn B to NC C-Legs Loops A-B Lo Flow" alarms

The BOP notes the following amperage indication

The crew will enter _____(1)_____ of AP/1/A/5500/019 (Loss of Residual Heat Removal System).

Once 1A ND is placed in service, the requirements of T.S. 3.9.4 (Residual Heat Removal and Coolant Circulation - High Water Level) _____(2)_____ met.

A. 1. Case II (Leak in ND) 2. are B. 1. Case II (Leak in ND) 2. are NOT C. 1. Case III (Loss of ND with Large Vent Path Established) 2. are D. 1. Case III (Loss of ND with Large Vent Path Established) 2. are NOT Page 4 of 100 Question:(1 point)5ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

1A1 KC Pump is in service

1B NV Pump is in service

OAC Alarm C1D2215 (KC Train A Low-Low Level Surge Tank Isol) is actuated due to an instrument failure

Following the receipt of the OAC alarm, and assuming no operator action:

1RN-291 (KC Hx 1A Outlet Throttle Valve) will throttle in the _____(1)_____ direction in order to maintain KC system temperature.

1B NV Pump motor bearing temperatures will _____(2)_____ .

A. 1. open 2. increase B. 1. open 2. remain constant C. 1. closed 2. increase D. 1. closed 2. remain constant Page 5 of 100 Question:(1 point)6ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

A DCS malfunction has caused an Alternate Action on SPP-1 (Selected PZR Pressure-1)

Subsequently:

The Main Turbine trips

The Reactor does NOT trip automatically or manually

The crew has entered EP/1/A/5000/FR-S.1 (Response to Nuclear Power Generation/ATWS)

As NC pressure increases, 1NC-32B (PZR PORV) _____(1)_____ automatically open.

FR-S.1 contains guidance to ensure PZR pressure remains below the PORV setpoint in order to _____(2)_____ .

A. 1. will 2. ensure sufficient boration flow B. 1. will 2. prevent challenging PZR Safety Valves C. 1. will NOT 2. ensure sufficient boration flow D. 1. will NOT 2. prevent challenging PZR Safety Valves Page 6 of 100 Question:(1 point)7ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 Reactor has experienced a Safety Injection due to Low PZR Pressure

The CRS has just completed reading the immediate action steps of EP/1/A/5000/E-0 (Reactor Trip or Safety Injection)

The following EMFs are in Trip 2 condition: o 1EMF-33 (Condenser Air Ejector Exhaust) o 1EMF-71 (S/G A Leakage) o 1EMF-74 (S/G D Leakage) o 1EMF-26 (Steam Line 1A)

1A S/G NR level is 35%

1B, 1C and 1D S/G NR levels are 23%

Total CA flow is approximately 1000 gpm

The RO has depressed the "CA SYS VLV CTRL TRN A" and "CA SYS VLV CTRL TRN B" reset buttons * "CA SYS VLV CTRL TRN A" reset light is DARK * "CA SYS VLV CTRL TRN B" reset light is LIT

CA flow to 1A S/G using the flow control valve _____(1)_____ under operator control

EP/1/A/5000/E-0 _____(2)_____ require isolation of CA flow to the 1A S/G at this time.

A. 1. is 2. does B. 1. is 2. does NOT C. 1. is NOT 2. does D. 1. is NOT 2. does NOT Page 7 of 100 Question:(1 point)8ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A seismic event has resulted in the following: o 2A S/G has experienced a complete shear of the Main Steam line at the S/G outlet o 2D S/G has experienced a complete shear of the Main Feed line at the S/G inlet

Which Steam Generator will decrease to 0% WR level FIRST?

Procedural guidance to isolate 2D S/G is contained in _____(2)_____ . A. 1. 2A S/G 2. EP/2/A/5000/E-1 (Loss of Reactor or Secondary Coolant) B. 1. 2A S/G 2. EP/2/A/5000/E-2 (Faulted Steam Generator Isolation) C. 1. 2D S/G 2. EP/2/A/5000/E-1 (Loss of Reactor or Secondary Coolant) D. 1. 2D S/G 2. EP/2/A/5000/E-2 (Faulted Steam Generator Isolation) Page 8 of 100 Question:(1 point)9ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Both units have experienced a Loss of Offsite Power (LOOP)

No backup Diesel VI compressor is available

EP/1/A/5000/ECA-0.0 (Loss of All AC Power) was implemented on Unit 1 and the crew has just transitioned to EP/1/A/5000/ECA-0.1 (Loss of All AC Power Recovery Without S/I Required)

Current conditions:

30 minutes have elapsed since the LOOP

NC Thots are STABLE

S/G pressures are STABLE at 725 PSIG

S/G levels are decreasing and approaching 11% NR

NC Tcolds are 490°F and STABLE

VI header pressure is 0 PSIG Natural Circulation flow (1) be verified In accordance with ECA-0.1, the Operators will increase CA flow (2) .

REFERENCE PROVIDED A. 1. can 2. using flow controllers in the control room B. 1. can 2. by notifying AO to throttle CA valves locally C. 1. can NOT 2. using flow controllers in the control room D. 1. can NOT 2. by notifying AO to throttle CA valves locally Page 9 of 100 Question:(1 point)10ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 tripped from 100% RTP due to a Loss of Offsite Power

NC cooldown is currently in progress in accordance with EP/1/A/5000/ES-0.2 (Natural Circulation Cooldown)

The trend of NC Temperature is as follows:

Time 0800 0830 0900 NC Temperature 535ºF 495ºF 455ºF

In accordance with ES-0.2:

NC Temperature will be determined by monitoring of _____(1)_____ .

At 0900, the required cooldown rate _____(2)_____ been exceeded.

A. 1. NC Tave 2. has B. 1. NC Tave 2. has NOT C. 1. NC Tcolds 2. has D. 1. NC Tcolds 2. has NOT Page 10 of 100 Question:(1 point)11ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

1ERPA is de-energized

The crew has entered AP/1/A/5500/029 (Loss of Vital or Aux Control Pwr)

In accordance with AP/29, an operator will be dispatched to 1A D/G room because _____(1)_____ .

Entry conditions of AP/1/A/5500/016 (Malfunction of Nuclear Instrumentation System) _____(2)_____ been met.

A. 1. D/G control power has been lost 2. have B. 1. D/G control power has been lost 2. have NOT C. 1. D/G Day Tank auto makeup capability has been lost 2. have D. 1. D/G Day Tank auto makeup capability has been lost 2. have NOT Page 11 of 100 Question:(1 point)12ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

The 125 VDC/120 VAC Auxiliary Control Power System is in normal alignment

The supply breaker from 1CDA to Static Inverter 1KXIA trips open Alternate power to 1KXPA will be _____(1)___ aligned. The crew can verify 1KXPA is energized by observing that _____(2)_____ .

A. 1. manually 2. NC pump vibration monitor is IN SERVICE B. 1. automatically 2. NC pump vibration monitor is IN SERVICE C. 1. manually 2. individual P-11 Channel 1 status light on 1SI-7 is LIT D. 1. automatically 2. individual P-11 Channel 1 status light on 1SI-7 is LIT Page 12 of 100 Question:(1 point)13ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Units 1 & 2 are at 100% RTP

1A RN pump is in service

1A and 2A KC trains are in service Subsequently:

The following Unit 1 annunciators are lit o 1AD-12 B/2 "RN PIT A Screen Hi D/P" o 1AD-12 B/1 "RN Pump Intake Pit A Level - LO" o 1AD-12 E/2 "RN Pit A Swap to SNSWP"

The crew has entered AP/0/A/5500/020 (Loss of Nuclear Service Water), Case II (Loss of RN Pit Level)

Enclosure 2 (RN Valve Alignment for RN Swap to SNSWP) will direct the BOP to ensure _____(1)_____ is closed .

Following system stabilization, the BOP is directed to "Ensure KC Hx Outlet Mode Switches - Properly Aligned". In response, the BOP _____(2)_____ required to reposition 1RN-291 (KC Hx 1A Outlet Throttle Valve) from its original alignment.

A. 1. 1RN-47A (RN Supply X-Over Isol) 2. is B. 1. 1RN-47A (RN Supply X-Over Isol) 2. is NOT C. 1. 1RN-48B (RN Supply X-Over Isol) 2. is D. 1. 1RN-48B (RN Supply X-Over Isol) 2. is NOT Page 13 of 100 Question:(1 point)14ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Both units are at 100% RTP

VI pressure is decreasing

The crew has entered AP/0/A/5500/022 (Loss of Instrument Air)

In accordance with AP/22, an operator will be dispatched to align air cylinders to operating compressors if VI pressure decreases below a MAXIMUM of _____(1)_____ in order to _____(2)_____ .

A. 1. 60 psig 2. provide a backup VI supply B. 1. 60 psig 2. prevent compressor damage C. 1. 76 psig 2. provide a backup VI supply D. 1. 76 psig 2. prevent compressor damage Page 14 of 100 Question:(1 point)15ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

Operators are controlling the Main Generator Voltage Regulator in MANUAL

An electrical grid disturbance results in the Main Generator operating with a power factor of 0.8 lagging

The concern with operating the Main Generator at a 0.8 lagging power factor is overheating of the generator _____(1)_____ .

If the OATC depresses the Voltage Adjust "RAISE" pushbutton, power factor will become _____(2)_____ lagging.

REFERENCE PROVIDED A. 1. core end 2. more B. 1. field windings 2. more C. 1. core end 2. less D. 1. field windings 2. less Page 15 of 100 Question:(1 point)16ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

EP/1/A/5000/ECA-1.2 (LOCA Outside Containment) has been entered following an event

Containment Phase A Isolation _____(1)_____ automatically occurred.

In accordance with ECA-1.2, the parameter used to verify that the LOCA has been isolated is _____(2)_____ .

A. 1. has NOT 2. pressurizer level B. 1. has NOT 2. NC pressure C. 1. has 2. pressurizer level D. 1. has 2. NC pressure Page 16 of 100 Question:(1 point)17ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 has experienced a Safety Injection due to Hi Containment Pressure

Containment pressure peaked at 2.7 psig and is now slowly decreasing

The crew has implemented EP/1/A/5000/FR-H.1 (Response to Loss of Secondary Heat Sink)

All attempts to restore CA flow have been unsuccessful

In accordance with FR-H.1:

The NEXT source of feed water attempted for restoration of flow to the S/Gs is through the CM/CF system using _____(1)_____ . The crew will be required to establish bleed and feed when W/R level in at least 3 S/Gs is less than a MAXIMUM level of _____(2)_____ .

A. 1. either Main Feed Water pump 2. 24% B. 1. either Main Feed Water pump 2. 36% C. 1. Hotwell and Booster pumps 2. 24% D. 1. Hotwell and Booster pumps 2. 36% Page 17 of 100 Question:(1 point)18ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Which of the following is a complete list of the reasons for the reactor operator performing a controlled depressurization of the Steam Generators (SGs) during the performance of EP/1/A/5000/ECA-1.1, (Loss of Emergency Coolant Recirculation)?

List of Reasons 1. Minimize NC dilution potential in case of a subsequent tube rupture 2. To establish conditions for injection of the Cold Leg accumulators 3. Minimize reactor coolant flow from the LOCA

4. To establish conditions for ND system operation

A. 1 and 2 ONLY B. 3 and 4 ONLY C. 1, 2, and 3 ONLY D. 2, 3, and 4 ONLY Page 18 of 100 Question:(1 point)19ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The crew is responding to a continuous rod withdrawal on Unit 1 per AP/1/A/5500/015 Case 2, Continuous Rod Motion

Control rods have been placed in manual and rod motion has stopped

Boron is being added with 1A boric acid transfer pump to return Tavg to Tref Subsequently:

1ETA experiences a loss of power

The blackout sequencer re-energized 1ETA

Which choice states the minimum action(s), if any, required to be completed before the operator can secure the 1A boric acid transfer pump using its control switch?

A. Reset the 1A diesel generator sequencer and then reset the 1A boric acid transfer pump B. Reset the 1A diesel generator sequencer C. Reset the 1A boric acid transfer pump D. No additional actions are required Page 19 of 100 Question:(1 point)20ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station In accordance with SLC 16.9-19 (Refueling Operations - Manipulator Crane), the overload cutoff limit for the Reactor Building Manipulator crane is required to be less than or equal to _____(1)_____ pounds.

The reason for the manipulator crane overload limits is to prevent damage to the _____(2)_____ .

A. 1. 2900 2. Core Internals B. 1. 3250 2. Core Internals C. 1. 2900 2. Refueling Bridge D. 1. 3250 2. Refueling Bridge Page 20 of 100 Question:(1 point)21ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Steam Generator Tube Leak has occurred on Unit 2

All NC pumps are running

The crew is preparing to depressurize the NC system to minimize subcooling Which of the following sets of operating conditions will result in the LEAST Primary-to-Secondary leakage?

NCS Temperature NCS Pressure A. 504°F 770 PSIG B. 512°F 826 PSIG C. 516°F 855 PSIG D. 524°F 885 PSIG Page 21 of 100 Question:(1 point)22ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 2 is at 50% Power.

