Text

1A- SRO Exam as Administered
	ML23079A092
Person / Time
Site:	Catawba
Issue date:	03/20/2023
From:	; Division of Reactor Safety II
To:	;
References
	Download: ML23079A092 (1)
	v • d • e

Question:

(1 point) 1 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

The Unit 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 (Minimum S/I Flowrate Versus Time After Trip)

Current conditions:

Unit 1 Reactor was tripped 60 minutes ago 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.

Which ONE of the following correctly completes the statements above?

REFERENCE PROVIDED A.

1.

2.

340 gpm 1A NV B.

1.

2.

360 gpm 1A NV C.

1.

2.

340 gpm 1B NI D.

1.

2.

360 gpm 1B NI Page 1 of 100 ML23079A092

Question:

(1 point) 2 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A reactor trip and safety injection have occurred due to a stuck open Pressurizer Safety valve EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) has been entered NC pressure is 1100 psig and lowering slowly Subcooling based on CETs is -20°F and lowering In accordance with E-0, to trip the NC pumps on a loss of subcooling requires at least one _____(1)_____ pump delivering S/I flow to the NC system.

The basis for tripping the NC pumps due to loss of subcooling is to _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. NV and NI

2. minimize mass loss from the NC system B.

1. NV and NI

2. minimize NCS heatup from energy added by the NC pumps C.

1. NV or NI

2. minimize mass loss from the NC system D.

1. NV or NI

2. minimize NCS heatup from energy added by the NC pumps Page 2 of 100

Question:

(1 point) 3 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A LOCA has occurred Containment pressure peaked at 3.2 PSIG and is now 2.4 PSIG and lowering The crew is determining whether S/I termination criteria are met per EP/1/A/5000/E-1 (Loss of Reactor or Secondary Coolant)

NC pressure is 1300 PSIG and stable All S/G N/R levels are 27% and stable Total CA flow is 350 GPM and stable Pressurizer level is 26% and stable Heat sink requirement for S/I termination _____(1)_____ met.

Pressurizer level requirement for S/I termination _____(2)_____ met.

Which ONE of the following completes the statements above?

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 3 of 100

Question:

(1 point) 4 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

Reactor trip and S/I have occurred (simultaneously) due to a Large Break LOCA from 100% RTP A CF Isolation signal was FIRST generated based on _____(1)_____.

The purpose of the CF Isolation signal _____(2)_____ to prevent uncontrolled cooldown of the NC system.

Which ONE of the following completes the statements above?

A.

1. NC system Tavg

2. is B.

1. NC system Tavg

2. is NOT C.

1. the Ss signal

2. is D.

1. the Ss signal

2. is NOT Page 4 of 100

Question:

(1 point) 5 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

Unit 1 is at 95% RTP due to electrical grid issues An underfrequency condition is sensed on 2/4 NCP Busses All Unit 1 NCPs have tripped Which ONE of the following completes the statement below?

In accordance with EP/1/A/5000/ES-0.1 (Reactor Trip Response), S/G NR levels greater than a MINIMUM of _____(1)_____ ensures that _____(2)_____.

A.

1. 11%

2. the S/Gs are maintained as a heat sink B.

1. 11%

2. the S/G tube bundle is covered to prevent S/G depressurization C.

1. 16%

2. the S/Gs are maintained as a heat sink D.

1. 16%

2. the S/G tube bundle is covered to prevent S/G depressurization Page 5 of 100

Question:

(1 point) 6 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 initial conditions:

The Unit is at 100% RTP The crew has entered AP/1/A/5500/012 (Loss of Charging or Letdown), Case I (Loss of Charging) following a trip of 1A NV Pump Subsequently:

The crew is in the process of re-establishing charging o 1B NV Pump has been started o 1NV-294 (NV Pmps A&B Disch Flow Ctrl) has been throttled to proper flow In accordance with AP/12, the BOP will next throttle 1NV-309 (Seal Water Injection Flow) in the _____(1)_____ direction to establish a MINIMUM of _____(2)_____

Total Seal Water Flow.

Which ONE of the following completes the statement above?

A.

1. OPEN

2. 40 gpm B.

1. CLOSED

2. 40 gpm C.

1. OPEN

2. 32 gpm D.

1. CLOSED

2. 32 gpm Page 6 of 100

Question:

(1 point) 7 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 2 initial conditions:

The Unit is in MODE 5 and drained to Mid Loop ND Train 2A is in service ND system flow rate is 3200 gpm NC system level is 6.5% and stable Subsequently:

2A ND pump amps and discharge pressure begin to oscillate The crew has entered AP/2/A/5500/019 (Loss of Residual Heat Removal System), Case IV (Loss of ND in Mid Loop or S/G Manway Removed)

Per AP/19 Case IV, the BOP will FIRST __________ to mitigate this issue.

Which ONE of the following completes the statements above?

A.

secure 2A ND pump B.

reduce ND flow to less than a MAXIMUM of 2000 gpm C.

reduce ND flow to less than a MAXIMUM of 1500 gpm D.

reduce ND flow to less than a MAXIMUM of 1000 gpm Page 7 of 100

Question:

(1 point) 8 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 timeline:

0800 Unit is at 100% RTP 1A2 KC Pump is in service 0830 1EMF-46B (Component Cooling Water) power supply fails 0900 1A2 KC Pump trips Crew enters AP/1/A/5500/021 (Loss of Component Cooling Water) 1B1 KC Pump is started Following trip of 1A2 KC Pump, 1RAD-1 D/5 1EMF-46A TRN A KC Loss Of Flow

_____(1)_____ alarm.

Entry into the action statement of SLC 16.7-10 (Radiation Monitoring for Plant Operations) was required at _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. did

2. 0830 B.

1. did

2. 0900 C.

1. did NOT

2. 0830 D.

1. did NOT

2. 0900 Page 8 of 100

Question:

(1 point) 9 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is at 100% RTP 1AD-6, D/8 (PZR LO PRESS ALERT) has alarmed Pressurizer Pressure Ch. 2 is reading 1700 PSIG and stable Pressurizer Pressure Ch. 1, 3, and 4 are reading 2240 PSIG and stable Tech Specs require the operators to verify that P-11 (PZR S/I BLOCK PERMISSIVE) is in the proper state in a MAXIMUM timeframe of _____(1)_____.

The crew _____(2)_____ required to enter AP/1/A/5500/011 (Pressurizer Pressure Anomalies).

Which ONE of the following completes the statements above?

A.

1. immediately

2. is B.

1. one hour

2. is C.

1. immediately

2. is NOT D.

1. one hour

2. is NOT Page 9 of 100

Question:

(1 point) 10 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

Unit 1 has experienced a S/G Tube Rupture The crew has transitioned to EP/1/A/5000/E-3 (Steam Generator Tube Rupture) and is preparing to initiate a cooldown to target CET temperature Concerning the cooldown, this procedure will specify _____(1)_____.

NC Pump trip criteria, based on NC subcooling _____(2)_____ apply after starting a controlled cooldown.

Which ONE of the following completes the statements above?

A.

1. as close as possible without exceeding 100° F per hour

2. does B.

1. as close as possible without exceeding 100° F per hour

2. does NOT C.

1. maximum rate while attempting to avoid a Main Steam Isolation

2. does D.

1. maximum rate while attempting to avoid a Main Steam Isolation

2. does NOT Page 10 of 100

Question:

(1 point) 11 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following on Unit 1:

Unit 1 has established feed and bleed while performing the actions of EP/1/A/5000/FR-H.1 (Response To Loss Of Secondary Heat Sink)

Subsequently:

The TDCA pump has been started and is available to feed the S/Gs The crew is transitioning to FR-H.1 Enclosure 7 (S/G CA Flow Restoration)

CETs are STABLE All S/G WR levels are indicating 0%

Per FR-H.1 Enclosure 7:

The criteria for restoration of CA flow is to restore cooling to _____(1)_____ at a rate not to exceed 100 GPM.

The primary basis for the restoration of flow criteria _____(2)_____ to minimize thermal stress of the S/G components.

Which ONE of the following completes the statements above?

A.

1. ONE S/G

2. is B.

1. ONE S/G

2. is NOT C.

1. ALL S/Gs

2. is D.

1. ALL S/Gs

2. is NOT Page 11 of 100

Question:

(1 point) 12 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A Loss of Offsite Power (LOOP) has occurred Both D/Gs failed to automatically start EP/1/A/5000/ECA-0.0 (Loss of All AC Power) has been entered If manual start of the D/Gs is unsuccessful, the next attempt at power restoration will be _____(1)_____.

If attempts to restore power are unsuccessful, long term shutdown margin is guaranteed by use of _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. local start of ESPS D/Gs

2. boric acid transfer pumps B.

1. initiation of Safety Injection

2. boric acid transfer pumps C.

1. local start of ESPS D/Gs

2. the standby makeup pump D.

1. initiation of Safety Injection

2. the standby makeup pump Page 12 of 100

Question:

(1 point) 13 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A reactor trip has occurred due to a Loss of Offsite Power Crew has entered EP/1/A/5000/E-0 (Reactor Trip or Safety Injection)

BOP is performing E-0 step 4 Verify 1ETA and 1ETB - ENERGIZED To verify that 1ETA and 1ETB are energized; LINE VOLTS meters on 1MC-8 _____(1)_____ be an accurate indication.

KC valve E30 switch position status lights _____(2)_____ be an accurate indication.

Which ONE of the following completes the statements above?

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 13 of 100

Question:

(1 point) 14 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

A blackout on 2ETA has occurred Crew entered AP/2/A/5500/007 (Loss of Normal Power) Case II (Loss of All Power to an Essential Train) when 2A D/G failed to automatically start Attempts to start 2A D/G locally have failed Crew has entered AP/2/A/5500/029 (Loss of Vital or Aux Control Power)

In accordance with AP/29:

Vital and Aux Control Power batteries are capable of carrying their associated loads for a MINIMUM of _____(1)_____ hours.

If a complete loss of voltage occurs to 2EPA, the Unit 2 MSIVs will fail _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. 3

2. as is B.

1. 3

2. closed C.

1. 2

2. as is D.

1. 2

2. closed Page 14 of 100

Question:

(1 point) 15 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

Both Units are operating at 100% RTP 1A RN pump is in service Subsequently:

1A RN pump trips Crew enters AP/0/A/5500/020 (Loss of Nuclear Service Water)

Based on the above information and in accordance with AD-OP-ALL-1000 (Conduct of Operations):

The tripping of 1A RN pump _____(1)_____ an appropriate example of when a Crew Update should be performed.

A plant wide announcement _____(2)_____ required prior to starting a standby RN pump per AP/20.

Which ONE of the following completes the statements above?