The crew enters AP/2/A/5500/023 (Loss of Main Condenser Vacuum) due to decreasing vacuum. The following conditions are noted: Time 1257 1258 1259 1300 A Cond. Vacuum 23.2" 22.0" 21.7" 21.5" B Cond. Vacuum 24.0" 22.6" 21.9" 21.7" C Cond. Vacuum 24.3" 22.9" 22.3" 22.0" Based upon the above information, which of the following describes: 1) The latest time the crew must take action(s) PRIOR to exceeding the turbine trip setpoint? 2) The action(s) the crew must take?

A. 1. 1258 2. Trip the Turbine and Reactor B. 1. 1258 2. Trip the Turbine ONLY C. 1. 1259 2. Trip the Turbine and Reactor D. 1. 1259 2. Trip the Turbine ONLY Page 22 of 100 Question:(1 point)23ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The Control Room has been evacuated due to toxic gas

Following Control Room evacuation, per AP/1/A/5500/017 (Loss of Control Room), which of the following completes the statements below?

Reactor shutdown condition can be verified by monitoring of _____(1)_____ .

Shutdown margin will be maintained by use of the _____(2)_____ .

A. 1. Source Range instrumentation 2. Standby Makeup Pump B. 1. Source Range instrumentation 2. Boric Acid Transfer Pumps C. 1. Wide Range Neutron instrumentation 2. Standby Makeup Pump D. 1. Wide Range Neutron instrumentation 2. Boric Acid Transfer Pumps Page 23 of 100 Question:(1 point)24ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station In accordance with AP/1/A/5500/018, (High Activity in Reactor Coolant) which of the following is a complete list of EMFs used to validate High Reactor Coolant Activity?

1. 1EMF-18 (NC Filter 1A)

2. 1EMF-48 (NC Sample Line Reactor Coolant)

3. 1EMF-39 (Containment Gas Hi Rad)

A. 1 ONLY B. 2 and 3 ONLY C. 1 and 2 ONLY D. 1, 2, and 3 Page 24 of 100 Question:(1 point)25ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station In accordance with EP/1/A/5000/ES-1.2 (Post LOCA Cooldown And Depressurization): The crew will FIRST attempt to establish an NC system cooldown using the _____(1)_____ .

The crew will cooldown _____(2)_____ .

A. 1. SM PORVs 2. as close as possible without exceeding 100°F in an hour B. 1. SM PORVs 2. at the maximum rate C. 1. Condenser Dumps 2. as close as possible without exceeding 100°F in an hour D. 1. Condenser Dumps 2. at the maximum rate Page 25 of 100 Question:(1 point)26ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Following an event, equipment failures have resulted in a RED condition on the NC Integrity CSF Status Tree.

NC Cooldown rate was approximately 220°F per hour

NC System temperature is currently 240°F

The crew is performing a soak in accordance with EP/1/A/5000/FR-P.1 (Response to Imminent Pressurized Thermal Shock Condition).

Which 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. Initiate a cooldown at less than 50°F per hour Page 26 of 100 Question:(1 point)27ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 has experienced an event

Containment pressure is 2.8 psig

Containment radiation is indicating 30 R/Hr

Containment sump level is 17 feet

Entry requirements for _____(1)_____ are currently met.

TS 3.3.3 (Post-Accident Monitoring Instrumentation) requires operability of _____(2)_____ .

A. 1. EP/1/A/5000/FR-Z.2 (Response to Containment Flooding) 2. Containment Radiation Instruments ONLY B. 1. EP/1/A/5000/FR-Z.2 (Response to Containment Flooding) 2. Containment Radiation AND Containment Sump Water Level Instruments C. 1. EP/1/A/5000/FR-Z.3 (Response to High Containment Radiation Level) 2. Containment Radiation Instruments ONLY D. 1. EP/1/A/5000/FR-Z.3 (Response to High Containment Radiation Level) 2. Containment Radiation AND Containment Sump Water Level Instruments Page 27 of 100 Question:(1 point)28ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Which of the following alarms would confirm a failure of a #2 NCP seal?

1A NCP Standpipe _____(1)_____ level alarm.

1A NCP #1 Seal Leakoff _____(2)_____ flow alarm.

A. 1. High 2. High B. 1. High 2. Low C. 1. Low 2. High D. 1. Low 2. Low Page 28 of 100 Question:(1 point)29ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A malfunction of 1NV-224 (H2 Sup to VCT Press Reg Vlv) has isolated the VCT Hydrogen supply

VCT Purge is in progress

1AD-7 H/2 (VCT HI/LO Pressure) is in alarm

As VCT pressure decreases, NCP #1 Seal leakoff flow will _____(1)_____ .

The Annunciator Response Procedure for 1AD-7 H/2 will direct Chemistry to _____(2)_____ .

A. 1. decrease 2. secure VCT purge B. 1. decrease 2. increase primary sample frequency C. 1. increase 2. secure VCT purge D. 1. increase 2. increase primary sample frequency Page 29 of 100 Question:(1 point)30ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station The 1A Boric Acid Transfer Pump receives power from __________ .

A. SMXG

B. 1EMXA

C. 1EMXG

D. 1MXW Page 30 of 100 Question:(1 point)31ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The Unit 1 NC system is currently SOLID in Mode 5

1A ND is in service

1B ND is secured

ND Letdown is in service

The BOP has been directed to perform a 10ºF heatup

To commence heatup, the BOP will manually operate _____(1)_____ .

In order to prevent exceeding LTOP limits during the heatup, the BOP will throttle 1NV-135 (ND Flow to Letdown HX) in the _____(2)_____ direction.

A. 1. 1ND-26 (ND HX 1A Outlet CTRL) 2. open B. 1. 1ND-26 (ND HX 1A Outlet CTRL) 2. closed C. 1. 1ND-27 (ND HX 1A Bypass CTRL) 2. open D. 1. 1ND-27 (ND HX 1A Bypass CTRL) 2. closed Page 31 of 100 Question:(1 point)32ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The crew is performing EP/1/A/5000/E-3 (Steam Generator Tube Rupture) following a Unit 1 event o 1B S/G has been isolated per EP/1/A/5000/E-3 o A plant cooldown has been initiated o The crew is preparing to depressurize the NC system o ALL NC pumps are in service

During NC cooldown, 1B NC Hot Leg Temperature (Thot) will be _____(1)_____ 1C NC Hot Leg (Thot) Temperature.

During NC depressurization 1NC-27 (PZR Spray Ctrl Frm Loop A) will be _____(2)_____ effective than 1NC-29 (PZR Spray Ctrl Frm Loop B)

A. 1. equal to 2. less B. 1. higher than 2. less C. 1. equal to 2. more D. 1. higher than 2. more Page 32 of 100 Question:(1 point)33ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in the process of drawing a bubble in the Pressurizer

The following step is taken from OP/1/A/6100/001 (Controlling Procedure For Unit Startup) following actions to vent Nitrogen from the PZR as the bubble is formed:

In order to determine when Nitrogen venting is complete, the operators will verify _____(1)_____ .

If necessary to lower PRT pressure following this evolution, PRT level will be decreased by draining to the _____(2)_____ .

A. 1. PRT temperature equalizes with PZR steam space temperature 2. Reactor Coolant Drain Tank B. 1. PRT temperature equalizes with PZR steam space temperature 2. Containment Floor and Equipment Sump C. 1. PRT level increases without a corresponding PRT pressure increase 2. Reactor Coolant Drain Tank D. 1. PRT level increases without a corresponding PRT pressure increase 2. Containment Floor and Equipment Sump Page 33 of 100 Question:(1 point)34ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Unit 1 event has occurred

Containment pressure peaked at 3.2 psig

The non-essential KC headers were automatically isolated by the _____(1)_____ signal.

Based on this alignment, Spent Fuel Pool temperature will _____(2)_____ .

A. 1. Phase A 2. increase B. 1. Phase A 2. decrease C. 1. Phase B 2. increase D. 1. Phase B 2. decrease Page 34 of 100 Question:(1 point)35ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

A slight cooldown of the NC system causes the "C" PZR heaters to be full "on"

A malfunction of two PZR pressure transmitters causes an Alternate Action to occur on the Pressurizer Pressure Control System

Assuming NO operator actions:

The PZR Pressure Master will be in _____(1)_____ control AND

"C" Heaters _____(2)_____ be energized.

A. 1. Manual 2. will B. 1. Manual 2. will NOT C. 1. Automatic 2. will D. 1. Automatic 2. will NOT Page 35 of 100 Question:(1 point)36ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 2 is at 100% RTP

2B Pressurizer heaters are manually energized to promote mixing following unit start-up

All plant parameters are at equilibrium

A fault causes breaker 2LXH-6C (2B NC PZR Heater Power Panel PHP2B Feeder) to OPEN.

1) How will the output of the pressurizer master control change? 2) What is the system's response to the loss of heat input?

A. 1. Increase 2. Pressurizer Spray Valves will close prior to "A" & "D" Pressurizer Heaters energizing B. 1. Decrease 2. Pressurizer Spray Valves will close prior to "A" & "D" Pressurizer Heaters energizing C. 1. Increase 2. "A" & "D" Pressurizer Heaters will energize prior to Pressurizer Spray Valves closing D. 1. Decrease 2. "A" & "D" Pressurizer Heaters will energize prior to Pressurizer Spray Valves closing Page 36 of 100 Question:(1 point)37ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Which of the following describes the impact on the Reactor Protection System for a loss of 125 VDC Panelboard 1EPA or 1EPD?

A. SSPS Logic Bay has lost one of two power supplies B. SSPS Output Bay has lost one of two power supplies C. The associated Reactor Trip Breaker cannot be opened by the UV trip D. The associated Reactor Trip Breaker cannot be opened by the Shunt trip Page 37 of 100 Question:(1 point)38ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Which of the following describes features of the Engineered Safety Features Actuation System (ESFAS) which are designed to prevent spurious actuations?

A loss of a MAXIMUM of _____(1)_____ Vital I&C channel(s) will NOT result in an actuation.

Hi Containment Pressure Safety Injection signal logic requires _____(2)_____ channels above setpoint for actuation.

A. 1. one 2. 2 of 3 B. 1. one 2. 2 of 4 C. 1. two 2. 2 of 3 D. 1. two 2. 2 of 4 Page 38 of 100 Question:(1 point)39ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

An accident has occurred on Unit 1

The Containment Air Return Fans will start _____(1)_____ following receipt of an Sp signal.

In order to open, Containment Air Return Dampers require differential pressure to be less than or equal to a MAXIMUM of _____(2)_____ .

A. 1. 10 seconds 2. 0.5 psid B. 1. 10 seconds 2. 0.9 psid C. 1. 9 minutes 2. 0.5 psid D. 1. 9 minutes 2. 0.9 psid Page 39 of 100 Question:(1 point)40ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Under normal conditions, Containment Ventilation Units are cooled by the _____(1)_____ system.

Containment Ventilation cooling water supply is isolated by a _____(2)_____ signal.

A. 1. YV 2. Phase A B. 1. YV 2. Phase B C. 1. RN 2. Phase A D. 1. RN 2. Phase B Page 40 of 100 Question:(1 point)41ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Unit 1 event has occurred

NC System pressure is 1940 PSIG and lowering slowly

Containment pressure is 2.3 PSIG and rising slowly

NF (ICE CONDENSER) system valve(s) will receive a signal to CLOSE.

COMPONENT LEGEND:

1NF-228A (Glycol Sup Cont Isol Otsd) 1NF-233B (Glycol Ret Cont Isol) 1NF-234A (Glycol Ret Cont Isol)

A. 1NF-233B ONLY B. 1NF-228A and 1NF-233B ONLY C. 1NF-228A and 1NF-234A ONLY D. 1NF-228A, 1NF-233B and 1NF-234A Page 41 of 100 Question:(1 point)42ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station The bistables required to permit Containment Spray Pump operation will _____(1)_____ to actuate. Once started, the NS pumps will secure on decreasing containment pressure at _____(2)_____ .

A. 1. energize 2. 0.9 psig B. 1. energize 2. 0.35 psig C. 1. de-energize 2. 0.9 psig D. 1. de-energize 2. 0.35 psig Page 42 of 100 Question:(1 point)43ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Unit 1 reactor trip has occurred due to a secondary system malfunction

EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) has been performed and a transition has been made to EP/1/A/5000/ES-0.1 (Reactor Trip Response)

The crew has entered EP/1/A/5000/FR-H.2 (Response to Steam Generator Overpressure)

The crew is preparing to dump steam from the affected S/G

FR-H.2 will only allow steam release from the affected S/G if NR level is less than a MAXIMUM of _____(1)_____ .

If the maximum level has been exceeded, an evaluation must be performed prior to release due to the potential effects of _____(2)_____ .

A. 1. 83% 2. steamline water hammer B. 1. 92% 2. steamline water hammer C. 1. 83% 2. condenser tube damage D. 1. 92% 2. condenser tube damage Page 43 of 100 Question:(1 point)44ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 2 is at 100% RTP Subsequently:

2A CFPT tripped

All S/G levels remain within a 48%-52% band for the 30 seconds following the CFPT trip

If reactor power was greater than a MINIMUM of _____(1)_____ prior to the event, S/G level control setpoints will change to a lower specific value for a period of _____(2)_____ prior to ramping back to the normal programmed value.

A. 1. 56% 2. 6 minutes B. 1. 65% 2. 6 minutes C. 1. 56% 2. 10 minutes D. 1. 65% 2. 10 minutes Page 44 of 100 Question:(1 point)45ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

The Steam Generator Water Level Control system will automatically runback CFPTs at 3% per minute if _____(1)_____ until the condition is cleared.

In order to ensure 1B CFPT assumes MORE load than 1A, "CFPT BIAS" must be adjusted _____(2)_____ than 0%

A. 1. CFPT speed exceeds 6226 rpm 2. greater B. 1. CFPT speed exceeds 6226 rpm 2. less C. 1. CF Header Pressure exceeds 1300 psig 2. greater D. 1. CF Header Pressure exceeds 1300 psig 2. less Page 45 of 100 Question:(1 point)46ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Unit 1 event has occurred

Containment pressure 4.5 psig

No Containment EMFs are in alarm

The crew has entered EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) S/G 1A 1B 1C 1D NR Level 28% increasing 26% increasing 9% decreasing 24% increasing CA Flow 225 gpm 200 gpm 250 gpm 150 gpm

Prior to being given specific S/G level control guidance by the CRS, the OATC can reduce feed flow to non-faulted S/Gs when level reaches _____(1)_____ .

Following completion of Enclosure 1 (Foldout Page) actions, S/G _____(2)_____ will be receiving the highest CA flow rate.

A. 1. 29% 2. 1A B. 1. 29% 2. 1D C. 1. 39% 2. 1A D. 1. 39% 2. 1D Page 46 of 100 Question:(1 point)47ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Which of the following describes how the concern for potential feed line voiding and water hammer in the discharge piping of the CA pumps is addressed?

If a _____(1)_____ condition exists, an OAC alarm is generated.

To mitigate this condition, the operators will _____(2)_____ .

A. 1. high pressure 2. run the CA pump to flush and reseat the CA pump discharge check valve B. 1. high temperature 2. run the CA pump to flush and reseat the CA pump discharge check valve C. 1. high temperature 2. operate 1CF-105 (S/G Tempering Flow Supply Throttle) to increase tempering flow D. 1. high pressure 2. operate 1CF-105 (S/G Tempering Flow Supply Throttle) to increase tempering flow Page 47 of 100 Question:(1 point)48ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

Battery Charger 1ECC needs to be removed from service for maintenance

Standby Charger 1ECS will be aligned to supply Distribution Center 1EDC

In accordance with OP/1/A/6350/008 (125V/120VAC Vital Instrument and Control Power System), 1ECS must be powered from _____(1)_____ . This power supply alignment is required in Modes _____(2)_____ .

A. 1. 1EMXA 2. 1 - 4 B. 1. 1EMXA 2. 1 - 6 C. 1. 1EMXJ 2. 1 - 4 D. 1. 1EMXJ 2. 1 - 6 Page 48 of 100 Question:(1 point)49ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 2 is in Mode 4 performing a cooldown for refueling

Train 2A ND is in service

ND flow is stable at 3400 gpm

Which of the following describes:

1) the effect on NC cooldown rate of losing power to Vital DC Bus 2EPA; AND

2) what action the operator will take to restore the desired cooldown rate per AP/2/A/5500/029 (Loss of Vital or Aux Control Power)?

A. 1. Rate INCREASES 2. Manually adjust setpoint of 2ND-26 (ND HX 2A OUTLET CTRL)

B. 1. Rate DECREASES 2. Place PWR DISCON FOR 2NI-173A switch to THROT and manually control 2NI-173A C. 1. Rate INCREASES 2. Place PWR DISCON FOR 2NI-173A switch to THROT and manually control 2NI-173A D. 1. Rate DECREASES 2. Manually adjust setpoint of 2ND-26 (ND HX 2A OUTLET CTRL) Page 49 of 100 Question:(1 point)50ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP Subsequently:

A Blackout of 1ETA occurs

While loading, the Accelerated Sequence halts at Load Group 2

Assuming D/G voltage and speed permissives are met, the Accelerated Sequence _____(1)_____ re-initiate once the Committed Sequence completes Load Group 2.

Following completion of all Load Groups, the sequencer _____(2)_____ need to be reset in order to start the 1A KF Pump.

A. 1. will 2. does B. 1. will 2. does NOT C. 1. will NOT 2. does D. 1. will NOT 2. does NOT Page 50 of 100 Question:(1 point)51ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Concerning the operation of 1EMF-46A (Component Cooling Water):

When a Trip 2 alarm is received, 1KC-122 (KC Surge Tank Vent) _____________ automatically close.

When the associated KC train pumps are NOT running, the EMF loss of flow alarm ________________ blocked from being received in the control room.

A. 1. will 2. is B. 1. will 2. is NOT C. 1. will NOT 2. is D. 1. will NOT 2. is NOT Page 51 of 100 Question:(1 point)52ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Per EP/1/A/5000/ES-1.3 (Transfer to Cold Leg Recirculation), Enclosure 2 (Aligning NS for Recirculation), in order to provide cooling to the Containment Spray (NS) Heat Exchanger, the operator will align _____(1)_____ at a flowrate less than the MAXIMUM of _____(2)_____.

A. 1. RN 2. 5000 gpm B. 1. RN 2. 4650 gpm C. 1. KC 2. 5000 gpm D. 1. KC 2. 4650 gpm Page 52 of 100 Question:(1 point)53ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The BOP is currently performing actions of EP/1/A/5000/ES-1.3 (Transfer to Cold Leg Recirculation), Enclosure 2 (Aligning NS for Recirculation)

Containment pressure is 6 psig

Containment sump level is 5 feet

Upon opening 1NS-18A (NS Pump A Suct From Cont Sump), Containment Integrity status will change to _____(1)_____ .

In order to return Containment Integrity status to a "Yellow" condition, the BOP will be required to _____(2)_____ .

A. 1. red 2. start 1A NS pump ONLY B. 1. orange 2. start 1A NS pump ONLY C. 1. red 2. start 1A NS pump and align 1A NS HX cooling water flow D. 1. orange 2. start 1A NS pump and align 1A NS HX cooling water flow Page 53 of 100 Question:(1 point)54ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 and 2 are operating at 100% RTP

AP/0/A/5500/022 (Loss of Instrument Air) has been entered due to decreasing VI header pressure

Attempts to restore VI pressure have been unsuccessful Subsequently:

Both units are tripped per direction of AP/22

VI Pressure continues to decrease

Per OMP 1-7 (Emergency/Abnormal Procedure Implementation Guidelines), the actions of AP/22 will be continued _____(1)_____ .

As VI pressure decreases, AP/22 will FIRST dispatch operators to manually control _____(2)_____ .

Procedure Legend:

EP/1(2)/A/5000/E-0 (Reactor Trip or Safety Injection) EP/1(2)/A/5000/ES-0.1 (Reactor Trip Response)

A. 1. following actions to control key plant parameters in ES-0.1 2. CA feed flow B. 1. following actions to control key plant parameters in ES-0.1 2. NV charging flow C. 1. IMMEDIATELY following completion of Immediate Actions in E-0 2. CA feed flow D. 1. IMMEDIATELY following completion of Immediate Actions in E-0 2. NV charging flow Page 54 of 100 Question:(1 point)55ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Unit 1 event has occurred

Containment pressure is 3.8 psig and slowly decreasing

SI has not been reset

Containment isolation signals have not been reset

In order to re-establish Instrument Air to Containment:

Safety Injection _____(1)_____ required to be RESET.

Phase "B" _____(2)_____ required to be RESET.

A. 1. is 2. is B. 1. is 2. is NOT C. 1. is NOT 2. is D. 1. is NOT 2. is NOT Page 55 of 100 Question:(1 point)56ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 50% RTP

Control Rod Bank Select Switch is in "Auto"

Control Bank "D" is at 195 steps withdrawn Subsequently:

STIP-1 (Selected Turbine Impulse Pressure Input to Reactor Control System) fails to the 100% value

Assuming no operator action, Control Bank "D" will __________ .

A. withdraw 5 steps

B. withdraw 36 steps

C. remain at 195 steps

D. withdraw until Reactor Power reaches 103%

Page 56 of 100 Question:(1 point)57ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 tripped from 100% RTP due to a loss of Offsite Power.

The crew has transitioned to EP/1/A/5000/ES-0.3 (Natural Circulation Cooldown With Steam Void In Vessel) due to void formation in the Vessel Head

All PZR Heaters are OFF

Under these conditions, if charging flow is greater than letdown flow, PZR level will _____(1)_____ .

Vessel UR Level must be maintained greater than 68% in order prevent _____(2)_____ .

A. 1. decrease 2. fuel uncovery B. 1. decrease 2. void entry into hot legs C. 1. increase 2. fuel uncovery D. 1. increase 2. void entry into hot legs Page 57 of 100 Question:(1 point)58ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP Consider each statement separately If AFD exceeds specified limits, the OATC will be notified by an alarm on the _____(1)_____ .

If the AFD monitor function is not available, T.S. 3.2.3 (Axial Flux Difference) requires _____(2)_____ .

A. 1. Operator Aid Computer 2. a power reduction to <50%

B. 1. Operator Aid Computer 2. manual logging of % delta flux C. 1. Detector Current Comparator 2. a power reduction to <50%

D. 1. Detector Current Comparator 2. manual logging of % delta flux Page 58 of 100 Question:

(1 point)59ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Containment Hydrogen Recombiners receive power from a(n) _____(1)_____ Motor Control Center.

Containment Hydrogen Recombiners should not be placed in service if Containment Hydrogen concentration is greater than or equal to a MAXIMUM of _____(2)_____ .

A. 1. unit 2. 4%

B. 1. unit 2. 6%

C. 1. essential 2. 4%

D. 1. essential 2. 6% Page 59 of 100 Question:

(1 point)60ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 has tripped from 100% power

1A Rx Trip Breaker is OPEN

1B Rx Trip Breaker is CLOSED

The MAXIMUM NC temperature at which all Steam Dumps will be fully closed is _____(1)_____ degrees F.

Atmospheric Dumps _____(2)_____ operate following the Reactor Trip.

A. 1. 557 2. will B. 1. 557 2. will NOT C. 1. 560 2. will D. 1. 560 2. will NOT Page 60 of 100 Question:

(1 point)61ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% power.

Generator Breaker 1A spuriously opens.

The ONLY annunciator associated with the Turbine or Generator that alarms is: 1AD-11, F/1, GEN BKR B OVERCURRENT

Which of the following describes:

1) another annunciator which SHOULD have alarmed;

AND

2) a manual action required for mitigating these conditions?

A. 1. 1AD-1, F/4, TURB RUNBACK INITIATED 2. Manually initiate a turbine runback to 48%

B. 1. 1AD-1, F/4, TURB RUNBACK INITIATED 2. Manually trip Generator Breaker 1B C. 1. 1AD-11, F/7, GEN BKR B TROUBLE 2. Manually initiate a turbine runback to 48%

D. 1. 1AD-11, F/7, GEN BKR B TROUBLE 2. Manually trip Generator Breaker 1B Page 61 of 100 Question:

(1 point)62ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

A S/G tube leak has developed

If 1EMF-33 exceeds the _____(1)_____ setpoint, the BOP will ensure the _____(2)_____ system AUTOMATICALLY re-aligns (assuming NO operator action).

A. 1. Trip 1 2. S/G Blowdown (BB)

B. 1. Trip 1 2. Condenser Offgas (ZJ)

C. 1. Trip 2 2. S/G Blowdown (BB)

D. 1. Trip 2 2. Condenser Offgas (ZJ) Page 62 of 100 Question:

(1 point)63ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The Condensate System is aligned for High Pressure Cleanup Subsequently:

The 1A CFPT experiences a complete loss of oil pressure due to an LF Pump trip

The windmill protection circuit will send a direct signal to trip the operating _____(1)_____ .

Low CFPT oil pressure will cause the 1A CFPT _____(2)_____ to close.

A. 1. Hotwell Pumps 2. recirc valve B. 1. Hotwell Pumps 2. discharge valve C. 1. Condensate Booster Pumps 2. recirc valve D. 1. Condensate Booster Pumps 2. discharge valve Page 63 of 100 Question:

(1 point)64ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

1A1 Cont Floor & Equip Sump Pump has automatically started due to normal operational leakage Subsequently:

1EMF-53A (Containment Hi Range) loses power

As a result of this power failure:

1A1 Cont Floor & Equip Sump Pump is _____(1)_____ .

Entry in the action statement of Tech Spec 3.3.3 (Post Accident Monitoring Instrumentation) _____(2)_____ required.

A. 1. ON 2. is B. 1. ON 2. is NOT C. 1. OFF 2. is D. 1. OFF 2. is NOT Page 64 of 100 Question:

(1 point)65ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 4% RTP following startup o Main Turbine warming is in progress o 1A CFPT is in service o 1B CFPT is Tripped o All S/G levels are 39%

The crew has entered AP/1/A/5500/023 (Loss of Condenser Vacuum) due to a loss of Condenser Circulating Water (RC) Pumps Subsequently:

1A CFPT trips on low vacuum actuation

The OATC _____(1)_____ required to trip the reactor.