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 15 of 100

Question:

(1 point) 16 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

AP/1/A/5500/037 (Generator Voltage and Electric Grid Disturbances), Case I (Abnormal Generator or Grid Voltage) has been entered following a Grid Disturbance The TCC has reported that Real Time Contingency Analysis (RTCA) indicates CNS switchyard voltage would NOT be adequate if the unit should trip SPOC is making preparation for installation of Jumpers per AM/1/A/5100/008 (4KV Essential Power System Degraded Voltage Logic)

Unit 1 ECCS is currently _____(1)_____.

Once jumpers are installed, LOCA sequencer actuation timing for the 4160V incoming breakers _____(2)_____ affected.

Which ONE of the following completes the statements above?

A. 1.

2.

operable is B. 1.

2.

operable is NOT C. 1.

2.

inoperable is D. 1.

2 inoperable is NOT Page 16 of 100

Question:

(1 point) 17 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A LOCA outside containment has occurred EP/1/A/5000/ECA-1.2 (LOCA Outside Containment) has been entered Crew is performing step 2 to attempt to identify and isolate the leak Subcooling is -5°F and stable In accordance with ECA-1.2:

The crew will FIRST attempt to isolate the leak by isolating the _____(1)_____ system from the NC system.

Rising NC pressure _____(2)_____ be the only parameter used to verify leak isolation.

Which ONE of the following completes the statements above?

A.

1. ND

2. will B.

1. NI

2. will C.

1. ND

2. will NOT D.

1. NI

2. will NOT Page 17 of 100

Question:

(1 point) 18 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Crew is performing steps in EP/1/A/5000/FR-H.1 (Loss of Secondary Heat Sink)

CA System Valve Control is RESET All CA pumps are secured Annunciator 1AD-8, B/1 UST LO LEVEL is LIT UST level is 8% and stable The CRS has entered AP/1/A/5500/006 (Loss of S/G Feedwater)

Case II (Loss of Normal CA Supply) to align CA suction to the hotwell Main Condenser Vacuum has been broken In accordance with AP/06 Case II, when aligning CA suction to the hotwell:

The CA Pump Low Suction Pressure Trips _____(1)_____ be blocked.

Preference should be given to first starting the _____(2)_____ CA pump(s).

Which ONE of the following completes the statements above?

A.

1. will

2. Turbine Driven B.

1. will

2. Motor Driven C.

1. will NOT

2. Turbine Driven D.

1. will NOT

2. Motor Driven Page 18 of 100

Question:

(1 point) 19 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following initial conditions on Unit 1:

Unit is stable at 50% RTP Control Bank D is at 180 steps withdrawn Subsequently:

Control rods begin stepping out at 8 steps per minute After rods stepped out 4 steps, the RO places the CRD Bank Select switch in MANUAL Crew has entered AP/1/A/5500/015 (Rod Control Malfunction)

Case II (Continuous Rod Movement)

Following the control rod withdrawal, the RO can expect Pressurizer Level to be

_____(1)_____ the value prior to the failure.

Once the failure has been repaired, AP/15 will have the crew place control rods back in AUTO once Tavg is within a MAXIMUM of _____(2)_____ of Tref.

Which ONE of the following completes the statements above?

A.

1. higher than

2. 1°F B.

1. higher than

2. 1.5°F C.

1. equal to

2. 1°F D.

1. equal to

2. 1.5°F Page 19 of 100

Question:

(1 point) 20 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following WR NIS Power trends:

TREND A TREND B Based on the trends above, _____(1)_____ indicates that a dropped rod has occurred.

For a dropped rod, the background color on the DRPI display for the affected rod group will be _____(2)_____.

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

A.

1. Trend A

2. orange B.

1. Trend A

2. black C.

1. Trend B

2. orange D.

1. Trend B

2. black Page 20 of 100

Question:

(1 point) 21 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A LOCA has occurred The VE system is in operation HVAC Annunciator 1AD-19, B/1 (VE CARBON FILTER TRAIN A HI-HI TEMP) is received Based on the conditions above, 1A VE Fan _____(1)_____ automatically trip as temperature rises.

An operator will be dispatched to _____(2)_____.

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

A.

1. will NOT

2. manually OPEN the Filter Sprinkler RF isolation valves B.

1. will NOT

2. verify automatic actuation of the Filter Sprinkler system C.

1. will

2. manually OPEN the Filter Sprinkler RF isolation valves D.

1. will

2. verify automatic actuation of the Filter Sprinkler system Page 21 of 100

Question:

(1 point) 22 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A LOCA has occurred The crew entered EP/1/A/5000/FR-C.1 (Response to Inadequate Core Cooling)

S/G depressurization was not effective in restoring adequate core cooling Core Exit Thermocouples (CET) are 1200°F and rising In accordance with FR-C.1:

The crew will _____(1)_____ and monitor until CETs lower below 1200°F.

Normal conditions for starting a NC pump _____(2)_____ required.

Which ONE of the following completes the statements above?

A.

1. start NC pumps one at a time

2. are B.

1. start NC pumps one at a time

2. are NOT C.

1. start all available NC pumps

2. are D.

1. start all available NC pumps

2. are NOT Page 22 of 100

Question:

(1 point) 23 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is in Mode 5 Loops Not Filled for a refueling outage Equipment Hatch is off to stage equipment for fuel offload Subsequently:

Pressurizer Level begins to lower Containment Floor and Equipment Sump (CFES) level is rising Entry conditions for AP/1/A/5500/010 (Reactor Coolant Leak) Case II (NC System Leak) _____(1)_____ met.

The Equipment Hatch _____(2)_____ required to be closed.

Which ONE of the following completes the statements above?

A.

1. are

2. is B.

1. are

2. is NOT C.

1. are NOT

2. is D.

1. are NOT

2. is NOT Page 23 of 100

Question:

(1 point) 24 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

EP/1/A/5000/FR-H.1 (Response to Loss of Secondary Heat Sink) has been implemented Following Bleed and Feed initiation, CA flow has been restored from CAPT #1 Containment pressure peaked at 3.2 psig and is now 2.1 psig In accordance with FR-H.1, which ONE of the following indicates the MINIMUM heat sink requirements that must be met to allow termination of NC system bleed and feed?

A.

NR level in at least ONE S/G > 11%

B.

NR level in at least ONE S/G > 29%

C.

WR level in at least ONE S/G > 24%

D.

WR level in at least ONE S/G > 36%

Page 24 of 100

Question:

(1 point) 25 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A LOCA has occurred NC pressure is 1300 PSIG and stable Crew has entered EP/1/A/5000/ES-1.2 (Post LOCA Cooldown and Depressurization)

The ECCS Steam Pressure Block pushbuttons have been depressed Current Steam pressure is 800 PSIG and lowering at 0.5 PSIG per second The low steam line pressure Main Steam Isolation signal _____(1)_____ active at this time.

If the current rate of lowering Steam Pressure is maintained over the next 5 minutes, a Main Steam Isolation _____(2)_____ occur.

Which ONE of the following completes the statements above?

A.

1. is

2. will B.

1. is NOT

2. will C.

1. is

2. will NOT D.

1. is NOT

2. will NOT Page 25 of 100

Question:

(1 point) 26 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit was operating at 100% RTP when a Loss of Offsite Power (LOOP) occurred 1A D/G failed to start Crew is performing actions of EP/1/A/5000/ES-0.2 (Natural Circulation Cooldown) 1AD-5, F/3 CAPT MECH OS TRIP is LIT In accordance with ES-0.2:

1A and 1B NC System Loops are considered _____(1)_____.

Cooldown of active loops will be performed using the _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. active

2. steam dumps B.

1. active

2. S/G PORVs C.

1. inactive

2. steam dumps D.

1. inactive

2. S/G PORVs Page 26 of 100

Question:

(1 point) 27 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

Unit is in Mode 3 All NC pumps are in service 2A NCP #1 Seal Leakoff flow is 4.5 GPM and rising slowly 2A NCP lower radial bearing temperature is 200°F and rising at 5°F/minute Entry conditions for AP/2/A/5500/008 (Reactor Coolant Pump Malfunction)

_____(1)_____ met.

2A NC pump lower radial bearing will reach a trip setpoint in a MINIMUM of

_____(2)_____ minutes.

Which ONE of the following completes the statements above?

A.

1. are NOT

2. 5 B.

1. are NOT

2. 7 C.

1. are

2. 5 D.

1. are

2. 7 Page 27 of 100

Question:

(1 point) 28 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is at 100% RTP S/G pressure and level begin to lower slowly NC system Tavg is lowering Containment pressure is rising slowly Pressurizer level is lowering slowly The FIRST procedurally directed action to stabilize Pressurizer level will have the operators _____(1)_____.

The accident in progress is a _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. lower letdown flow

2. steam leak B.

1. lower letdown flow

2. NC system leak C.

1. raise charging flow

2. steam leak D.

1. raise charging flow

2. NC system leak Page 28 of 100

Question:

(1 point) 29 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

Unit is in Mode 4 performing a cooldown for refueling Train 2A ND is in service ND flow is stable at 3400 gpm Which ONE (1) 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?

A.

1. Rate RISES

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

B.

1. Rate LOWERS

2. Place PWR DISCON FOR 2NI-173A switch to THROT and manually control 2NI-173A C.

1. Rate RISES

2. Place PWR DISCON FOR 2NI-173A switch to THROT and manually control 2NI-173A D.

1. Rate LOWERS

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

Page 29 of 100

Question:

(1 point) 30 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A Steamline break has occurred on the equalization header NC system pressure is 1200 PSIG and lowering Main Steam equalization header pressure is 500 psig and lowering slowly MSIVs on 1B & 1C S/Gs did NOT close on the Main Steam Isolation signal 1A and 1D S/G N/R levels are 30% and rising Per EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) Enclosure 1 (Foldout Page):

1NV-202B & 1NV-203A (NV Pumps A&B Recirc Isol) were required to be closed by the operators when NC system pressure lowered to less than a MAXIMUM of

_____(1)_____ PSIG.

Auxiliary Feedwater flow _____(2)_____ required to be isolated to the 1B & 1C S/Gs.

Which ONE of the following completes the statements above?

A.

1. 1500

2. is B.

1. 1500

2. is NOT C.

1. 2000

2. is D.

1. 2000

2. is NOT Page 30 of 100

Question:

(1 point) 31 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 timeline:

1000 A load rejection resulted in a reactor trip from 100% RTP Following the trip, a Pressurizer Safety valve opens, and will NOT reseat The PRT rupture disks function as designed Containment pressure is 0.1 psig and rising at 0.03 psig every 5 minutes Lower Containment temperature is 110°F and rising at 2°F every 5 minutes Assuming these conditions remain constant, and concerning only the application of LCOs 3.6.4 and 3.6.5 Plant conditions will FIRST require entry into LCO _____(1)_____.

At 1030, conditions for entry into _____(2)_____ will be met.