1A and 1B CA pumps _____(2)_____ automatically start.

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 65 of 100 Question:

(1 point)66ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Note the table below from SLC 16.5-3, Chemistry (excerpt). PARAMETER STEADY STATE LIMIT TRANSIENT LIMIT Dissolved Oxygen = 0.10 ppm = 1.00 ppm Chloride = 0.15 ppm = 1.50 ppm Fluoride = 0.15 ppm = 1.50 ppm

1) In accordance with SLC 16.5-3 (Chemistry), Fluoride is a _____(1)_____ chemistry parameter.

2) In accordance with the applicable abnormal operating procedure, if dissolved oxygen were at a value in Action Level 3, a shutdown to Mode 3 would be required in a MAXIMUM of _____(2)_____ hours.

A. 1. Primary 2. 3 B. 1. Primary 2. 24 C. 1. Secondary 2. 3 D. 1. Secondary 2. 24 Page 66 of 100 Question:

(1 point)67ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

1AD-7 C/4 "NCP Seal Water Lo Flow" alarms o The OAC indicates 1B NCP Seal Injection Flow is 0 gpm

In order to determine that seal water flow to the 1B NCP has been isolated, the BOP would verify an increase in flow on _____(1)_____ .

Following an actual loss of seal injection flow, the BOP _____(2)_____ expect to receive OAC alarm C1A0831 "NC Pump B Lower Brg Water Temp Hi-Hi".

A. 1. 1NVP5650 (Total Seal Wtr Flow) 2. will B. 1. 1NVP5650 (Total Seal Wtr Flow) 2. will NOT C. 1. 1NVP5330 (NCP 1A Seal In Flow) 2. will D. 1. 1NVP5330 (NCP 1A Seal In Flow) 2. will NOT Page 67 of 100 Question:

(1 point)68ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

During an outage, a planned maintenance activity will result in a YELLOW Defense in Depth (DID) sheet configuration.

In accordance with NSD 403 (Shutdown Risk Management):

Defense in Depth (DID) sheets are first initiated once _____(1)_____is reached on a shutdown.

A RISK MANAGEMENT PLAN is _____(2)_____ for a YELLOW risk condition.

A. 1. Mode 3 2. optional B. 1. Mode 3 2. required C. 1. Mode 4 2. optional D. 1. Mode 4 2. required Page 68 of 100 Question:

(1 point)69ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 3

In accordance with Tech Spec 2.1.2 (RCS Pressure Safety Limit), NC system pressure shall be less than or equal to a MAXIMUM of _____(1)_____ PSIG.

Based on the conditions above, if the NC system Safety Limit is exceeded, NC system pressure must be restored to within limits in a MAXIMUM of _____(2)_____ .

A. 1. 2485 2. 5 minutes B. 1. 2735 2. 5 minutes C. 1. 2485 2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. 1. 2735 2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Page 69 of 100 Question:

(1 point)70ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 6

A fuel handling accident involving fuel damage occurs o 1EMF-17 (Reactor Building Refueling Bridge) is in alarm

In accordance with AP/1/A/5500/025 (Damaged Spent Fuel), additional radiation monitoring will be provided by _____(1)_____ .

If this radiation monitor reaches Trip 2 setpoint, the Containment Evacuation Alarm _____(2)_____ .

A. 1. 1EMF-53A/B (Containment Train A/B) 2. will AUTOMATICALLY actuate B. 1. 1EMF-53A/B (Containment Train A/B) 2. must be MANUALLY actuated C. 1. 1EMF-38 (Containment Particulate Monitor) 2. will AUTOMATICALLY actuate D. 1. 1EMF-38 (Containment Particulate Monitor) 2. must be MANUALLY actuated Page 70 of 100 Question:

(1 point)71ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 3.

The NV system is being aligned for startup.

The procedure in use requires independent verification of a single valve located in a room with a general dose rate of 130 mREM/hr.

Estimated time to independently verify the valve's position is 10 minutes.

There are no known hot spots in the area.

There is no airborne activity in this room.

The room has no surface contamination areas.

Necessary approvals are obtained.

Per NSD 700 (Verification Techniques), independent verification of the valve above _____(1)_____ be waived because _____(2)_____ .

A. 1. may 2. the general area dose rate is GREATER than 100 mREM/hr B. 1. may NOT 2. the general area dose rate is LESS THAN 1 REM/hr C. 1. may 2. the radiation exposure for a single verification would EXCEED the allowable limit D. 1. may NOT 2. the radiation exposure for a single verification is WITHIN the allowable limit Page 71 of 100 Question:

(1 point)72ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Prior to a Refueling Outage, an Operator's exposure for the current year is 1250 mREM

The Operator's first assigned task during the Refueling Outage results in a dose of 600 mREM

Based on the conditions above, the next time the Operator logs on to the EDC computer, they will receive an _____(1)_____ notification flag.

In accordance with PD-RP-ALL-001 (RADIATION WORKER RESPONSIBILITIES), the Operator must _____(2)_____ prior to entering the RCA.

A. 1. ALERT 2. notify their supervisor AND obtain a dose extension B. 1. EXCLUDE 2. notify their supervisor AND obtain a dose extension C. 1. ALERT 2. notify their supervisor ONLY D. 1. EXCLUDE 2. notify their supervisor ONLY Page 72 of 100 Question:

(1 point)73ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A large fire was reported at the Operations Training Building

1) Which of the following describes the availability of the Fire Brigade to respond to this fire? 2) Which procedure contains the guidance for requesting offsite fire department support?

A. 1. The fire brigade is not allowed to respond to fires outside the Protected Area 2. AP/0/A/5500/045 (Plant Fire)

B. 1. The fire brigade is not allowed to respond to fires outside the Protected Area 2. RP/0/B/5000/029 (Fire Brigade Response)

C. 1. The fire brigade may respond to fires outside the Protected Area as resources permit 2. AP/0/A/5500/045 (Plant Fire)

D. 1. The fire brigade may respond to fires outside the Protected Area as resources permit 2. RP/0/B/5000/029 (Fire Brigade Response) Page 73 of 100 Question:

(1 point)74ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

1A S/G has experienced a Tube Rupture

The crew has transitioned to EP/1/A/5000/E-3 (Steam Generator Tube Rupture)

NC pressure is 1650 psig

The CRS has just read a series of NOTES prior to E-3, Step 10 "Initiate NC System cooldown as follows:"

Once the cooldown is initiated:

the OATC will dump steam while maintaining steam pressure negative rate less than 2 psig per second in order to prevent a _____(1)_____ .

if NC system subcooling is lost, the crew _____(2)_____ secure NC pumps.

A. 1. main steam isolation 2. will B. 1. main steam isolation 2. will NOT C. 1. S/G lo-lo level actuation 2. will D. 1. S/G lo-lo level actuation 2. will NOT Page 74 of 100 Question:

(1 point)75ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Fire alarm has actuated in the Unit 1 CA Pump Room

An Operator dispatched to the area reports that there is smoke and some cables with glowing embers but, NO visible flames

In accordance with AP/1/A/5500/045 (Plant Fire), this _____(1)_____ classified as an ACTIVE fire.

In accordance with RP/0/B/5000/029 (Fire Brigade Response), in addition to making an announcement on the Fire Brigade Radio and activating the Fire Brigade Pagers, a Plant PA announcement _____(2)_____ required when dispatching the Fire Brigade.

A. 1. is 2. is B. 1. is 2. is NOT C. 1. is NOT 2. is D. 1. is NOT 2. is NOT Page 75 of 100 Question:

(1 point)76ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 4

1A and 1B CA Pumps are aligned for standby readiness

In service: o 1A ND Pump o 1B NC Pump

Removed from service per R&R: o 1B ND Pump o 1A, 1C, and 1D NC Pumps Subsequently:

1B NCP Supply Breaker fails o The crew has entered the actions of LCO 3.4.6 (RCS Loops - Mode 4)

S/G Levels are as follows: S/G 1A 1B 1C 1D NR Level 11% 14% 12% 11% Consider each statement separately Restoring 1B ND pump _____(1)_____ allow the crew to exit the actions of LCO 3.4.6.

Restoring 1A NC pump _____(2)_____ allow the crew to exit the actions of LCO 3.4.6.

A. 1. will 2. will B. 1. will 2. will NOT C. 1. will NOT 2. will D. 1. will NOT 2. will NOT Page 76 of 100 Question:

(1 point)77ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 was at 100% RTP when the Main Turbine tripped

Automatic and Manual Reactor Trips have failed

The crew has entered EP/1/A/5000/FR-S.1 (Response to Nuclear Power Generation/ATWS) o Emergency Boration has been initiated Subsequently:

The Reactor Trip and Bypass Breakers are opened locally

In order to initiate Emergency Boration, a MINIMUM of _____(1)_____ NV Pump(s) must be in operation.

The crew will transition to EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) following verification that _____(2)_____ .

A. 1. one 2. the reactor is subcritical ONLY B. 1. two 2. the reactor is subcritical ONLY C. 1. one 2. SDM is verified per Reactivity Balance Procedure D. 1. two 2. SDM is verified per Reactivity Balance Procedure Page 77 of 100 Question:

(1 point)78ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 has experienced a Loss of Offsite Power

1B D/G failed to start

Safety Injection was initiated due to decreasing PZR level following Reactor Trip

Given the following parameters o All CA flow control valves are closed o All S/G pressures are approximately 1125 psig and stable o 1A S/G NR level is 67% and increasing o 1B, 1C, & 1D S/G NR levels are 39% and slowly decreasing o NC Subcooling is 8ºF and stable Which of the following describes the correct classification for this event in accordance with RP/0/A/5000/001 (Classification of Emergency)?

REFERENCE PROVIDED A. 4.1.S.2

B. 4.1.S.3

C. 4.1.A.1

D. 4.5.A.2 Page 78 of 100 Question:

(1 point)79ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 was at 100% power when a total loss of onsite and offsite power occurred.

(1) Which procedure contains voltage values on essential DC busses (EDA, EDB, EDC, and EDD) requiring essential batteries (EBA, EBB, EBC, EBD) to be removed from service?

(2) After power is restored and the battery chargers are placed in service, in accordance with Tech Spec 3.8.4 (DC Sources - Operating), what is the MINIMUM voltage required for the essential batteries to be OPERABLE?

Procedure Legend:

AP/1/A/5500/007 (Loss of Normal Power)

AP/1/A/5500/029 (Loss of Vital or Aux Control Power)

A. 1. AP/07 2. 125 volts B. 1. AP/29 2. 125 volts C. 1. AP/07 2. 110 volts D. 1. AP/29 2. 110 volts Page 79 of 100 Question:

(1 point)80ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

The TCC has reported that "Real Time Contingency Analysis" (RTCA) indicates INADEQUATE switchyard voltage

The crew has entered AP/1/A/5500/037 (Generator Voltage and Electric Grid Disturbances)

Main Generator operating conditions are as follows o Hydrogen Pressure (psig) 73 o Generator VARS 750 o Generator MW 1200

In accordance with AP/37:

the CRS will direct the OATC to _____(1)_____ .

once required jumpers are placed, both trains of offsite power _____(2)_____ remain inoperable.

REFERENCE PROVIDED A. 1. decrease turbine load 2. do B. 1. decrease turbine load 2. do NOT C. 1. decrease generator voltage 2. do D. 1. decrease generator voltage 2. do NOT Page 80 of 100 Question:

(1 point)81ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A Unit 1 event has occurred from 100% RTP

NC Pressure is 300 psig

All S/G Pressures are 700 psig

Containment sump level is currently 3.4 feet and slowly increasing

EP/1/A/5000/ES-1.3 (Transfer to Cold Leg Recirculation) has been implemented

1NI-185A (ND Pump 1A Cont Sump Suct) AND 1NI-184B (ND Pump 1B Cont Sump Suct) will NOT open

Implementation of _____(1)_____ is required.

The implemented procedure will direct the crew to _____(2)_____ .

Procedure Legend:

EP/1/A/5000/ECA-1.1 (Loss of Emergency Coolant Recirculation) EP/1/A/5000/ECA-1.3 (Containment Sump Blockage)

A. 1. ECA-1.1 2. depressurize S/Gs to cooldown the NC system B. 1. ECA-1.1 2. depressurize the NC system to increase SI flow C. 1. ECA-1.3 2. depressurize S/Gs to cooldown the NC system D. 1. ECA-1.3 2. depressurize the NC system to increase SI flow Page 81 of 100 Question:

(1 point)82ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Unit 1 is operating at 98% power. PT/1/A/4600/001 (RCCA Movement Test) is in progress. As Control Bank D was being moved, one control rod in Control Bank D slipped to 120 steps withdrawn and stopped. Below is an incore thermocouple map one minute later.