Which ONE of the following correctly completes the statements above?

LEGEND:

LCO 3.6.4 (Containment Pressure)

LCO 3.6.5 (Containment Air Temperature)

A.

1.

3.6.4

2.

3.6.4 ONLY B.

1.

3.6.4

2.

3.6.4 AND 3.6.5 C.

1.

3.6.5

2.

3.6.5 ONLY D.

1.

3.6.5

2.

3.6.4 AND 3.6.5 Page 31 of 100

Question:

(1 point) 32 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

A Safety Injection has occurred The BOP notices PRT level, pressure, and temperature rising slowly NO ESFAS related signals have been reset Per OP/2/A/6150/004 (Pressurizer Relief Tank), the temperature in the PRT shall not exceed a MAXIMUM of _____(1)_____ above lower containment ambient temperature.

Based on the conditions above, cooling of the PRT _____(2)_____ available.

Which ONE of the following correctly completes the statements above?

A.

1.

8°F

2.

is B.

1.

8°F

2.

is NOT C.

1.

10°F

2.

is D.

1.

10°F

2.

is NOT Page 32 of 100

Question:

(1 point) 33 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

Unit is operating at 100% RTP at BOL NV system letdown is through the mixed bed demineralizers A loss of instrument air (VI) to valve 2KC-132 (Letdn Hx Otlt Temp Ctrl) has occurred The above failure will result in overall KC system flow _____(1)_____.

If the above failure is not corrected, entry conditions for AP/2/A/5500/013 (Boron Dilution) _____(2)_____ be met.

Which ONE of the following completes the statements above?

A.

1. rising

2. will B.

1. rising

2. will NOT C.

1. lowering

2. will D.

1. lowering

2. will NOT Page 33 of 100

Question:

(1 point) 34 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is operating at 100% RTP A failure has caused valve 1KC-132 to close 1AD-7, F/3 LETDN HX OUTLET HI TEMP is LIT The BOP has placed 1KC-132 in MANUAL Current letdown temperature is 129°F and rising slowly The BOP will operate button _____(1)_____ to restore normal letdown temperature.

Per the 1AD-7, F/3 ARP, if letdown temperature can be restored to normal, entry in AP/1/A/5500/021 (Loss of Component Cooling Water) _____(2)_____ required.

Which ONE of the following completes the statements above?

A.

1. 1

2. is NOT B.

1. 2

2. is NOT C.

1. 1

2. is D.

1. 2

2. is Page 34 of 100

Question:

(1 point) 35 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 2 conditions:

An input issue is causing OAC alarm C2P1361 (PZR STM TEMP-PZR SPRAY B TEMP D/T) to intermittently come in HIGH RO gets permission from the CRS to delete this alarm while waiting for W/O planning by Maintenance In accordance with AD-OP-ALL-1000 (Conduct of Operations):

The deleted OAC alarm _____(1)_____ required to be logged in eSOMS.

An audit of deleted computer alarms is required ____(2)_____.

Which ONE of the following correctly completes the statements above?

A.

1. is

2. weekly B.

1. is

2. monthly C.

1. is NOT

2. weekly D.

1. is NOT

2. monthly Page 35 of 100

Question:

(1 point) 36 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Which ONE (1) of the following describes the impact on the Reactor Protection System for a loss of 125 VDC Panelboard 1EPA or 1EPD?

A.

The associated Reactor Trip Breaker cannot be opened by the Shunt trip B.

The associated Reactor Trip Breaker cannot be opened by the UV trip C.

SSPS Output Bay has lost one of two power supplies D.

SSPS Logic Bay has lost one of two power supplies Page 36 of 100

Question:

(1 point) 37 ILT23 CNS SRO NRC Examination Catawba Nuclear Station The purpose of the Aux Safeguards Test Cabinet is to test the _____(1)_____ relays in the Solid State Protection System.

_____(2)_____ testing is where the final actuation device is actuated and its operation is verified by observation.

Which ONE of the following completes the statements above?

A.

1. Master

2. GO B.

1. Master

2. BLOCK C.

1. Slave

2. GO D.

1. Slave

2. BLOCK Page 37 of 100

Question:

(1 point) 38 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

Unit 2 was at 100% RTP A complete loss of offsite power (LOOP) occurs Which ONE of the following lists ALL Containment Ventilation system components that will be operating after D/G load sequence is complete?

1. Pipe Tunnel Booster fans

2. Lower Containment Ventilation Units

3. Upper Containment Ventilation Units

4. Containment Auxiliary Charcoal Filter Unit A.

1, 2, and 3 B.

2, 3, and 4 C.

1 and 2 ONLY D.

3 and 4 ONLY Page 38 of 100

Question:

(1 point) 39 ILT23 CNS SRO NRC Examination Catawba Nuclear Station When manually starting a Containment Air Return Fan via the control room switch:

A CPCS signal _____(1)_____ required.

The time delay _____(2)_____ be bypassed.

Which ONE of the following completes the statements above?

A.

1. is

2. will B.

1. is

2. will NOT C.

1. is NOT

2. will D.

1. is NOT

2. will NOT Page 39 of 100

Question:

(1 point) 40 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is at 100% RTP It has been determined that four Ice Condenser Intermediate Deck doors will not open due to excessive ice buildup Entry into the Required Actions of Tech Spec 3.6.13 (Ice Condenser Doors)

_____(1)_____ required.

During an accident, the ice condenser is designed to maintain a MAXIMUM Containment design pressure of _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. is

2. 20 psig B.

1. is NOT

2. 20 psig C.

1. is

2. 15 psig D.

1. is NOT

2. 15 psig Page 40 of 100

Question:

(1 point) 41 ILT23 CNS SRO NRC Examination Catawba 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 41 of 100

Question:

(1 point) 42 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

A LOCA has occurred inside Containment All Containment Spray (NS) pump discharge valves are OPEN ECCS has been RESET In order to start the 1A Containment Spray (NS) pump, the BOP must verify that CPCS pressure is greater than a MINIMUM of _____(1)_____ and must depress the _____(2)_____.

Which ONE of the following correctly completes the statement above?

A.

1.

2.

0.90 psig 1A NS pump ON pushbutton ONLY B.

1.

2.

0.35 psig 1A NS pump ON pushbutton ONLY C.

1.

2.

0.90 psig 1A DG Sequencer RESET pushbutton AND 1A NS pump ON pushbutton D.

1.

2.

0.35 psig 1A DG Sequencer RESET pushbutton AND 1A NS pump ON pushbutton Page 42 of 100

Question:

(1 point) 43 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following on Unit 2:

Unit 2 is at 25% RTP A tube leak has developed on the 2B S/G A Trip 2 on _____(1)_____ will cause the S/G BB flow control valves to close.

2EMF-72 (Steam Generator B Leakage) _____(2)_____ accurate as power level rises to 75%.

Which ONE of the following completes the statements above?

A.

1. 2EMF-33 (Condenser Air Ejector Exhaust)

2. remains as B.

1. 2EMF-33 (Condenser Air Ejector Exhaust)

2. is more C.

1. 2EMF-11 (Steam Line 2B)

2. remains as D.

1. 2EMF-11 (Steam Line 2B)

2. is more Page 43 of 100

Question:

(1 point) 44 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

Unit was initially 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.

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

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) 45 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following initial conditions on Unit 1:

The Condensate System is aligned for High Pressure Cleanup Subsequently:

The 1A CFPT experiences a complete loss of oil pressure The windmill protection circuit will send a signal to DIRECTLY trip the operating

_____(1)_____.

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

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

A.

1. Hotwell Pumps

2. discharge valve B.

1. Hotwell Pumps

2. recirc valve C.

1. Condensate Booster Pumps

2. discharge valve D.

1. Condensate Booster Pumps

2. recirc valve Page 45 of 100

Question:

(1 point) 46 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

The unit is in Mode 4 The crew is performing unit heatup and pressurization The CA PUMP AUTO START DEFEAT lights are LIT Based on the conditions above:

If both Main Feedwater pumps trip, the 1A and 1B CA pumps _____(1)_____ auto start.

If a Safety Injection occurs, the 1A and 1B CA pumps _____(2)_____ auto start.

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

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 46 of 100

Question:

(1 point) 47 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

Unit 1 is at 100% RTP The following Annunciators actuate:

o 1AD-11 A/8 125VDC Diesel Gen A Control Pwr Sys Trbl o 1AD-11 B/7 D/G A Panel Trouble The dispatched operator reports that the 1A D/G Control Power supply breaker 1DGCA-B304 has tripped open and appears to be damaged Which ONE of the following completes the statement below?

Upon a loss of all offsite power, the 1A D/G _____(1)_____ automatically start because _____(2)_____.

A.

1. will NOT

2. the D/G has lost control power B.

1. will NOT

2. the sequencer has lost control power C.

1. will

2. control power is available from 1EDA D.

1. will

2. the D/G auto starts upon a loss of control power Page 47 of 100

Question:

(1 point) 48 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

Unit 1 is at 100% RTP Vital DC battery chargers 1ECA, 1ECB, 1ECC, and 1ECD are in service and in FLOAT Spare Vital DC battery charger 1ECS is energized from 1EMXJ Subsequently:

An Overvoltage condition develops in the output of charger 1ECA The Overvoltage condition on the output of charger 1ECA will cause the

_____(1)_____ breaker to open.

In accordance with OP/1/A/6350/008 (125VDC/120 V AC Vital Instrument and Control Power System), battery charger 1ECS _____(2)_____ be aligned to supply Vital DC bus 1EDA in its current configuration.

Which ONE of the following completes the statements above?

A.

1. AC input

2. can B.

1. AC input

2. can NOT C.

1. DC output

2. can D.

1. DC output

2. can NOT Page 48 of 100

Question:

(1 point) 49 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is operating at 100% RTP Maintenance has determined that an equalizing charge must be placed on vital battery 1EBA In accordance with OP/1/A/6350/008 (125VDC/120VAC Vital Instrumentation and Control Power System) Encl. 4.13 (Placing and Removing Equalize Charge on 1EBA):

The equalize charge for 1EBA will be from battery charger _____(1)_____.

1EBA Battery room ventilation _____(2)_____ required to be in service.

Which ONE of the following completes the statements above?

A.

1. 1ECA

2. is B.

1. 1ECA

2. is NOT C.

1. ECS

2. is D.

1. ECS

2. is NOT Page 49 of 100

Question:

(1 point) 50 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

D/G 1A has been started and fully loaded for a Periodic Test The following alarms are received:

o 1AD-12, D/1 (DIESEL GEN HX A OUTLET FLOW - LO) o 1AD-11, B/7 (D/G A PANEL TROUBLE)

Lube oil temperature has risen rapidly and is now 207°F Jacket water outlet temperature is 185°F and rising D/G 1A continues to run and is fully loaded In accordance with the Annunciator Response Procedure, which ONE (1) of the following actions will be taken FIRST to address the above conditions?