R P N M L K J H G F E D C B A 1 576 572 576 2 546 599 611 603 605 3 605 624 BAD 569 4 597 602 609 617 5 602 625 616 614 6 563 611 611 626 623 568 7 602 606 617 620 8 561 612 634 572 626 9 619 611 613 614 10 548 610 606 613 574 11 627 BAD 610 614 12 612 599 608 622

13 620 619 576 626 562 14 540 628 613 608 614 15 BAD 622 580

Rod _____(1)_____ is misaligned.

In order to continue Mode 1 operation, Tech. Spec. 3.1.4 (Rod Group Alignment Limits) will require a power reduction, SDM verification, and completion of _____(2)_____ .

A. 1. D-12 2. FNDH(X,Y) surveillance ONLY B. 1. M-4 2. FNDH(X,Y) surveillance ONLY C. 1. D-12 2. FNDH(X,Y) AND FQ(X,Y,Z) surveillances D. 1. M-4 2. FNDH(X,Y) AND FQ(X,Y,Z) surveillances Page 82 of 100 Question:

(1 point)83ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 6 offloading fuel

N-31 and N-32 are in service

1A and 1B BDMS are in service Subsequently:

Source Range Instrument (N-31) fails

The CRS _____(1)_____ required to enter the action statement of LCO 3.9.2 (Nuclear Instrumentation).

In order to meet the operability requirements of LCO 3.9.2, source range audible indication _____(2)_____ required.

A. 1. is 2. is B. 1. is 2. is NOT C. 1. is NOT 2. is D. 1. is NOT 2. is NOT Page 83 of 100 Question:(1 point)84ILT15 CNS SRO NRC ExaminationCatabwa Nuclear StationGiven the following: *A Unit 1 event has occurred*ES-1.3, the OATC reports the following Status Indicators /Annunciators:oReactor Coolant Integrity Status is REDo1AD-9 B/6 "ND Trn B To C-Legs Loops A-B Lo Flow" is LIT due to trip of1B ND PumpThe CRS is required to implement __________ . Procedure Legend: EP/1A/5000/ES-1.3 (Transfer to Cold Leg Recirculation) EP/1A/5000/FR-P.1 (Response to Imminent Pressurized Thermal Shock) EP/1A/5000/ECA-1.1 (Loss of Emergency Coolant Recirculation) A. FR-P.1 IMMEDIATELY B. FR-P.1 at step 8 of ES-1.3 C. ECA-1.1 IMMEDIATELY D. ECA-1.1 at step 4 of ES-1.3 Page 84 of 100 Question:

(1 point)85ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 was initially at 100% power.

The reactor tripped due to a spurious turbine trip.

The crew has performed and exited EP/1/A/5000/E-0, (Reactor Trip or Safety Injection).

1B S/G pressure is being maintained at approximately 1210 psig.

The MAXIMUM number of S/G safety relief valves (associated with 1B S/G) that did NOT function as designed is _____(1)_____ .

To mitigate this event, the CRS would enter _____(2)_____ .

Procedure Legend:

EP/1/A/5000/FR-H.2, (Response to Steam Generator Overpressure) EP/1/A/5000/FR-H.4, (Response to Loss of Normal Steam Release Capabilities)

A. 1. 3 2. FR-H.2 B. 1. 5 2. FR-H.2 C. 1. 3 2. FR-H.4 D. 1. 5 2. FR-H.4 Page 85 of 100 Question:

(1 point)86ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

Containment Pressure Channel Two has failed low

All required Tech Spec actions have been completed Subsequently:

A faulty Containment Pressure bistable has initiated a Reactor Trip and Safety Injection The faulty bistable initiating this event was associated with Containment Pressure Channel _____(1)_____ .

The CRS will direct the SI termination sequence per _____(2)_____ .

A. 1. One 2. EP/1/A/5000/E-0 (Reactor Trip or Safety Injection)

B. 1. One 2. EP/1/A/5000/ES-1.1 (Safety Injection Termination)

C. 1. Four 2. EP/1/A/5000/E-0 (Reactor Trip or Safety Injection)

D. 1. Four 2. EP/1/A/5000/ES-1.1 (Safety Injection Termination) Page 86 of 100 Question:

(1 point)87ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The crew has entered EP/1/A/5000/ECA-2.1 (Uncontrolled Depressurization of All Steam Generators) following a Unit 1 event

All attempts to close MSIVs from the Control Room have failed Subsequently:

The crew has reached the step (in ECA-2.1) just prior to verification of S/I termination criteria

1A MSIV indicates closed on the main control board

1A S/G pressure is slowly increasing In accordance with ECA-2.1, the crew is required to transition to EP/1/A/5000/E-2 (Faulted Steam Generator Isolation) _____(1)_____ based on indication of _____(2)_____ .

A. 1. immediately 2. increasing S/G pressure B. 1. following completion of S/I termination in ECA-2.1 2. increasing S/G pressure C. 1. immediately 2. MSIV closure D. 1. following completion of S/I termination in ECA-2.1 2. MSIV closure Page 87 of 100 Question:

(1 point)88ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 4

1A ND train is in RHR

CA system aligned per T.S. minimum requirements Subsequently:

PRT level is increasing

PZR level is decreasing

PZR Relief Valve and Safety Valve Temperatures indicate 105 ºF and stable

AP/1/A/5500/027 (Shutdown LOCA) has been entered

Following completion of applicable steps in AP/27, the CRS will transition to _____(1)_____ .

If needed, additional heat removal capability will be available via operation of _____(2)_____ CA Pump(s).

A. 1. AP/1/A/5500/010 (Reactor Coolant Leak) 2. one B. 1. AP/1/A/5500/010 (Reactor Coolant Leak) 2. two C. 1. AP/1/A/5500/019 (Loss of Residual Heat Removal System) 2. one D. 1. AP/1/A/5500/019 (Loss of Residual Heat Removal System) 2. two Page 88 of 100 Question:

(1 point)89ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

1000 Both Units are at 100% RTP 1AD-11 E/7 "D/G B Panel Trouble" is received Dispatched operator reports B/1 "Low Level Lube Tank" in alarm Operator also reports leak at D/G Lube Oil Tank The crew enters the actions of LCO 3.8.1 (AC Sources - Operating) 1005 The crew enters OP/0/A/6400/006C, Enclosure 4.11 and then transitions to Enclosure 4.12B 1020 Enclosure 4.12B alignment is completed At 1005 Unit 2 _____(1)_____ in LCO 3.7.8 (Nuclear Service Water System), Condition "A" (One NSWS train inoperable).

At 1020 Unit 2 _____(2)_____ in LCO 3.7.8 (Nuclear Service Water System), Condition "A" (One NSWS train inoperable).

Procedure Legend for OP/0/A/6400/006C (Nuclear Service Water System): Enclosure 4.11 (Operability Actions With One RN Pump and/or Its Associated D/G Inoperable With Both Units Entering an Action Statement) Enclosure 4.12B (Alignment For Single Pump Flow Balance Due To One Train B RN Pump and/or Its Associated D/G Inoperable)

A. 1. is 2. is B. 1. is 2. is NOT C. 1. is NOT 2. is D. 1. is NOT 2. is NOT Page 89 of 100 Question:(1 point)90ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

The VI system on Unit 1 has become heavily contaminated with oil * "E" VI Dryer has become clogged

The crew enters AP/0/A/5500/022 (Loss of Instrument Air)

Backup VI Compressors CANNOT be started Subsequently:

The VI system is restored

A Temporary Engineering change is required to align alternate air supply to CA Flow Control valves for system oil removal

During performance of AP/22, the CRS FIRST directed operators to verify proper operation of _____(1)_____ as pressure decreased.

The required Temporary Engineering Change _____(2)_____ require a 10CFR50.59 evaluation.

A. 1. 1VI-670 (VI Dryer Auto Bypass) 2. does B. 1. 1VI-670 (VI Dryer Auto Bypass) 2. does NOT C. 1. 1VI-487 (Skid "E" Purge Exhaust to Atmos) 2. does D. 1. 1VI-487 (Skid "E" Purge Exhaust to Atmos) 2. does NOT Page 90 of 100 Question:(1 point)91ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

1AD-9 E/1 "Accum Tank A Hi/Lo Press" has alarmed

The BOP has determined 1A Cold Leg Accumulator Pressure is low The CRS will direct a Nitrogen makeup to the 1A CLA in order to ensure the T.S. 3.5.1 (Accumulators) MINIMUM required pressure of _____(1)_____ psig is not violated.

A MINIMUM of _____(2)_____ operable CLAs will sufficiently cover the core prior to significant fuel clad melting following a LOCA.

A. 1. 585 2. Three B. 1. 585 2. Four C. 1. 678 2. Three D. 1. 678 2. Four Page 91 of 100 Question:(1 point)92ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

Channel 4 PZR Pressure has failed low

All required T.S. actions have been completed Subsequently:

1B NC Loop Tcold experiences a power supply failure

Channel 2 OTdeltaT setpoint will _____(1)_____ .

Assuming no other actions are taken, the latest time that Unit 1 will be required to reach Mode 3 is in _____(2)_____ hours.

A. 1. increase 2. 7 B. 1. decrease 2. 7 C. 1. increase 2. 78 D. 1. decrease 2. 78 Page 92 of 100 Question:(1 point)93ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Concerning operation of the Reactor Building Manipulator Crane:

In order to prevent dropping a fuel assembly, the load cell must indicate less than _____(1)_____ in order to operate the Gripper.

In accordance with MP/1/A/7150/026 B (Unit 1 Reactor Manipulator Crane Operation), any Interlock bypass, not approved by procedure, requires a MINIMUM of _____(2)_____ .

A. 1. 500 lbs 2. Fuel Handling SRO approval ONLY B. 1. 500 lbs 2. Fuel Handling SRO approval AND CRS notification when bypassed and restored C. 1. 1200 lbs 2. Fuel Handling SRO approval ONLY D. 1. 1200 lbs 2. Fuel Handling SRO approval AND CRS notification when bypassed and restored Page 93 of 100 Question:(1 point)94ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Concerning Fuel Handling requirements:

PT/1/A/4550/001C (Refueling Communications Test) is required to be performed within a MAXIMUM of _____(1)_____ of commencing fuel movement.

In order to unlatch Control Rods per MP/0/A/7150/067 (Rod Cluster Control Assembly Drive Rod Latching & Unlatching), the Refueling SRO _____(2)_____ required to be in the Reactor Building.

A. 1. 30 minutes 2. is B. 1. 30 minutes 2. is NOT C. 1. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 2. is D. 1. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 2. is NOT Page 94 of 100 Question:(1 point)95ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is at 100% RTP

Unit 2 is in Mode 4

In order to meet the MINIMUM requirements of OMP 1-10 (Shift Manning and Overtime Requirements), _____(1)_____ licensed SROs will be onsite at all times.

Per NSD 200 (Work Hour Guidelines and Limits), all hours worked during a declared emergency _____(2)_____ be included in future work hour calculations.

A. 1. 4 2. will B. 1. 4 2. will NOT C. 1. 5 2. will D. 1. 5 2. will NOT Page 95 of 100 Question:(1 point)96ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station For high activity in the reactor coolant, to ensure that the requirements of Tech. Spec. 3.4.16, (RCS Specific Activity) are met, the SRO is ALSO required to apply the requirements of AP/1/A/5500/018, (High Activity in Reactor Coolant) for NC System Dose Equivalent _____(1)_____ .

These RCS Specific Activity limits are based on the _____(2)_____ Design Basis Accident.

A. 1. I-131 2. LOCA outside containment B. 1. I-131 2. Steam Generator Tube Rupture C. 1. Xe-133 2. LOCA outside containment D. 1. Xe-133 2. Steam Generator Tube Rupture Page 96 of 100 Question:(1 point)97ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 6

1AD-4 B/1 (S/G A Level Deviation) will not illuminate. IAE has Blocked the inputs for outage related activities

1AD-9 B/7 (FWST Emerg Lo Temp) will not illuminate. IAE has disconnected the associated transmitter in order to temporarily change the alarm setpoint for an upcoming test

In accordance with OMP 2-31(Control Room Instrumentation Status):

OMP 2-31, Attachment 1 (TMs Affecting Control Room Annunciators) will be used to track the inoperability of _____(1)_____ .

Attachment 1 will be filed in the _____(2)_____ .

A. 1. 1AD-4 B/1 2. Ops Shift Routine Logbook B. 1. 1AD-9 B/7 2. Ops Shift Routine Logbook C. 1. 1AD-4 B/1 2. Shift Work Manager Logbook D. 1. 1AD-9 B/7 2. Shift Work Manager Logbook Page 97 of 100 Question:(1 point)98ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

Unit 1 is in Mode 3 following a refueling outage

The status of the Personnel Air Locks (PAL) is as follows: o Upper Airlock Inner Door Operable o Upper Airlock Outer Door Operable o Lower Airlock Inner Door Inoperable o Lower Airlock Outer Door Operable

Repairs required are on the barrel (airlock side of the inner door)

The guidance for Containment entry, to repair the Lower Airlock Door, is contained in _____(1)_____ .

The Lower Airlock Outer Door _____(2)_____ be opened to make the repair.