A.

Start an additional RN pump to increase D/G heat exchanger cooling water flow.

B.

Ensure that 1RN-232A (D/G 1A Hx Inlet Isol) is FULLY open.

C.

Reduce loading on D/G 1A.

D.

Manually trip D/G 1A.

Page 50 of 100

Question:

(1 point) 51 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 2 conditions:

Unit is in Mode 6 VP is in service and refueling is in progress 2A and 2B SSPS are in TEST Subsequently:

A power supply failure occurs on 2EMF-39(L) (Containment Gas Monitor)

Based on the conditions above:

2A VP train is _____(1)_____.

2B VP Supply and Exhaust dampers are _____(2)_____.

Which ONE of the following correctly completes the statements above?

A.

1.

operating

2.

open B.

1.

operating

2.

closed C.

1.

secured

2.

open D.

1.

secured

2.

closed Page 51 of 100

Question:

(1 point) 52 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

Both units are at 100% power RN is aligned to Lake Wylie 1A D/G is removed from service for bar and roll 1EMXG is aligned to Unit 2 Subsequently:

Unit 1 experiences a loss of offsite power 1AD-12 E/2 RN Pit-A Swap To SNSWP is received Which ONE of the following completes the statements below?

1RN-3A (RN P/H Pit A Isol From SNSWP) is _____(1)_____.

2RN-47A (RN Supply X-Over Isolation) is _____(2)_____.

A.

1. open

2. open B.

1. open

2. closed C.

1. closed

2. open D.

1. closed

2. closed Page 52 of 100

Question:

(1 point) 53 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

The Waste Gas (WG) system is in a normal lineup with A Catalytic Hydrogen Recombiner (CHR) in service aligned to A Waste Gas Decay Tank (WGDT)

Subsequently:

A loss of VI to A CHR Oxygen addition valve (1WG-191) occurs 1WG-191 _____(1)_____ close to secure Oxygen addition to A CHR.

Per SLC 16.11-18 (Explosive Gas Mixture) COMMITMENT, the concentration of oxygen in the WG system shall be limited to < _____(2)_____ by volume when hydrogen concentration is > 4%.

Which ONE of the following completes the statements above?

A.

1. will

2. 4%

B.

1. will

2. 2%

C.

1. will NOT

2. 4%

D.

1. will NOT

2. 2%

Page 53 of 100

Question:

(1 point) 54 ILT23 CNS SRO NRC Examination Catawba Nuclear Station During a normal containment air release (VQ), valve _____(1)_____ automatically closes to prevent containment pressure from lowering below 0 PSIG.

The limit on negative (-) containment pressure per Tech Spec 3.6.4 (Containment Pressure) is _____(2)_____ PSIG.

Which ONE of the following completes the statements above?

A.

1. VQ-10 (VQ Fans Disch To Unit Vent)

2. -0.1 B.

1. VQ-3B (VQ Fan Suct From Cont Isol)

2. -0.1 C.

1. VQ-10 (VQ Fans Disch To Unit Vent)

2. -0.3 D.

1. VQ-3B (VQ Fan Suct From Cont Isol)

2. -0.3 Page 54 of 100

Question:

(1 point) 55 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is operating at 100% RTP Subsequently:

A CRDM failure causes Control Bank D rod M4 to fully drop into the core To verify rod M4 has fully dropped into the core, DRPI will show a _____(1)_____ rod and RB indication.

Per Tech Spec 3.1.4 (Rod Group Alignment Limits), if rod M4 can NOT be realigned then the crew will need to perform a Shutdown Margin calculation or initiate boration within a MAXIMUM of _____(2)_____.

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

A.

1. green

2. one hour B.

1. green

2. fifteen minutes C.

1. red

2. one hour D.

1. red

2. fifteen minutes Page 55 of 100

Question:

(1 point) 56 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

The Unit is at 100% RTP Control Banks A, B, & C are 228 steps withdrawn Control Bank D is 215 steps withdrawn 1AD-2 E/10 RPI Non-Urgent Failure is illuminated DRPI displays Data A Failure Subsequently:

The OATC performs the following evolution in accordance with PT/1/A/4600/001 (RCCA Movement Test) o Withdraws Control Bank B to 238 steps as indicated by the applicable step demand counter o Inserts Control Bank B to 228 steps as indicated by the applicable step demand counter Following completion of this step (and plant stabilization):

DRPI accuracy is _____(1)_____.

NC system temperature _____(2)_____ be approximately the same as compared to temperature prior to testing.

Which ONE of the following completes the statements above?

A. 1.

2.

-10/+4 steps will NOT B. 1.

2.

-10/+4 steps will C. 1.

2.

+10/-4 steps will NOT D. 1.

2.

+10/-4 steps will Page 56 of 100

Question:

(1 point) 57 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Shutdown with a Keff of 0.92 Stable count rate on N31 & N32 of 2x102 cps Subsequently:

Reactor startup has commenced All control rod motion stopped when Keff reached 0.995 Immediately following stopping of control rod motion, N31 & N32 count rate was 1.8x103 cps Once count rate stabilizes, N31 & N32 will read approximately __________.

Which ONE of the following completes the statement above?

REFERENCE PROVIDED A.

1.8x103 cps B.

3.2x103 cps C.

3.4x103 cps D.

5.0x103 cps Page 57 of 100

Question:

(1 point) 58 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

The following instrument diagram represents a typical CET measuring circuit:

If the thermocouple measuring junction experiences an open circuit, the associated CET will indicate _____(1)_____ temperature.

If the thermocouple reference junction temperature lowers by 50°F, the associated CETs will indicate 50°F _____(2)_____.

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

A.

1. HIGH

2. HIGHER B.

1. HIGH

2. LOWER C.

1. LOW

2. HIGHER D.

1. LOW

2. LOWER Page 58 of 100

Question:

(1 point) 59 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A Safety Injection has occurred BOP is performing EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) Encl. 2 (Automatic S/I Actions Verification) step to verify proper VE system operation Annulus pressure on both trains reads -2.0 IN. WC Per E-0 Encl. 2; Annulus pressure is too _____(1)_____.

To determine which train of VE may be causing the issue, the operators will observe

_____(2)_____ on each train.

Which ONE of the following completes the statements above?

A.

1. low

2. VE system total flow B.

1. high

2. VE system total flow C.

1. low

2. VE flow to the stack D.

1. high

2. VE flow to the stack Page 59 of 100

Question:

(1 point) 60 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following initial conditions on Unit 1:

Initial:

The Unit was shutdown on March 2 for a refueling outage Current:

The Unit is at 100% power 1A KF Pump is in service Due to an emergent issue, 1B KF Pump has been tagged out and is currently disassembled The date is March 29 Spent Fuel Pool level is 39.9 ft and temperature is 119.6°F.

Subsequently:

1A KF Pump trips for an unknown reason The crew enters AP/1/A/5500/041, (Loss of Spent Fuel Pool Cooling or Level).

The indication on 1EMF-15 (Spent Fuel Bldg refuel Brdg) will FIRST begin to rise significantly following a minimum of _____(1)_____ hours.

To restore Spent Fuel Pool cooling and exit the AP, one KF pump must be restored and the Spent Fuel Pool temperature must be reduced to NO GREATER THAN

_____(2)_____.

Which ONE of the following completes the statements above?

REFERENCE PROVIDED A.

1. 16

2. 125° F B.

1. 36

2. 125° F C.

1. 16

2. 140° F D.

1. 36

2. 140° F Page 60 of 100

Question:

(1 point) 61 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

A reactor trip occurred from 100% RTP All S/G pressures are 1105 PSIG Steam Dump Control is in AUTO

% STM DUMP DEMAND is 76%

Steam Dump Select switch is in PRESS NC Tavg:

o A Loop - 552°F o B Loop - 554°F o C Loop - 553°F o D Loop - 552°F C7A Status Light - LIT C7B Status Light - DARK Tref - 557°F Which ONE of the following identifies ALL steam dump banks which are open (if any)?

A.

No Banks open B.

Bank 1 C.

Banks 1, 2, and 3 D.

Banks 1, 2, 3, 4, and 5 Page 61 of 100

Question:

(1 point) 62 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is operating at 100% RTP Secondary Chemistry contacted the Control Room and informed the crew that Unit 1 is in Action Level 1 for sodium Secondary Chemistry suspects a Main Condenser tube is leaking Crew enters AP/1/A/5500/034 (Secondary Chemistry Out of Spec)

Subsequently:

Secondary Chemistry has verified Unit 1 is in Action Level 1 for sodium via confirmatory sample AP/34 _____(1)_____ require power to be reduced to < 50% RTP within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Once the Unit is < 50% RTP, power operation at < 50% RTP _____(2)_____ continue indefinitely with an Engineering justification.

Which ONE of the following completes the statements above?

A.

1. does

2. can NOT B.

1. does

2. can C.

1. does NOT

2. can NOT D.

1. does NOT

2. can Page 62 of 100

Question:

(1 point) 63 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is operating at 100% RTP RP has requested the setpoints be changed for Trip 1 and 2 on 1EMF1 (522 FF-57 AUX BLDG 522)

Current 1EMF-1 setpoints are as follows:

o Trip 1 - 7.0E+1 cpm o Trip 2 - 1.0E+2 cpm If the BOP inputs a Trip 1 value of 1.2E+2, the EMF module _____(1)_____ accept it.

When performing a source check on an EMF, the Trip 1 and Trip 2 lights

_____(2)_____ illuminate.

Which ONE of the following completes the statements above?

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 63 of 100

Question:

(1 point) 64 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Which ONE of the following evolutions requires a plant announcement in accordance with AD-OP-ALL-1000 (CONDUCT OF OPERATIONS)?

A.

Rolling the Main Turbine B.

Starting the Unit 1 FW Pump C.

Starting a Liquid Waste Release D.

Starting the Weekly Main Turbine Trip Test Page 64 of 100

Question:

(1 point) 65 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

Unit 1 has suffered a loss of Main Feed Pump runback from 100% RTP o Control rods failed to automatically insert on the runback Unit 2 is currently raising power to 100% RTP following Control Valve Movement Testing o An ILT student is manipulating control rods under the instruction of the OATC In accordance with AD-OP-ALL-0203 (Reactivity Management):

An additional Reactor Operator _____(1)_____ required to peer check the Unit 1 OATC manually operating failed control rods.

An additional Reactor Operator _____(2)_____ required to peer check control rod manipulations performed by the ILT student.

Which ONE of the following correctly completes the statements above?

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 65 of 100

Question:

(1 point) 66 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

The unit is at 100% RTP In accordance with Tech Spec 2.1.1 (Reactor Core SLs), Departure from Nucleate Boiling Ratio (DNBR) shall be maintained greater than or equal to _____(1)_____.

If this Safety Limit is exceeded, DNBR must be restored within limits in a MAXIMUM of _____(2)_____.

Which ONE of the following completes the statements above?

A. 1.

2.

1.14 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. 1.

2.

1.14 5 minutes C. 1.

2.

1.3 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. 1.

2.

1.3 5 minutes Page 66 of 100

Question:

(1 point) 67 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Refer to the Actions table in Example 1.3-3 of Tech Spec 1.3 to answer the following:

One Function X train was declared inoperable at 0700 on 01/05 One Function Y train was declared inoperable at 0700 on 01/11 Subsequently:

Function X is restored to OPERABLE at 1000 on 01/11 Which ONE of the following identifies the LATEST time and date that Function Y train must be restored to OPERABLE status?

REFERENCE PROVIDED A.

0700 on 01/12 B.

0700 on 01/14 C.

1000 on 01/14 D.

0700 on 01/15 Page 67 of 100

Question:

(1 point) 68 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

An Operator is performing a valve lineup in the Unit 1 Auxiliary Building While working in the area, the Operator receives a Dose Rate alarm on his Electronic Dosimeter (ED)

After a few seconds, the Dose Rate alarm automatically clears The possibility of a Dose Rate alarm was NOT discussed during the RP brief In accordance with PD-RP-ALL-0001 (Radiation Worker Responsibilities):

the Operator _____(1)_____.

if the Operator receives a Dose alarm, the alarm _____(2)_____.

Which ONE of the following correctly completes the statements above?

A.

1.

must stop work, exit the area, and notify RP immediately

2.

will not clear until the ED is reset B.

1.

must stop work, exit the area, and notify RP immediately

2.

will automatically clear after 10 seconds C.

1.

may continue to work unless two additional dose rate alarms are received

2.

will not clear until the ED is reset D.

1.

may continue to work unless two additional dose rate alarms are received

2.

will automatically clear after 10 seconds Page 68 of 100

Question:

(1 point) 69 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Valid reactor trip annunciator is lit Intermediate Range Startup Rate is positive Power Ranges indicate 6%

All 6900v busses are deenergized 1ETA is deenergized 1B D/G is OFF Safety Injection Actuated status light is lit All S/G levels are 10% and lowering

1 CAPT has tripped due to mechanical overspeed Which ONE (1) of the following indicates the procedure that will have the HIGHEST priority for the conditions above?

A.

ECA-0.0 (Loss of All AC Power)

B.

E-0 (Reactor Trip or Safety Injection)

C.

FR-H.1 (Response to Loss of Secondary Heat Sink)

D.

FR-S.1 (Response to Nuclear Power Generation/ATWS)

Page 69 of 100

Question:

(1 point) 70 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following initial conditions on Unit 1:

Unit operating at 100% RTP for the last two weeks Subsequently:

Reactor power reduced to 50% over a one hour period To maintain reactor power stable during the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , which ONE of the following incremental control rod manipulations will be required?

A.

Withdraw control rods slowly during the entire period B.

Withdraw rods slowly at first, and then insert rods slowly C.

Insert rods slowly during the entire period D.

Insert rods slowly at first, and then withdraw rods slowly Page 70 of 100

Question:

(1 point) 71 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Which ONE of the following describes the reason why burnable poisons are installed in a new reactor core instead of using a higher reactor coolant boron concentration for reactivity control?

A.

To prevent boron precipitation during normal operation.

B.

To establish a more negative moderator temperature coefficient.

C.

To minimize the distortion of the neutron flux distribution caused by soluble boron.

D.

To allow the loading of excessive reactivity in the form of higher fuel enrichment.

Page 71 of 100

Question:

(1 point) 72 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Reactor startup is in progress following a one month shutdown Upon reaching criticality, the operator establishes a positive 0.5 DPM startup rate and stops control rod motion After an additional 5 minutes, reactor power will be _____(1)_____ and startup rate will be _____(2)_____. (Assume reactor power remains below the point of adding heat.)

Which ONE of the following completes the statement above?

A.

1. constant

2. constant B.

1. constant

2. rising C.

1. rising

2. constant D.

1. rising

2. rising Page 72 of 100

Question:

(1 point) 73 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

A saturated steam-water mixture with an inlet quality of 60 percent is flowing through a moisture separator The moisture separator is 100 percent efficient for removing moisture How much moisture will be removed by the moisture separator from 50 lbm of the steam-water mixture?

A.

10 lbm B.

20 lbm C.

30 lbm D.

40 lbm Page 73 of 100

Question:

(1 point) 74 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

The unit is operating at 50% RTP Pzr pressure is 2235 PSIG Pzr Relief Tank (PRT) pressure is 20 PSIG PRT temperature is 125°F A Pzr code safety valve is leaking by its seat Which ONE (1) of the following identifies the approximate temperature that is indicated on the leaking safety valve discharge RTD?

REFERENCE PROVIDED A.

123 - 127°F B.

161 - 165°F C.

227 - 231°F D.

258 - 262°F Page 74 of 100

Question:

(1 point) 75 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

The unit is at 100% RTP Pressurizer Pressure Channel 1 fails low Based on the conditions above, the Reactor Protection System (RPS) setpoint for Channel 1 of _____(1)_____ will _____(2)_____.

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

LEGEND:

T OP

- OVERPOWER T

T OT - OVERTEMPERATURE T

A.

1.

T OP

2. lower B.

1.

T OP

2. rise C.

1.

T OT

2. lower D.

1.

T OT

2. rise Page 75 of 100

Question:

(1 point) 76 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following initial conditions on Unit 1:

Unit is at 100% RTP 1A1 KC pump is in service Subsequently:

Annunciator 1AD-10, A/1 KC SURGE TANK A LO-LO LEVEL - LIT Current 1A KC surge tank level is 10% and lowering AO reports a significant leak in the area of the KC surge tanks on the 594 elevation of the Auxiliary Building The procedure that contains the specific steps to ensure the KC trains are split is

_____(1)_____.

Following the splitting of the KC trains, a reactor trip _____(2)_____ be required.

Which ONE of the following completes the statements above?

A.

1. AP/0/A/5500/030 (Plant Flooding)

2. will B.

1. AP/0/A/5500/030 (Plant Flooding)

2. will NOT C.

1. AP/1/A/5500/021 (Loss of Component Cooling)

2. will D.

1. AP/1/A/5500/021 (Loss of Component Cooling)

2. will NOT Page 76 of 100

Question:

(1 point) 77 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 initial conditions:

The Unit is at 100% RTP Containment Pressure Channel 1 failed High Pressurizer Pressure Channel 1 failed High All required Tech Spec actions have been completed Subsequently:

A loss of 1ERPB occurs 1FO-1, D/5 Hi Cont Press S/I Rx Trip _____(1)_____ be lit.

If the Unit 1 Reactor does NOT trip, an ATWS _____(2)_____ in progress.

Which ONE of the following completes the statements above?

A. 1.

2.

will is B. 1.

2.

will is NOT C. 1.

2.

will NOT is D. 1.

2.

will NOT is NOT Page 77 of 100

Question:

(1 point) 78 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Reactor trip and S/I have occurred from 100% RTP due to a 1B S/G fault inside containment All MSIVs failed to close automatically or manually Transition to EP/1/A/5000/ECA-2.1 (Uncontrolled Depressurization of All Steam Generators) will be directly from _____(1)_____.

Once a controlled cooldown to Mode 5 has begun in ECA-2.1, any boration needed to maintain shutdown margin will be made from the _____(2)_____.

Which ONE of the following completes the statements above?

Procedure Legend:

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

EP/1/A/5000/E-2 (Faulted Steam Generator Isolation)

A.

1. E-0

2. Boric Acid Tank B.

1. E-0

2. FWST C.

1. E-2

2. Boric Acid Tank D.

1. E-2

2. FWST Page 78 of 100

Question:

(1 point) 79 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

Unit is at 100% RTP 1ERPA has de-energized due to an inverter failure The crew is preparing to re-energize 1ERPA via 1VRD per AP/1/A/5500/029 (Loss of Vital or Aux Control Power)

The power supply transfer required to re-align 1ERPA to 1VRD will be performed via a(n) _____(1)_____.

Once 1ERPA is aligned to 1VRD, the crew _____(2)_____ exit the action statement of Tech Spec 3.8.7 (Inverters - Operating).

Which ONE of the following correctly completes the statements above?

A.

1. manual bypass switch

2. will B.

1. manual bypass switch

2. will NOT C.

1. automatic transfer switch

2. will D.

1. automatic transfer switch

2. will NOT Page 79 of 100

Question:

(1 point) 80 ILT23 CNS SRO NRC Examination Catawba 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 In accordance with AP/22:

As pressure lowers, the CRS will FIRST direct performance of _____(1)_____.

The CRS will direct performance of valve failure position verification once pressure lowers below _____(2)_____ psig.

Which ONE of the following correctly completes the statements above?

PROCEDURE LEGEND AP/22 Enclosure 2 (Isolating VI Dryer Purge Valves)

OP/0/A/6450/013 (Instrument Air System), Enclosure 4.1 (Startup and Operation of the Station Air System)

A.

1.

2.

40 B.

1.

2.

60 C.

1..1

2.

40 D.

1..1

2.

60 Page 80 of 100

Question:

(1 point) 81 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

A LOCA has occurred on Unit 1 NC system is saturated at 900 PSIG RVLIS Lower Range level is 70%

The crew has implemented EP/1/A/5000/ECA-1.1 (Loss Of Emergency Coolant Recirc) and has reached the step to Verify SI termination criteria Which ONE of the following completes the statements below?

In accordance with ECA-1.1, based on the conditions above, SI can be terminated when NC system subcooling is greater than a MINIMUM of

_____(1)_____.

If Safety Injection CANNOT be terminated, ECA-1.1 will direct the crew to reduce SI flow to _____(2)_____.

A.

1. 10°F

2. conserve FWST inventory B.

1. 50°F

2. conserve FWST inventory C.

1. 10°F

2. allow Cold Leg Accumulator injection D.

1. 50°F

2. allow Cold Leg Accumulator injection Page 81 of 100

Question:

(1 point) 82 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 2 initial conditions:

The OATC is manually inserting control rods per AP/2/A/5500/003 (Load Rejection)

Subsequently:

While inserting rods, the OATC notes the following:

o DRPI indication for Control Bank D rod M4 is YELLOW with a DATA B indication above the rod o DRPI indication for rod M4 indicates 198 steps o Demand position counters for Groups 1 & 2 are blinking with position indication of 187 steps In accordance with T.S. 3.1.7 (Rod Position Indication) bases, LCO 3.1.7

_____(1)_____ met.

The rod group alignment limits of T.S. 3.1.4 (Rod Group Alignment Limits)

_____(2)_____ met.

Which ONE of the following completes the statements above?