A. 1. Tech. Spec. 3.6.2, (Containment Air Locks) 2. may B. 1. Tech. Spec. 3.6.2, (Containment Air Locks) 2. may NOT C. 1. Site Directive 3.1.2, (Access to Reactor Building and Areas Having High Pressure Steam Relief Devices) 2. may D. 1. Site Directive 3.1.2, (Access to Reactor Building and Areas Having High Pressure Steam Relief Devices) 2. may NOT Page 98 of 100 Question:(1 point)99ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station For a planned release of the contents of a Monitor Tank to the Low Pressure Service Water discharge (RL) via the Nuclear Service Water System (RN), which of the following describes a condition that would PREVENT approval of the release?

A. The pH of the contents is 8.8. B. A boron release with a boron concentration at 480 ppm C. Planned release flowrate is 95 gpm D. RN is aligned to the Standby Nuclear Service Water Pond Page 99 of 100 Question:(1 point)100ILT15 CNS SRO NRC ExaminationCatabwa Nuclear Station Given the following:

A General Emergency has been declared.

The TSC, OSC, and EOF have NOT been activated.

In accordance with RP/0/A/5000/005 (TSC Activation Procedure):

The SM/Emergency Coordinator's responsibility of making Protective Action Recommendations _____(1)_____ be delegated.

Turnover of command and control to the TSC or EOF_____(2)_____ relieve the SM of classification, notification, and Protective Action Recommendation (PAR) responsibilities.

A. 1. can 2. does B. 1. can 2. does NOT C. 1. can NOT 2. does D. 1. can NOT 2. does NOT Page 100 of 100 Examination KEY for:ILT15 CNS SRO NRC ExaminQuestionNumberAnswer1C 2C 3B 4C 5C 6C 7C 8D 9A10C 11C 12B 13D 14B 15B 16D 17A 18D 19A 20A 21D 22B 23B 24C 25CPrinted 5/6/2015 2:51:19 PMPage 1 of 4 Examination KEY for:ILT15 CNS SRO NRC ExaminQuestionNumberAnswer26C 27B 28B 29C 30B 31A 32C 33C 34C 35A 36B 37D 38A 39C 40B 41D 42B 43B 44D 45D 46D 47B 48B 49C 50BPrinted 5/6/2015 2:51:19 PMPage 2 of 4 Examination KEY for:ILT15 CNS SRO NRC ExaminQuestionNumberAnswer51C 52B 53D 54B 55C 56A 57B 58B 59D 60D 61A 62C 63A 64D 65C 66A 67D 68C 69B 70C 71C 72B 73D 74B 75CPrinted 5/6/2015 2:51:19 PMPage 3 of 4 Examination KEY for:ILT15 CNS SRO NRC ExaminQuestionNumberAnswer76B 77A 78B 79B 80C 81A 82D 83D 84B 85C 86C 87A 88C 89B 90C 91A 92A 93B 94C 95C 96B 97B 98A 99D100CPrinted 5/6/2015 2:51:19 PMPage 4 of 4

Unit 1 Data BookSource: CNM-1300.00153, Fig. 17-3, Dwg 435HA148 (Rev. 2)Figure 43 - Generator Capability CurvesPage 1-1000-750-500-25002505007501000125015000200400600800100012001400160018001000 kVARS1000 kilowatts30 PSIG45 PSIG60 PSIG75 PSIG0.6 PF0.7 PF0.8 PF0.85 PF0.9 PF0.95 PF0.98 PF0.95 PF0.9 PF0.85 PF0.8 PF0.7 PF0.6 PFATB 4 POLE, 1,450,000 KVA, 1800 RPM, 2200 VOLTS0.9 PF, 0.50 SCR, 75 PSIG HYDROGEN PRESSURE, 545 VOLTS EXCITATIONAAAABBBBCCCCD Unit 1 Data BookSource: CNM-1300.00153, Fig. 17-3, Dwg 435HA148 (Rev. 2)Figure 43 - Generator Capability CurvesPage 1-1000-750-500-25002505007501000125015000200400600800100012001400160018001000 kVARS 1000 kilowatts 30 PSIG45 PSIG60 PSIG75 PSIG0.6 PF 0.7 PF 0.8 PF 0.85 PF 0.9 PF 0.95 PF 0.98 PF 0.95 PF 0.9 PF 0.85 PF 0.8 PF 0.7 PF 0.6 PF ATB 4 POLE, 1,450,000 KVA, 1800 RPM, 2200 VOLTS 0.9 PF, 0.50 SCR, 75 PSIG HYDROGEN PRESSURE, 545 VOLTS EXCITATION A A A A B B B B C C C C D Procedure No. Revision No. Electronic Reference No. PERFORMANCE RP/0/A/5000/001 Page 2 of 4 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 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 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/5000/001 Page 3 of 4 2. Immediate Actions 2.1 IF performing this procedure due to security-related event(s) considered to be a CREDIBLE THREAT or HOSTILE ACTION, perform the following.

IF Security reports that a SECURITY CONDITION or HOSTILE ACTION is imminent (15 minutes) or in-progress, notify station personnel via the plant page to take the appropriate protective actions. Refer to RP/0/B/5000/026, Site Response to Security Events, as soon as possible for scripted message.

Perform accelerated notification to the NRC within 15 minutes of the occurrence of the event. 2.2 Assessment, classification and declaration of any applicable emergency condition should be completed within 15 minutes after the availability of indications or information to COGNIZANT FACILITY STAFF that an EAL threshold has been exceeded. (Refer to Enclosure 4.9, Emergency Declaration Guidelines, as needed) 2.3 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. 2.4 IF the plant was in Mode 1-4 and a valid condition affects fission product barriers, proceed to Enclosure 4.1 (Fission Product Barrier Matrix). 2.5 IF a General Emergency is NOT declared in Step 2.4 OR the condition does not affect fission product barriers, review the listing of enclosures to determine if the event is applicable to one the categories shown. 2.6 Compare actual plant conditions to the Emergency Action Levels evaluated in step 2.4 and/or 2.5 and declare the appropriate Emergency Class as indicated. 2.7 Document the declaration time. __________________ 2.8 Activate the ERO per the appropriate Response Procedure (RP) utilizing the Control Room ERO Notification Job Aid. 2.9 IF the declaration is made in the Control Room, announce the classification and declaration time to the Control Room Crew at the first opportunity that will not interfere with the performance of the crew or the flow of the Emergency Procedure. 2.10 IF the declaration is made in the TSC or EOF, announce the classification and declaration time to the applicable facility personnel. _____ _____ _____ _____ _____ _____ _____ _____ _____ _____

RP/0/A/5000/001 Page 4 of 4 2.11 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/5000/002 Alert RP/0/A/5000/003 Site Area Emergency RP/0/A/5000/004 General Emergency RP/0/A/5000/005 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 4.9. 3.2 Refer to enclosure 4.9, Emergency Declaration Guidelines, as needed. 3.3 Refer to Section D of the Catawba Emergency Plan for basis information about the Emergency Classification System as needed. 3.4 Refer to RP/0/A/5000/020, "TSC Activation Procedure" concerning the use of 10CFR50.54(x). If the TSC is activated, contact the TSC Emergency Coordinator (EC) for concurrence when using 10CFR50.54(x). 4. 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 Fires/Explosions 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 Reading for Enclosure 4.3 EALs _____ _____ _____ _____ _____

Enclosure 4.1 RP/0/A/5000/001 Fission Product Barrier Matrix Page 1 of 5 1. Use EALs to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all barriers. Classify according to the table below. Note 1: 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 judgment and classify as if the thresholds are exceeded. Note 2: 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, even if the conditions do not currently exist. Note 3: 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 that a CSF has been validated prior to the CSF being used as a basis to classify an emergency. Example: If ECA-0.0, Loss of All AC Power Procedure, is implemented with an appropriate CSF alarm condition valid and sustained, the 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. IC Unusual Event IC Alert IC Site Area Emergency IC General Emergency 4.1.U.1 Potential Loss of Containment 4.1.A.1 Loss OR Potential Loss of Nuclear Coolant System 4.1.S.1 Loss OR Potential Loss of Both Nuclear Coolant System AND Fuel Clad 4.1.G.1 Loss of All Three Barriers 4.1.U.2 Loss of Containment 4.1.A.2 Loss OR Potential Loss of Fuel Clad 4.1.S.2 Loss AND Potential Loss Combinations of Both Nuclear Coolant System AND Fuel Clad 4.1.G.2 Loss of Any Two Barriers AND Potential Loss of the Third 4.1.A.3 Potential Loss of Containment AND Loss OR Potential Loss of Any Other Barrier 4.1.S.3 Loss of Containment AND Loss OR Potential Loss of Any Other Barrier Enclosure 4.1 RP/0/A/5000/001 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 table 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 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 "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 5, under one of the four "classification" columns, select the event number (e.g. 4.1.A.1 for Loss of Nuclear Coolant System) that best fits the loss of barrier descriptions. 7. Using the number (e.g. 4.1.A.1), select the preprinted notification form OR a blank notification form and complete the required information for Emergency Coordinator approval and transmittal.

Enclosure 4.1 RP/0/A/5000/001 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 - (1 Point) LOSS - (3 Points) POTENTIAL LOSS - (4 Points) LOSS - (5 Points) POTENTIAL LOSS - (4 Points) LOSS - (5 Points) 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status

Containment-RED

Core cooling-RED Path is indicated for >15 minutes

Not applicable

NCS Integrity-Red

Heat Sink-Red

Not applicable

Core Cooling-Orange

Heat Sink-Red

Core Cooling-Red 2. Containment Conditions 2. NCS Leak Rate 2. Primary Coolant Activity Level

Containment Pressure > 15 PSIG

H2 concentration > 9%

Containment pressure greater than 3 psig with less than one full train of NS and a VX-CARF operating after actuation.

Rapid unexplained decrease in containment pressure following initial increase

Containment pressure or sump level response not consistent with LOCA conditions. NOTE: Refer to Emergency Plan, Sect. D, 4.1.C.2, last paragraph for inability to maintain normal annulus pressure.

Unisolable leak exceeding the capacity of one charging pump in the normal charging mode with letdown isolated.

GREATER THAN available makeup capacity as indicated by a loss of NCS subcooling.

Not applicable

Coolant Activity GREATER THAN 300 µCi/cc Dose Equivalent Iodine (DEI) I-131 CONTINUED CONTINUED CONTINUED Enclosure 4.1 RP/0/A/5000/001 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 - (1 Point) LOSS - (3 Points) POTENTIAL LOSS - (4 Points) LOSS - (5 Points) POTENTIAL LOSS - (4 Points) LOSS - (5 Points) 3. Containment Isolation Valves Status After Containment Isolation Actuation 3. SG Tube Rupture 3. Containment Radiation Monitoring

Not applicable

Containment isolation is incomplete and a direct release path from containment exists to the environment

Primary-to-Secondary leak rate exceeds the capacity of one charging pump in the normal charging mode with letdown isolated.

Indication that a SG is Ruptured and has a Non-Isolable secondary line fault

Indication that a SG is ruptured and a prolonged release of contaminated secondary coolant is occurring from the affected SG to the environment

Not applicable

Containment radiation monitor 53 A or 53 B Reading at time since Shutdown. 0-0.5 hrs > 99 R/hr 0.5-2 hrs > 43 R/hr 2-4 hrs > 31 R/hr 4-8 hrs > 22 R/hr

>8 hrs > 13 R/hr 4. SG Secondary Side Release With Primary-to-Secondary Leakage 4. Containment Radiation Monitoring 4. Emergency Coordinator/EOF Director Judgement

Not applicable

Release of secondary side to the environment with primary to secondary leakage GREATER THAN Tech Spec allowable

Not applicable

Not applicable

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 fuel clad barrier. END CONTINUED CONTINUED Enclosure 4.1 RP/0/A/5000/001 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 - (1 Point) LOSS - (3 Points) POTENTIAL LOSS - (4 Points) LOSS - (5 Points) POTENTIAL LOSS - (4 Points) LOSS - (5 Points) 5. Significant Radioactive Inventory In Containment 5. Emergency Coordinator/EOF Director Judgement

Containment Rad. Monitor EMF53A or 53B Reading at time since shutdown:

0 - 0.5 hr > 390 R/hr 0.5 - 2 hr > 170 R/hr 2 - 4 hr > 125 R/hr 4 - 8 hr > 90 R/hr

> 8 hr > 53 R/hr

Not applicable

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 NCS barrier. END 6. Emergency Coordinator /EOF Director Judgment

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/5000/001 System Malfunctions Page 1 of 2 UNUSUAL EVENT 4.2.U.1 Inability to Reach Required Shutdown Within Technical Specification Limits. OPERATING MODE: 1, 2, 3, 4 4.2.U.1-1 Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time. 4.2.U.2 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes. OPERATING MODE: 1 , 2, 3, 4 4.2.U.2-1 The following conditions exist: Unplanned loss of most (>50%) annunciators associated with safety systems for greater than 15 minutes. AND In the opinion of the Operations Shift Manager/Emergency Coordinator/EOF Director, the loss of the annunciators or indicators requires additional personnel (beyond normal shift compliment) to safely operate the unit. CONTINUED ALERT 4.2.A.1 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2)

Compensatory Non-Alarming Indicators Unavailable. OPERATING MODE: 1, 2, 3, 4 4.2.A.1-1 The following conditions exist: Unplanned loss of most (>50%) annunciators associated with safety systems for greater than 15 minutes. AND In the opinion of the Operations Shift Manager/Emergency Coordinator/EOF Director, the loss of the annunciators or indicators requires additional personnel (beyond normal shift compliment) to safely operate the unit. AND EITHER of the following:

A SIGNIFICANT PLANT TRANSIENT is in progress

Loss of the OAC. END SITE AREA EMERGENCY 4.2.S.1 Inability to Monitor a Significant Transient in Progress. OPERATING MODE: 1, 2, 3, 4 4.2.S.1-1 The following conditions exist: Loss of most (>50%)

Annunciators associated with safety systems. AND A SIGNIFICANT PLANT TRANSIENT is in progress. AND Loss of the OAC. AND Inability to provide manual monitoring of any of the following Critical Safety Functions:

subcriticality

core cooling

heat sink

containment. END GENERAL EMERGENCY END Enclosure 4.2 RP/0/A/5000/001 System Malfunctions Page 2 of 2 UNUSUAL EVENT 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 Specifications allowable limit. 4.2.U.4 Reactor Coolant System (NCS) Leakage. OPERATING MODE: 1, 2, 3, 4 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 ALERT SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.3 RP/0/A/5000/001 Abnormal Rad Levels/Radiological Effluent Page 1 of 5 UNUSUAL EVENT 4.3.U.1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the SLC Limits for 60 Minutes or Longer. OPERATING MODE: ALL 4.3.U.1-1 A valid Trip 2 alarm on radiation monitor EMF-49L or EMF-57 for > 60 minutes or will likely continue for > 60 minutes which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure HP/0/B/1009/014.