A.

1. is

2. are B.

1. is NOT

2. are C.

1. is

2. are NOT D.

1. is NOT

2. are NOT Page 82 of 100

Question:

(1 point) 83 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 initial conditions:

The crew has begun a Reactor startup Intermediate Range Channel N-35 begins to operate erratically Reactor Power is stabilized at 10-3 % RTP AP/1/A/5500/016 (Malfunction of Nuclear Instrumentation System) Case III (Intermediate Range Malfunction) is entered Subsequently:

Power remains stable at 10-3 % RTP AP/16 actions are complete The 1/N-35A I/R CHANNEL 1 TRIP BYPASS status light on 1SI-19

_____(1)_____ LIT.

Per Tech Spec 3.3.1, (RTS Instrumentation), the startup to MODE 1

_____(2)_____ continue.

Which ONE of the following correctly completes the statements above?

A.

1. is

2. may B.

1. is

2. may NOT C.

1. is NOT

2. may D.

1. is NOT

2. may NOT Page 83 of 100

Question:

(1 point) 84 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Core refueling is in progress Following insertion of a fuel assembly, the gripper failed to dis-engage To dis-engage the gripper from the fuel assembly, the Fuel Handling SRO will verify the load cell reading is less than a MAXIMUM of _____(1)_____.

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)_____.

Which ONE of the following completes the statements above?

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 84 of 100

Question:

(1 point) 85 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following on Unit 1:

Unit is at 100% RTP The Upper Airlock Inner door will not close and has been declared INOPERABLE Based on the conditions above:

Per TS 3.6.2 (Containment Air Locks), the Upper Airlock Outer Door is required to be CLOSED in ____(1)____ hour(s).

Per TS 3.6.1 (Containment), following closure of Upper Airlock Outer Door, Containment ____(2)____ OPERABLE.

Which ONE of the following completes the statements above?

A.

1. one

2. is NOT B.

1. one

2. is C.

1. four

2. is NOT D.

1. four

2. is Page 85 of 100

Question:

(1 point) 86 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following timeline on Unit 1:

2/1 0200:

Unit operating at 100% RTP Mixed Bed Demineralizer 1B placed in service to replace the resin in Mixed Bed Demineralizer 1A 2/1 1000:

Primary Chemistry reports that Unit 1 is in Action Level 2 for Chlorides (sample has been verified)

AP/0/A/5500/035 (Primary Chemistry Out of Specification) entered 2/1 2200:

Primary Chemistry reports that Unit 1 is now in Action Level 3 for Chlorides (sample has been verified)

In accordance with AP/35:

Unit 1 must be in Mode 3 by _____(1)_____.

Once in Mode 3, within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , Unit 1 must be cooled down to less than a MAXIMUM of _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. 2/2 at 0100

2. 200°F B.

1. 2/2 at 0100

2. 250°F C.

1. 2/2 at 1300

2. 200°F D.

1. 2/2 at 1300

2. 250°F Page 86 of 100

Question:

(1 point) 87 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

Unit is at 100% RTP Steam dump valves are in PRESS mode with steam dump controller in MANUAL and closed Subsequently:

1B NC pump tripped An automatic reactor trip has NOT occurred and attempts to manually trip the reactor from the control room have been unsuccessful The immediate actions of EP/1/A/5000/FR-S.1 (Response to Nuclear Power Generation/ATWS) have just been completed Per FR-S.1, if Pressurizer pressure exceeds the PORV lift setpoint, Pressurizer PORVs will be operated to ensure Pressurizer pressure is maintained less than a MAXIMUM of _____(1)_____ PSIG in order to _____(2)_____.

Which ONE of the following correctly completes the statements above?

A.

1. 2135

2. allow NC system boron injection B.

1. 2135

2. prevent lifting a Pressurizer Safety Valve C.

1. 2315

2. allow NC system boron injection D.

1. 2315

2. prevent lifting a Pressurizer Safety Valve Page 87 of 100

Question:

(1 point) 88 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 2 conditions:

Safety Injection termination is in progress per EP/2/A/5000/ECA-2.1 (Uncontrolled Depressurization Of All Steam Generators)

OATC reports that 2B S/G MSIV has closed and that 2B S/G pressure is trending up Based on the conditions above, which ONE of the following indicates the response directed by the CRS?

PROCEDURE LEGEND:

EP/2/A/5000/E-2 (Faulted S/G Isolation)

EP/2/A/5000/ES-1.1 (SI Termination)

A.

Immediately transition to E-2 to verify that 2B S/G is intact B.

Immediately transition to step 10 of ES-1.1 to complete Safety Injection termination C.

Complete Safety Injection termination per ECA-2.1 and then transition to E-2 D.

Complete Safety Injection termination and initiate a cooldown to cold shutdown per ECA-2.1 Page 88 of 100

Question:

(1 point) 89 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 conditions:

1A D/G is paralleled to the grid for testing per OP/1/A/6350/002 1A RN Pump is in service Subsequently, the following sequence of events occurs:

1A D/G Load is reduced to 200KW in anticipation of opening 1ETA-18 (Diesel Gen Bkr)

Prior to opening 1ETA-18, 1ETA-03 (ETA Norm Fdr Frm ATC) spuriously opens Following this event:

The Blackout Sequencer Actuated Train A status light on 1SI-14 _____(1)_____

illuminated.

The procedure the crew will use to RESTORE normal power to 1ETA and to SECURE the 1A Diesel Generator is _____(2)_____.

Which ONE of the following correctly completes the statements above?

PROCEDURE LEGEND:

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

OP/1/A/6350/002 (Diesel Generator Operation)

A.

1.

is

2.

OP/1/A/6350/002 B.

1.

is NOT

2.

OP/1/A/6350/002 C.

1.

is

2.

AP/1/A/5500/007 D.

1.

is NOT

2.

AP/1/A/5500/007 Page 89 of 100

Question:

(1 point) 90 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

Both Units operating at 100% RTP 2A RN pump is in service Alarm 2AD-12, C/2 RN PMP A STRAINER HI D/P is received Valve 2RN-30A (RN Strnr 2A Backflush Isol) is closed and can NOT be opened manually Unit 1 _____(1)_____ in LCO 3.7.8 Condition A (One NSWS train inoperable).

Per the Tech Spec 3.7.8 bases, sufficient RN capacity to supply post LOCA loads on one unit and shutdown and cooldown loads on the other unit _____(2)_____ exist.

Which ONE of the following completes the statements above?

A.

1. is

2. does B.

1. is NOT

2. does C.

1. is

2. does NOT D.

1. is NOT

2. does NOT Page 90 of 100

Question:

(1 point) 91 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 2:

Unit is at 100% RTP Pressurizer Pressure Channel 2 has failed HIGH All required Tech Spec actions have been completed Subsequently:

A fault on the Pressurizer Pressure Channel 4 bistable has caused it to fail low Automatic reactor trip and S/I _____(1)_____ occur.

If S/I does occur, terminating S/I and establishing normal charging will be performed by completing steps in _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. did NOT

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

B.

1. did NOT

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

C.

1. did

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

D.

1. did

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

Page 91 of 100

Question:

(1 point) 92 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions on Unit 1:

Unit is at 10% RTP and holding during a power ascension Subsequently, One condenser steam dump fails OPEN The CRS has implemented AP/1/A/5500/028 (Secondary Steam Leak)

Pzr level is stable NC system temperature is 555°F and lowering at 1° per minute Entry into the Action Statement of T.S 3.4.2 (RCS Minimum Temperature for Criticality) will FIRST be required in _____(1)_____ minutes.

In accordance with T. S. 3.4.2 (RCS Minimum Temperature for Criticality) Bases, maintaining NCS temperature within the nominal operating envelope is required to ensure _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. 3

2. the reactor remains subcritical in the event of a Rx trip B.

1. 3

2. proper indication and response of protective instrumentation C.

1. 5

2. the reactor remains subcritical in the event of a Rx trip D.

1. 5

2. proper indication and response of protective instrumentation Page 92 of 100

Question:

(1 point) 93 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

0815 A planned release of the contents of Waste Monitor Tank (WMT) B was initiated per OP/0/B/6500/113 (Operations Liquid Waste Release) 0820 During a control board walkdown, the BOP sees 0EMF-49 (Liquid Waste Discharge Lo Range) Trip 1 & 2 lights lit 0EMF-49 is indicating 5E+06 cpm and stable 1WL-124 (Waste Monitor Tank Pumps Discharge) is OPEN and can NOT be closed from the control room 0830 Unit Supervisor dispatches SPOC to attempt to locally close 1WL-124 0915 SPOC reports that they will be able to close 1WL-124, but it will take an additional 10 minutes to complete 0EMF-49 _____(1)_____ FUNCTIONAL per SLC 16.11-2 (Radioactive Liquid Effluent Monitoring Instrumentation).

At time 0915, the SM _____(2)_____ declare an Unusual Event.

Which ONE of the following completes the statements above?

REFERENCE PROVIDED A.

1. is NOT

2. should B.

1. is NOT

2. should NOT C.

1. is

2. should D.

1. is

2. should NOT Page 93 of 100

Question:

(1 point) 94 ILT23 CNS SRO NRC Examination Catawba Nuclear Station In accordance with AD-OP-ALL-0111 (Operations Communications):

_____(1)_____ are used to distribute information that does NOT impact the operation of the plant to the Operations department and must be approved for issuance at a MINIMUM by the _____(2)_____.

Which ONE of the following completes the statement above?

A.

1. Operations Supplemental Information Packages (OSIP)

2. SM B.

1. Operations Supplemental Information Packages (OSIP)

2. AOM Shift C.

1. Standing Instructions

2. SM D.

1. Standing Instructions

2. AOM Shift Page 94 of 100

Question:

(1 point) 95 ILT23 CNS SRO NRC Examination Catawba Nuclear Station In accordance with Tech Spec 5.1 (RESPONSIBILITY), an STA with an active SRO license on the unit, may assume the duties of the Control Room Supervisor provided:

the CRS or relief SRO is available to return to the control room within

_____(1)_____ minutes AND each period in which the STA assumes SRO duties _____(2)_____ exceed 15 minutes in duration.

(Assume MODE 1 conditions)

Which ONE of the following correctly completes the statements above?

A.

1.

10

2.

can B.

1.

10

2.

can NOT C.

1.

15

2.

can D.

1.

15

2.

can NOT Page 95 of 100

Question:

(1 point) 96 ILT23 CNS SRO NRC Examination Catawba Nuclear Station In accordance with AD-WC-ALL-0420 (Shutdown Risk Management) and AD-WC-ALL-0340 (Outage Schedule Development and Revision Process):

Defense in Depth (DID) sheets are first REQUIRED for risk management once

_____(1)_____ is reached during the shutdown.

The Plant Condition Mode Change (PCMC) reports _____(2)_____ intended to track operability of structures, systems, or components (SSC) required by Technical Specifications (TS) or Selected License Commitments (SLC).