4.3.U.1-2 A valid indication on radiation monitor EMF- 36L of 3.00E+04 cpm for > 60 minutes or will likely continue for 60 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure AD-EP-ALL-0202. (Continued) ALERT 4.3.A.1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the SLC limits for 15 Minutes or Longer. OPERATING MODE: ALL 4.3.A.1-1 A valid indication on radiation monitor EMF- 49L or EMF-57 of 1.2E+05 cpm for > 15 minutes or will likely continue for 15 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure HP/0/B/1009/014.

(Continued) SITE AREA EMERGENCY 4.3.S.1 Boundary Dose Resulting from an Actual or Imminent Release of Radioactivity Exceeds 100 mrem TEDE or 500 mrem CDE Adult Thyroid for the Actual or Projected Duration of the Release. OPERATING MODE: ALL 4.3.S.1-1 A valid indication on radiation monitor EMF-36L of > 2.7E+06 cpm sustained for > 15 minutes.

4.3.S.1-2 Dose assessment team calculations indicate dose consequences greater than 100 mrem TEDE or 500 mrem CDE Adult Thyroid at the site boundary. (Continued)

GENERAL EMERGENCY 4.3.G.1 Boundary Dose Resulting from an Actual or Imminent Release of Radioactivity that Exceeds 1000 mrem TEDE or 5000 mrem CDE Adult Thyroid for the Actual or Projected Duration of the Release. OPERATING MODE: ALL 4.3.G.1-1 A valid indication on radiation monitor EMF-36H of > 8.3E+03 cpm sustained for > 15 minutes. 4.3.G.1-2 Dose assessment team calculations indicate dose consequences greater than 1000 mrem TEDE or 5000 mrem CDE Adult Thyroid at the site boundary. (Continued)

Enclosure 4.3 RP/0/A/5000/001 Abnormal Rad Levels/Radiological Effluent Page 2 of 5 UNUSUAL EVENT 4.3.U.1-3 Gaseous effluent being released exceeds two times SLC 16.11-6 for > 60 minutes as determined by RP procedure.

4.3.U.1-4 Liquid effluent being released exceeds two times SLC 16.11-1 for > 60 minutes as determined by RP procedure.

Note: If the monitor reading is sustained for the time period indicated in the EAL AND the required assessments (procedure calculations) cannot be completed within this time period, declaration must be made based on the valid radiation monitor reading. (Continued)

ALERT 4.3.A.1-2 A valid indication on radiation monitor EMF- 36L of 5.4E+05 cpm for > 15 minutes or will likely continue for 15 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure AD-EP-ALL-0202. 4.3.A.1-3 Gaseous effluent being released exceeds 200 times the level of SLC 16.11-6 for > 15 minutes as determined by RP procedure. 4.3.A.1-4 Liquid effluent being released exceeds 200 times the level of SLC 16.11-1 for > 15 minutes as determined by RP procedure.

Note: If the monitor reading is sustained for the time period indicated in the EAL AND the required assessments (procedure calculations) cannot be completed within this time period, declaration must be made based on the valid radiation monitor reading. (Continued) SITE AREA EMERGENCY 4.3.S.1-3 Analysis of field survey results or field survey samples indicates dose consequences greater than 100 mrem TEDE or 500 mrem CDE Adult Thyroid at the site boundary.

Note 1: These EMF readings are calculated based on average annual meteorology, site boundary dose rate, and design unit vent flow rate.

Calculations by the dose assessment team use actual meteorology, release duration, and unit vent flow rate. Therefore, these EMF readings should not be used if dose assessment team calculations are available. Note 2: If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor reading should be used for emergency classification. END GENERAL EMERGENCY 4.3.G.1-3 Analysis of field survey results or field survey samples indicates dose consequences greater than 1000 mrem TEDE or 5000 mrem CDE Adult Thyroid at the site boundary.

Note 1: These EMF readings are calculated based on average annual meteorology, site boundary dose rate, and design unit vent flow rate.

Calculations by the dose assessment team use actual meteorology, release duration, and unit vent flow rate. Therefore, these EMF readings should not be used if dose assessment team calculations are available. Note 2: If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor reading should be used for emergency classification. END Enclosure 4.3 RP/0/A/5000/001 Abnormal Rad Levels/Radiological Effluent Page 3 of 5 UNUSUAL EVENT 4.3.U.2 Unexpected Increase in Plant Radiation or Airborne Concentration.

OPERATING MODE: ALL 4.3.U.2-1 Indication of uncontrolled water level decrease of greater than 6 inches in the reactor refueling cavity with all irradiated fuel assemblies remaining covered by water.

4.3.U.2-2 Uncontrolled water level decrease of greater than 6 inches in the spent fuel pool and fuel transfer canal with all irradiated fuel assemblies remaining covered by water. 4.3.U.2-3 Unplanned valid area EMF reading increases by a factor of 1,000 over normal levels as shown in Enclosure 4.10. END ALERT 4.3.A.2 Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel. Does not apply to spent fuel in dry cask storage. Refer to EPLAN Section D basis document OPERATING MODE: ALL 4.3.A.2-1 An unplanned valid trip II alarm on any of the following radiation monitors: Spent Fuel Building Refueling Bridge 1EMF-15 2EMF-4 Spent Fuel Pool Ventilation 1EMF-42 2EMF-42 Reactor Building Refueling Bridge (applies to Mode 6 and No Mode Only) 1EMF-17 2EMF-2 Containment Noble Gas Monitor (Applies to Mode 6 and No Mode Only) 1EMF-39 2EMF-39 (Continued) SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.3 RP/0/A/5000/001 Abnormal Rad Levels/Radiological Effluent Page 4 of 5 UNUSUAL EVENT ALERT 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 <95% 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) SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.3 RP/0/A/5000/001 Abnormal Rad Levels/Radiological Effluent Page 5 of 5 UNUSUAL EVENT ALERT 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 1EMF-12 greater than 15 mrem/hr in the Control Room. 4.3.A.3-2 Valid indication of radiation levels greater than 15 mrem/hr in the Central Alarm Station (CAS) or Secondary Alarm Station (SAS).

4.3.A.3-3 Valid radiation monitor reading exceeds the levels shown in Enclosure 4.10. END SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.4 RP/0/A/5000/001 Loss of Shutdown Functions Page 1 of 3 UNUSUAL EVENT END ALERT 4.4.A.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was Successful. OPERATING MODE: 1, 2, 3 4.4.A.1-1 The following conditions exist: Valid reactor trip signal received or required and automatic reactor trip was not successful. AND Manual reactor trip from the control room is successful and reactor power is less than 5%

and decreasing. (Continued) SITE AREA EMERGENCY 4.4.S.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was NOT Successful. OPERATING MODE: 1 4.4.S.1-1 The following conditions exist: Valid reactor trip signal received or required and automatic reactor trip was not successful. AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing. (Continued) GENERAL EMERGENCY 4.4.G.1 Failure of the Reactor Protection System to Complete an Automatic Trip and Manual Trip Was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core. OPERATING MODE: 1 4.4.G.1-1 The following conditions exist: Valid reactor trip signal received or required and automatic reactor trip was not successful. AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing. AND EITHER of the following conditions exist:

Core Cooling CSF-RED

Heat Sink CSF-RED. END Enclosure 4.4 RP/0/A/5000/001 Loss of Shutdown Functions Page 2 of 3 UNUSUAL EVENT ALERT 4.4.A.2 Inability to Maintain Plant in Cold Shutdown.

OPERATING MODE: 5, 6 4.4.A.2-1 Total loss of ND and/or RN and/or KC. AND One of the following:

Inability to maintain reactor coolant temperature below 200ºF

Uncontrolled reactor coolant temperature rise to

>180ºF. END SITE AREA EMERGENCY 4.4.S.2 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown. OPERATING MODE: 1, 2, 3, 4 4.4.S.2-1 Subcriticality CSF-RED. 4.4.S.2-2 Heat Sink CSF-RED. 4.4.S.3 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel. OPERATING MODE: 5, 6 4.4.S.3-1 Failure of heat sink causes loss of cold shutdown conditions. AND Lower range Reactor Vessel Level Indication System (RVLIS) decreasing after initiation of NC system makeup. 4.4.S.3-2 Failure of heat sink causes loss of cold shutdown conditions. AND Reactor Coolant (NC) system mid or wide range level less than 11% and decreasing after initiation of NC system makeup. (Continued) GENERAL EMERGENCY Enclosure 4.4 RP/0/A/5000/001 Loss of Shutdown Functions Page 3 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY 4.4.S.3-3 Failure of heat sink causes loss of cold shutdown conditions. AND Either train ultrasonic level indication less than 7.25% and decreasing after initiation of NC system makeup. END GENERAL EMERGENCY Enclosure 4.5 RP/0/A/5000/001 Loss of Power Page 1 of 2 UNUSUAL EVENT 4.5.U.1 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes. OPERATING MODE: 1, 2, 3, 4 4.5.U.1-1 The following conditions exist: Loss of offsite power to essential buses ETA and ETB for greater than 15 minutes. AND Both emergency diesel generators are supplying power to their respective essential busses. (Continued) ALERT 4.5.A.1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode. OPERATING MODE: 5, 6, No Mode 4.5.A.1-1 Loss of all offsite and onsite AC power as indicated by: Loss of power on essential buses ETA and ETB. AND Failure to restore power to at least one essential bus within 15 minutes. (Continued) SITE AREA EMERGENCY 4.5.S.1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses. OPERATING MODE: 1, 2, 3, 4 4.5.S.1-1 Loss of all offsite and onsite AC power as indicated by: Loss of power on essential buses ETA and ETB. AND Failure to restore power to at least one essential bus within 15 minutes. (Continued) GENERAL EMERGENCY 4.5.G.1 Prolonged Loss of All (Offsite and Onsite) AC Power.