Which ONE of the following correctly completes the statements above?

A.

1.

Mode 3

2.

are NOT B.

1.

Mode 3

2.

are C.

1.

Mode 4

2.

are NOT D.

1.

Mode 4

2.

are Page 96 of 100

Question:

(1 point) 97 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following:

2/10/23 @ 1000 While checking 1A D/G governor oil level, a technician inadvertently moved an adjustment knob.

2/11/23 @ 1200 1A D/G failed to meet the acceptance criteria during a scheduled PT. Misadjusted governor control is found during the PT.

2/11/23 @ 1800 1A D/G clearance is hung for governor repair.

2/12/23 @ 1000 Clearance is removed and 1A D/G is returned to service.

2/12/23 @ 2000 1A D/G surveillance PT is completed successfully.

The 1A D/G was required to be declared inoperable at _____(1)_____.

The 1A D/G can be declared operable at _____(2)_____.

Which ONE of the following completes the statements above?

A.

1. 2/10/23 1000

2. 2/12/23 1000 B.

1. 2/10/23 1000

2. 2/12/23 2000 C.

1. 2/11/23 1200

2. 2/12/23 1000 D.

1. 2/11/23 1200

2. 2/12/23 2000 Page 97 of 100

Question:

(1 point) 98 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

A General Emergency has been declared An AO must be dispatched from the Operations Support Center to an area with an identified radiation field of 110 R/hr in order to isolate the pathway for a large release to the environment The operator will be in the area for 15 minutes In accordance with AD-EP-ALL-0205 (Emergency Exposure Controls), the AO selected to perform this isolation _____(1)_____ required to be a volunteer.

Per AD-RP-ALL-4012 (Planned Special Exposure), this job _____(2)_____

considered a Planned Special Exposure (PSE).

Which ONE of the following completes the statements above?

A.

1. is

2. is B.

1. is NOT

2. is C.

1. is

2. is NOT D.

1. is NOT

2. is NOT Page 98 of 100

Question:

(1 point) 99 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following Unit 1 initial conditions:

Initially operating at 100% RTP when a SGTL occurs on 1A S/G Crew entered AP/1/A/5500/010 (Reactor Coolant Leak) Case 1 (Steam Generator Tube Leak) 1EMF-33 (Condenser Air Ejector Exhaust) is in Trip 2 Pressurizer Level stabilized at 40%

VCT level was not able to be maintained by normal makeup NV pump suction was aligned to the FWST and the reactor was tripped Currently:

The crew has transitioned from EP/1/A/5000/E-0 (Reactor Trip or Safety Injection)

Pressurizer Level is stable at 20%

AOs have reported that 1A S/G PORV is leaking by and steam has been observed coming from the top of the outside doghouse The classification for this event is _____(1)_____.

Emergency Notification Form Line 6 release Is Occurring block _____(2)_____

required to be checked.

Which ONE of the following correctly completes the statements above?

REFERENCE PROVIDED A.

1.

SU5.1

2.

is NOT B.

1.

FA1.1

2.

is NOT C.

1.

SU5.1

2.

is D.

1.

FA1.1

2.

is Page 99 of 100

Question:

(1 point) 100 ILT23 CNS SRO NRC Examination Catawba Nuclear Station Given the following conditions:

The Control Room has been notified by the NRC Headquarters Operations Center that a 747 commercial aircraft has been hijacked Ground intelligence indicates a nuclear plant is the intended target The airplane's current flight path will intersect with Catawba in 20 minutes In accordance with AP/0/A/5500/046 (Hostile Aircraft Activity),

The CRS will transition to _____(1)_____.

Non-essential plant personnel _____(2)_____ be directed to relocate to Warehouse 9.

Which ONE of the following correctly completes the statements above?

PROCEDURE LEGEND:

Enclosure 1 (Imminent Hostile Aircraft Response)

Enclosure 2 (Probable Hostile Aircraft Response)

A.

1.

2.

will B.

1.

2.

will C.

1.

2.

will NOT D.

1.

2.

will NOT Page 100 of 100

S/I FLOW REQUIRED TO MATCH DECAY HEAT PAGE NO.

CNS LOSS OF EMERGENCY COOLANT RECIRCULATION - Page 1 of 1 82 of 88 EP/1/A/5000/ECA-1.1 Revision 43 Minimum S/I Flowrate Versus Time After Trip

ES 4.3, Page 7 of 7 EQUATIONS Q = m cpT N = S/(1 Keff)

Q = m h CR11 Keff1= CR21 Keff2 Q = UAT 1/M = CR1/CRx Q m Nat Circ 3

A = r2 T m Nat Circ 2

F = PA Keff = 1/(1 )

m = Av

= (Keff 1)/Keff W Pump = m P SUR = 26.06/

P = I2R

= eff eff P = IE PA = 3IE

=

+

eff 1 + eff PT = 3IEpf

= 1.0 x 104 sec PR = 3IEsin eff = 0.1 sec1 (for > 0)

Thermal Efficiency = Net Work Out/Energy In DRW tip 2 /avg 2

g(z2 z1) gc

+ (v2 2 v1 2) 2gc

+ (P2 P1) + (u2 u1) + (q w) = 0 P = Poet/

P = Po10SUR(t)

A = Aoet g = 32.2 ft/sec2 gc = 32.2 lbm-ft/lbf-sec2 CONVERSIONS 1 MW = 3.41 x 106 Btu/hr

= (5/9)(32) 1 ftwater 3

= 7.48 gal 1 hp = 2.54 x 103 Btu/hr

= (9/5)() + 32 1 galwater = 8.35 lbm 1 Btu = 778 ft-lbf 1 kg = 2.21 lbm 1 Curie = 3.7 x 1010 dps Form 4.3-1 Generic Fundamental Equations and Conversion Sheet

The heatup tables in this enclosure were created from data documented in DPC-1201.30-00-0012 Loss of Spent Fuel Pool Cooling Heat Up Times Due to Decay Heat.

NOTE

If level lower than 39.0 feet, then tables are non-conservative.

Determine Unit 1 Spent Fuel heatup time as follows:

1.

WHEN heatup time determined, THEN notify Station Management.

a.

Determine heatup time by performing one of the following:

b.

IF in Mode 6 and fuel offload has commenced or No Mode (Full Core Offload), THEN observe Note prior to Step 2 and GO TO Step 2.

OR IF in Mode 1, 2, 3, 4, 5 or 6 (No Core Offload), THEN observe Note prior to Step 3 and GO TO Step 3.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 1 of 8 29 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

"Time After Shutdown" starts when reactor was shutdown for the current refueling outage.

NOTE Use the following table to determine approximate time (hours) for Unit 1 Spent Fuel Pool to reach 200°F during Mode 6 or No Mode (Full Core Offload):

2.

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F-Full Core Offload Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 0

12 11 11 10 10 9

8 7

6 5

4 3

12 11 11 10 10 9

8 7

6 5

4 6

12 11 11 10 10 9

8 7

6 5

4 9

16 15 15 14 13 12 10 9

8 6

5 12 18 17 16 16 15 13 12 10 9

7 6

15 20 19 18 17 16 14 13 11 9

8 6

18 21 20 19 18 17 16 14 12 10 8

7 21 22 21 20 19 18 17 15 13 11 9

7 24 24 22 21 20 19 17 15 13 11 9

7 27 25 24 22 21 20 18 16 14 12 10 8

30 26 25 23 22 21 19 17 15 12 10 8

Table continues on next page PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 2 of 8 30 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F-Full Core Offload Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 33 27 25 24 23 22 20 17 15 13 11 8

36 28 26 25 24 23 20 18 16 13 11 9

39 28 27 26 25 24 21 19 16 14 11 9

42 29 28 27 25 25 22 19 17 14 12 9

45 30 29 27 26 26 22 20 17 15 12 10 48 31 29 28 27 26 23 20 18 15 12 10 51 32 30 29 27 27 23 21 18 15 13 10 54 32 31 29 28 28 24 21 18 16 13 10 57 33 32 30 29 28 25 22 19 16 13 10 60 34 32 31 29 29 25 22 19 16 14 11 70 36 34 33 31 30 27 23 20 17 14 11 80 38 36 34 33 32 28 25 22 18 15 12 90 39 38 36 34 34 29 26 23 19 16 13 100 41 39 37 36 35 31 27 23 20 17 13 110 43 41 39 37 36 32 28 24 21 17 14 120 44 42 40 38 38 33 29 25 21 18 14 Greater than 120 days refer to CNEI-0400-225 Loss of Spent Fuel Pool Cooling Heat Up Times Due to Decay Heat PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 3 of 8 31 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

2. (Continued)

"Time After Shutdown" is the time the reactor was shutdown for a refueling outage. If current core offload has not begun, the number of days since the last refueling outage began is used. If this date is unknown, it is acceptable to use the following formula to estimate "Time After Shutdown":

NOTE

______________+ 20 days = ____________________________

Current EFPDs "Time After Shutdown (Days)"

This table is used prior to core offload or after core is completely reloaded.

Use the following table to determine approximate time (hours) for Unit 1 Spent Fuel Pool to reach 200°F during Modes 1, 2, 3, 4, 5 or 6 (No Core Offload):

3.