OPERATING MODE: 1, 2, 3, 4 4.5.G.1-1 Prolonged loss of all offsite and onsite AC power as indicated by: Loss of power on essential buses ETA and ETB for greater than 15 minutes. AND Standby Shutdown Facility (SSF) fails to supply NC pump seal injection OR CA supply to Steam Generators. AND (Continued)

Enclosure 4.5 RP/0/A/5000/001 Loss of Power Page 2 of 2 UNUSUAL EVENT OPERATING MODE: 5, 6, No Mode 4.5.U.1-2 The following conditions exist: Loss of offsite power to essential buses ETA and ETB for greater than 15 minutes. AND One emergency diesel generator is supplying power to its respective essential bus. 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 busses: EBA and EBD both <112 VDC, and EBB and EBC both <109 VDC. AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss. END ALERT 4.5.A.2 AC power to essential busses reduced to a single power source for greater than 15 minutes such that an additional single failure could result in station blackout. OPERATING MODE: 1, 2, 3, 4 4.5.A.2-1 The following condition exists: AC power capability has been degraded to one essential bus powered from a single power source for > 15 min. due to the loss of all but one of: SATA SATB ATC ATD D/G A D/G B END SITE AREA EMERGENCY 4.5.S.2 Loss of All Vital DC Power. OPERATING MODE: 1, 2, 3, 4 4.5.S.2-1 The following conditions exist: Unplanned loss of both unit related busses: EBA and EBD both <112 VDC, and EBB and EBC both <109 VDC. AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss. END GENERAL EMERGENCY At least one of the following conditions exist:

Restoration of at least one essential bus within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is NOT likely

Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring. END Enclosure 4.6 RP/0/A/5000/001 Fire/Explosion and Security Events Page 1 of 3 UNUSUAL EVENT 4.6.U.1 Fire Within Protected Area Boundary NOT Extinguished Within 15 Minutes of Detection OR Explosion Within the Protected Area Boundary. OPERATING MODE: ALL 4.6.U.1-1 Fire in any of the following areas NOT extinguished within 15 minutes of control room notification or verification of a control room fire alarm. Reactor Building Auxiliary Building Diesel Generator Rooms Control Room RN Pumphouse SSF CAS SAS Doghouses FWST Turbine Building Service Building Monitor Tank Building ISFSI Unit 1/2 Transformer Yard Areas (Continued) ALERT 4.6.A.1 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown. OPERATING MODE: 1, 2, 3, 4, 5, 6 4.6.A.1-1 The following conditions exist: (Non-security events) Fire or explosion in any of the following areas:

Reactor Building Auxiliary Building Diesel Generator Rooms Control Room RN Pumphouse SSF CAS SAS FWST Doghouses (Applies in Mode 1, 2, 3, 4 only). AND One of the following:

Affected safety system parameter indications show degraded performance (Continued)

SITE AREA EMERGENCY 4.6.S.1 HOSTILE ACTION within the PROTECTED AREA OPERATING MODE: ALL 4.6.S.1-1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the CNS Security Shift Supervision END GENERAL EMERGENCY 4.6.G.1 HOSTILE ACTION Resulting in Loss of Physical Control of the Facility. OPERATING MODE: ALL 4.6.G.1-1 A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions.

4.6.G.1-2 A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely for a FRESHLY OFF-LOADED REACTOR CORE in pool. END Enclosure 4.6 RP/0/A/5000/001 Fire/Explosion and Security Events Page 2 of 3 UNUSUAL EVENT 4.6.U.1-2 Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment or a loaded cask in the ISFSI. 4.6.U.2 Confirmed SECURITY CONDITION or Threat Which Indicates a Potential Degradation in the Level of Safety of the Plant.

OPERATING MODE: All 4.6.U.2-1 A SECURITY CONDITION that does NOT involve a HOSTILE ACTION as reported by the CNS Security Shift Supervision. 4.6.U.2-2 A credible site-specific security threat notification. 4.6.U.2-3 A validated notification from NRC providing information of an aircraft threat. END ALERT

Plant personnel report visible damage to permanent structures or equipment within the specified area required to establish or maintain safe shutdown within the specifications. Note: Only one train of a system needs to be affected or damaged in order to satisfy this condition. 4.6.A.2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown. OPERATING MODE: No Mode 4.6.A.2-1 The following conditions exist: (Non-security events) Fire or explosion in any of the following areas:

Spent Fuel Pool Auxiliary Building. RN Pump house AND One of the following:

Spent Fuel Pool level and/or temperature show degraded performance (Continued) SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.6 RP/0/A/5000/001 Fire/Explosion and Security Events Page 3 of 3 UNUSUAL EVENT ALERT

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 CNS Security Shift Supervision 4.6.A.3-2 A validated notification from NRC of airliner attack threat within 30 minutes of the site. END SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.7 RP/0/A/5000/001 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 1 of 4 UNUSUAL EVENT 4.7.U.1 Natural and Destructive Phenomena Affecting the Protected Area. OPERATING MODE: ALL 4.7.U.1-1 Tremor felt and valid alarm on the Syscom Seismic Monitoring System (OAC C1D2252). 4.7.U.1-2 Report by plant personnel of tornado striking within protected area boundary/ISFSI. 4.7.U.1-3 Vehicle crash into plant structures or systems within protected area boundary/ISFSI. 4.7.U.1-4 Report of turbine failure resulting in casing penetration or damage to turbine or generator seals. 4.7.U.1-5 Independent Spent Fuel Cask tipped over or dropped greater than 24 inches. 4.7.U.1-6 Uncontrolled flooding in the ISFSI area. 4.7.U.1-7 Tornado generated missiles(s) impacting the ISFSI. (Continued) ALERT 4.7.A.1 Natural and Destructive Phenomena Affecting the Plant Vital Area. OPERATING MODE: ALL 4.7.A.1-1 Valid "OBE Exceeded" Alarm on 1AD-4,B/8 4.7.A.1-2 Tornado or high winds: Tornado striking plant structures within the vital area:

Reactor Building

Auxiliary Building

FWST

Diesel Generator Rooms

Control Room

RN Pumphouse

SSF

Doghouses

CAS

SAS OR sustained winds 74 mph for > 15 minutes. (Continued)

SITE AREA EMERGENCY 4.7.S.1 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established. OPERATING MODE: ALL 4.7.S.1-1 The following conditions exist: Control Room evacuation has been initiated per AP/1(2)/A/5500/017 AND Control of the plant cannot be established from the ASP or the SSF within 15 minutes. 4.7.S.2 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of Site Area Emergency. OPERATING MODE: ALL 4.7.S.2-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate actual or likely major failures of plant functions needed for protection of the public. END GENERAL EMERGENCY 4.7.G.1 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of General Emergency. OPERATING MODE: ALL 4.7.G.1-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate: (1) actual or imminent substantial core degradation with potential for loss of containment OR (2) potential for uncontrolled radionuclide releases. These releases can reasonably be expected to exceed Environmental Protection Agency Protective Action Guideline levels outside the site boundary. END Enclosure 4.7 RP/0/A/5000/001 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 2 of 4 UNUSUAL EVENT 4.7.U.2 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant. OPERATING MODE: ALL 4.7.U.2-1 Report or detection of toxic or flammable gases that could enter within the site boundary in amounts that can affect safe operation of the plant. 4.7.U.2-2 Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event. 4.7.U.3 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of an Unusual Event. OPERATING MODE: ALL 4.7.U.3-1 Other conditions exist which in the judgement of the Emergency Coordinator/EOF Director indicate a potential degradation of the level of safety of the plant. END ALERT 4.7.A.1-3 Visible structural damage caused by either:

Vehicle crashes OR

Turbine failure generated missiles, OR

Other catastrophic events on any of the following plant structures:

Reactor Building

Auxiliary Building

FWST

Diesel Generator Rooms

Control Room

RN Pump House

SSF

Doghouses

CAS

SAS (Continued) SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.7 RP/0/A/5000/001 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 3 of 4 UNUSUAL EVENT ALERT 4.7.A.2 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown. OPERATING MODE: ALL 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. Structures for the above EALs:

Reactor Building

Auxiliary Building

Diesel Generator Rooms

Control Room

RN Pumphouse

SSF

CAS

SAS (Continued) SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.7 RP/0/A/5000/001 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 4 of 4 UNUSUAL EVENT ALERT 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. 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 SITE AREA EMERGENCY GENERAL EMERGENCY Enclosure 4.8 RP/0/A/5000/001 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) - Modes 1, 2, 3,4,5,6 and No Mode (Defueled)

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

CARF - Containment Air Return Fan.

CIVIL DISTURBANCE - A group of ten (10) or more people violently protesting station operations or activities at the site. A civil disturbance is considered to be violent when force has been used in an attempt to injure site personnel or damage plant property.

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.).

CREDIBLE THREAT - A threat should be considered credible when:

Physical evidence supporting the threat exists.

Information independent (law enforcement) from the actual threat message exists that supports the threat.

A specific group or organization claims responsibility for the threat.

EPA PAG - Environmental Protection Agency Protective Action Guidelines for exposure to a release of radioactive material.

EMERGENCY RELEASE - Any unplanned, quantifiable radiological release to the environment during an emergency event. The release does not have to be related to a declared emergency.

EXPLOSION - A rapid, violent unconfined combustion, or a catastrophic failure of pressurized equipment (e.g., a steamline or feedwater line break) that imparts energy sufficient to potentially damage or creates shrapnel to actually damage permanent structures, systems or components. An electrical breaker flash that creates shrapnel and results in damage to other components beyond scorching should also be considered.

EXTORTION - An attempt to cause an action at the site by threat of force.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flames is preferred but is NOT required if large quantities of smoke and heat are observed. An electrical breaker flash that creates high temperatures for a short duration and merely localized scorching to that breaker and its compartment should not be considered a fire.

Enclosure 4.8 RP/0/A/5000/001 Definitions/Acronyms Page 2 of 4 FRESHLY OFF-LOADED REACTOR CORE - The complete removal and relocation of all fuel assemblies from the reactor core and placed in the spent fuel pool. (Typical of a "No Mode" operation during a refuel outage that allows safety system maintenance to occur and results in maximum decay heat load in the spent fuel pool system.)

FUNCTIONAL - A component is fully capable of meeting its design function. It would be declared INOPERABLE if unable to meet Technical Specifications.

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 guideline exposure levels offsite for more than the immediate site area.

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

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

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. INOPERABLE - A component does not meet Technical Specifications. The component may be functional, capable of meeting its design. 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 - Includes the components approved for loading and storage of spent fuel assemblies.

LOSS - A component is INOPERABLE and not FUNCTIONAL.

NO MODE - Defueled.

OWNER CONTROLLED AREA - Area outside the protected area fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.

Enclosure 4.8 RP/0/A/5000/001 Definitions/Acronyms Page 3 of 4 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 misoperation of plant equipment with the intent to render the equipment unavailable. Equipment found tampered with or damaged due to malicious mischief may not meet the definition of SABOTAGE until this determination is made by security supervision.

SECURITY CONDITION - Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

SIGNIFICANT PLANT 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 Guidelines to exposure levels beyond the site boundary. SITE BOUNDARY - That area, including the protected area, in which Duke Energy 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). TERMINATION - Exiting the emergency condition.

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

Enclosure 4.8 RP/0/A/5000/001 Definitions/Acronyms Page 4 of 4 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.

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 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/5000/001 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 with 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 Catawba Nuclear Site Emergency Plan) is available for review if any questions arise over proper classification.

Emergencies are declared for the site. If an event results in multiple emergency action levels on a unit or different emergency action levels on each unit, then the emergency declaration shall be based on the higher classification. Information relating to the unit with the lesser classification will be noted as additional information on the Emergency Notification Form (ENF).

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 (Mode 1-4) hot shutdown or higher. An event that is recognized at cold shutdown or lower (Mode 5 or 6) shall not be classified using the fission product barrier matrix. Reference would be made to the other 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 that a condition exist for a specific duration prior to declaration.

a. 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. b. 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.

Enclosure 4.9 RP/0/A/5000/001 Emergency Declaration Guidelines Page 2 of 2 2. 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. 3. If an emergency classification is warranted, but the plant condition is corrected prior to declaration and notification, the Emergency Coordinator must consider the potential that the initiating condition (e.g. Failure of Reactor Protection System or earthquake) may have caused plant damage that warrants augmenting the on-shift personnel via activation of the Emergency Response Organization. The following action shall be taken:

a. For UNUSUAL EVENTS, the condition shall be declared and notifications made. The event may be terminated in the same notification or in a follow-up notification.

b. For ALERT, SITE AREA EMERGENCY, and GENERAL EMERGENCY, the event shall be declared and the emergency response organization activated.

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 minutes 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/5000/001 Radiation Monitor Readings for Enclosure 4.3 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 Event mR/hr Alert mR/hr 1EMF-1 522' FF, 57 Auxiliary Building Corridor 500 5000 1EMF-3 543' GG, 55 Unit 1 Charging Pump Area 100 5000 1EMF-4 543' GG, 59 Unit 2 Charging Pump Area 100 5000 1EMF-7 560' NN, 55 Unit 1 Auxiliary Building Corridor 1500 5000 1EMF-8 560' NN, 59 Unit 2 Auxiliary Building Corridor 500 5000 1EMF-9 577' LL, 55 Unit 1 Aux. Building Filter Hatch 100 5000 1EMF-10 577' LL, 58 Unit 2 Aux. Building Filter Hatch 100 5000 1EMF-22 594' KK, 53 Containment Purge Filter Area 100 5000 2EMF-9 594' KK, 61 Containment Purge Filter Area 100 5000

CNS PAGE NO. EP/1/A/5000/FR-P.1 18 of 48 Revision 25

17. (Continued)

c. Verify RVLIS indication adequate as c. RETURNTO Step 14. follows: IF all NC pumps are off, THEN verify "REACTOR VESSEL LR LEVEL" -

GREATER THAN 61%. IF any NC pump is on, THEN verify "REACTOR VESSEL D/P" -

GREATER THAN REQUIRED D/P FROM TABLE BELOW:

d. Dispatch operator to restore power to all CLA discharge isolation valves. REFERTO EP/1/A/5000/G-1 (Generic Enclosures), Enclosure 9 (Power Alignment for CLA Valves).

e. Maintain NC pressure greater than CLA pressure until the CLAs are isolated or vented. SHOCK CONDITIONRESPONSE TO IMMINENT PRESSURIZED THERMAL

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