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F-Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 0

30 28 27 26 25 22 19 17 14 12 9

3 30 28 27 26 25 22 19 17 14 12 9

6 36 34 33 31 30 27 24 21 18 14 11 9

40 39 37 35 33 30 27 23 20 16 13 12 44 42 40 38 36 33 29 25 21 18 14 15 47 45 43 41 39 35 31 27 23 19 15 18 49 47 45 43 41 36 32 28 24 20 16 21 52 49 47 45 42 38 34 29 25 21 16 24 53 51 48 46 44 39 35 30 26 21 17 27 55 52 50 48 45 41 36 31 27 22 18 30 56 54 52 49 47 42 37 32 28 23 18 Table continues on next page PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 4 of 8 32 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 33 58 55 53 50 48 43 38 33 28 23 19 36 59 57 54 52 49 44 39 34 29 24 19 39 61 58 55 53 50 45 40 35 30 25 20 42 62 59 57 54 51 46 41 36 30 25 20 45 63 60 58 55 52 47 42 36 31 26 20 48 64 62 59 56 53 48 42 37 31 26 21 51 65 63 60 57 54 49 43 38 32 27 21 54 66 64 61 58 55 50 44 38 33 27 21 57 67 65 62 59 56 50 44 39 33 27 22 60 68 65 63 60 57 51 45 39 34 28 22 70 71 68 65 62 59 53 47 41 35 29 23 80 74 71 68 65 61 55 49 43 36 30 24 90 77 73 70 67 63 57 50 44 38 31 25 100 79 75 72 69 65 59 52 45 39 32 25 110 81 77 74 71 67 60 53 47 40 33 26 Table continues on next page.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 5 of 8 33 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 120 83 79 76 72 69 62 55 48 41 34 27 130 85 81 78 74 70 63 56 49 42 34 27 140 87 83 79 75 72 65 57 50 42 35 28 150 88 84 81 77 73 66 58 51 43 36 29 160 90 86 82 78 74 67 59 52 44 37 29 170 91 87 84 80 76 68 60 53 45 37 30 180 93 89 85 81 77 69 61 53 46 38 30 190 94 90 86 82 78 70 62 54 46 38 30 200 96 91 87 83 79 71 63 55 47 39 31 210 97 93 89 84 80 72 64 56 48 39 31 220 98 94 90 86 81 73 65 56 48 40 32 230 99 95 91 87 82 74 66 57 49 40 32 240 100 96 92 88 83 75 66 58 49 41 33 250 102 97 93 89 84 76 67 58 50 41 33 260 103 98 94 90 85 77 68 59 50 42 33 270 104 99 95 91 86 77 68 60 51 42 34 Table continues on next page.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 6 of 8 34 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 280 105 100 96 91 87 78 69 60 52 43 34 290 106 101 97 92 88 79 70 61 52 43 34 300 107 102 98 93 89 80 71 62 53 44 35 310 108 103 99 94 90 80 71 62 53 44 35 320 109 104 100 95 90 81 72 63 53 44 35 330 110 105 100 96 91 82 72 63 54 45 36 340 111 106 101 97 92 83 73 64 54 45 36 350 112 107 102 97 93 83 74 64 55 45 36 360 112 108 103 98 93 84 74 65 55 46 36 370 113 108 104 99 94 85 75 65 56 46 37 380 114 109 104 100 95 85 75 66 56 47 37 390 115 110 105 100 95 86 76 66 56 47 37 400 116 111 106 101 96 86 76 67 57 47 38 410 117 112 107 102 97 87 77 67 57 48 38 420 117 112 107 102 97 88 77 68 58 48 38 430 118 113 108 103 98 88 78 68 58 48 38 Table continues on next page.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 7 of 8 35 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 440 119 114 109 104 99 89 78 68 58 48 39 450 120 115 109 104 99 89 79 69 59 49 39 460 120 115 110 105 100 90 79 69 59 49 39 470 121 116 111 106 100 90 80 70 59 49 39 480 122 117 111 106 101 91 80 70 60 50 39 490 122 117 112 107 102 91 81 70 60 50 40 500 123 118 113 107 102 92 81 71 61 50 40 510 124 119 113 108 103 92 82 71 61 50 40 520 125 119 114 109 103 93 82 72 61 51 40 530 125 120 115 109 104 93 83 72 62 51 41 540 126 120 115 110 104 94 83 72 62 51 41 550 126 121 116 110 105 94 83 73 62 52 41 560 127 122 116 111 105 95 84 73 62 52 41 570 128 122 117 111 106 95 84 74 63 52 41 580 128 123 117 112 106 96 85 74 63 52 42 590 129 123 118 112 107 96 85 74 63 53 42 600 130 124 119 113 107 97 88 75 64 53 42 PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 8 of 8 36 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Completion Times 1.3 1.3 Completion Times (continued)

Catawba Units 1 and 2 1.3-7 Amendment Nos. 298/294 EXAMPLES EXAMPLE 1.3-3 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One Function X train inoperable.

A.1 Restore Function X train to OPERABLE status.

7 days AND 10 days from discovery of failure to meet the LCO B.

One Function Y train inoperable.

B.1 Restore Function Y train to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND 10 days from discovery of failure to meet the LCO C.

One Function X train inoperable.

AND One Function Y train inoperable.

C.1 Restore Function X train to OPERABLE status.

OR C.2 Restore Function Y train to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 72 hours

S/I FLOW REQUIRED TO MATCH DECAY HEAT PAGE NO.

CNS LOSS OF EMERGENCY COOLANT RECIRCULATION - Page 1 of 1 82 of 88 EP/1/A/5000/ECA-1.1 Revision 43 Minimum S/I Flowrate Versus Time After Trip

ES 4.3, Page 7 of 7 EQUATIONS Q = m cpT N = S/(1 Keff)

Q = m h CR11 Keff1= CR21 Keff2 Q = UAT 1/M = CR1/CRx Q m Nat Circ 3

A = r2 T m Nat Circ 2

F = PA Keff = 1/(1 )

m = Av

= (Keff 1)/Keff W Pump = m P SUR = 26.06/

P = I2R

= eff eff P = IE PA = 3IE

=

+

eff 1 + eff PT = 3IEpf

= 1.0 x 104 sec PR = 3IEsin eff = 0.1 sec1 (for > 0)

Thermal Efficiency = Net Work Out/Energy In DRW tip 2 /avg 2

g(z2 z1) gc

+ (v2 2 v1 2) 2gc

+ (P2 P1) + (u2 u1) + (q w) = 0 P = Poet/

P = Po10SUR(t)

A = Aoet g = 32.2 ft/sec2 gc = 32.2 lbm-ft/lbf-sec2 CONVERSIONS 1 MW = 3.41 x 106 Btu/hr

= (5/9)(32) 1 ftwater 3

= 7.48 gal 1 hp = 2.54 x 103 Btu/hr

= (9/5)() + 32 1 galwater = 8.35 lbm 1 Btu = 778 ft-lbf 1 kg = 2.21 lbm 1 Curie = 3.7 x 1010 dps Form 4.3-1 Generic Fundamental Equations and Conversion Sheet

The heatup tables in this enclosure were created from data documented in DPC-1201.30-00-0012 Loss of Spent Fuel Pool Cooling Heat Up Times Due to Decay Heat.

NOTE

If level lower than 39.0 feet, then tables are non-conservative.

Determine Unit 1 Spent Fuel heatup time as follows:

1.

WHEN heatup time determined, THEN notify Station Management.

a.

Determine heatup time by performing one of the following:

b.

IF in Mode 6 and fuel offload has commenced or No Mode (Full Core Offload), THEN observe Note prior to Step 2 and GO TO Step 2.

OR IF in Mode 1, 2, 3, 4, 5 or 6 (No Core Offload), THEN observe Note prior to Step 3 and GO TO Step 3.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 1 of 8 29 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

"Time After Shutdown" starts when reactor was shutdown for the current refueling outage.

NOTE Use the following table to determine approximate time (hours) for Unit 1 Spent Fuel Pool to reach 200°F during Mode 6 or No Mode (Full Core Offload):

2.

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F-Full Core Offload Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 0

12 11 11 10 10 9

8 7

6 5

4 3

12 11 11 10 10 9

8 7

6 5

4 6

12 11 11 10 10 9

8 7

6 5

4 9

16 15 15 14 13 12 10 9

8 6

5 12 18 17 16 16 15 13 12 10 9

7 6

15 20 19 18 17 16 14 13 11 9

8 6

18 21 20 19 18 17 16 14 12 10 8

7 21 22 21 20 19 18 17 15 13 11 9

7 24 24 22 21 20 19 17 15 13 11 9

7 27 25 24 22 21 20 18 16 14 12 10 8

30 26 25 23 22 21 19 17 15 12 10 8

Table continues on next page PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 2 of 8 30 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F-Full Core Offload Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 33 27 25 24 23 22 20 17 15 13 11 8

36 28 26 25 24 23 20 18 16 13 11 9

39 28 27 26 25 24 21 19 16 14 11 9

42 29 28 27 25 25 22 19 17 14 12 9

45 30 29 27 26 26 22 20 17 15 12 10 48 31 29 28 27 26 23 20 18 15 12 10 51 32 30 29 27 27 23 21 18 15 13 10 54 32 31 29 28 28 24 21 18 16 13 10 57 33 32 30 29 28 25 22 19 16 13 10 60 34 32 31 29 29 25 22 19 16 14 11 70 36 34 33 31 30 27 23 20 17 14 11 80 38 36 34 33 32 28 25 22 18 15 12 90 39 38 36 34 34 29 26 23 19 16 13 100 41 39 37 36 35 31 27 23 20 17 13 110 43 41 39 37 36 32 28 24 21 17 14 120 44 42 40 38 38 33 29 25 21 18 14 Greater than 120 days refer to CNEI-0400-225 Loss of Spent Fuel Pool Cooling Heat Up Times Due to Decay Heat PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 3 of 8 31 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

2. (Continued)

"Time After Shutdown" is the time the reactor was shutdown for a refueling outage. If current core offload has not begun, the number of days since the last refueling outage began is used. If this date is unknown, it is acceptable to use the following formula to estimate "Time After Shutdown":

NOTE

______________+ 20 days = ____________________________

Current EFPDs "Time After Shutdown (Days)"

This table is used prior to core offload or after core is completely reloaded.

Use the following table to determine approximate time (hours) for Unit 1 Spent Fuel Pool to reach 200°F during Modes 1, 2, 3, 4, 5 or 6 (No Core Offload):

3.

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F-Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 0

30 28 27 26 25 22 19 17 14 12 9

3 30 28 27 26 25 22 19 17 14 12 9

6 36 34 33 31 30 27 24 21 18 14 11 9

40 39 37 35 33 30 27 23 20 16 13 12 44 42 40 38 36 33 29 25 21 18 14 15 47 45 43 41 39 35 31 27 23 19 15 18 49 47 45 43 41 36 32 28 24 20 16 21 52 49 47 45 42 38 34 29 25 21 16 24 53 51 48 46 44 39 35 30 26 21 17 27 55 52 50 48 45 41 36 31 27 22 18 30 56 54 52 49 47 42 37 32 28 23 18 Table continues on next page PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 4 of 8 32 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 33 58 55 53 50 48 43 38 33 28 23 19 36 59 57 54 52 49 44 39 34 29 24 19 39 61 58 55 53 50 45 40 35 30 25 20 42 62 59 57 54 51 46 41 36 30 25 20 45 63 60 58 55 52 47 42 36 31 26 20 48 64 62 59 56 53 48 42 37 31 26 21 51 65 63 60 57 54 49 43 38 32 27 21 54 66 64 61 58 55 50 44 38 33 27 21 57 67 65 62 59 56 50 44 39 33 27 22 60 68 65 63 60 57 51 45 39 34 28 22 70 71 68 65 62 59 53 47 41 35 29 23 80 74 71 68 65 61 55 49 43 36 30 24 90 77 73 70 67 63 57 50 44 38 31 25 100 79 75 72 69 65 59 52 45 39 32 25 110 81 77 74 71 67 60 53 47 40 33 26 Table continues on next page.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 5 of 8 33 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 120 83 79 76 72 69 62 55 48 41 34 27 130 85 81 78 74 70 63 56 49 42 34 27 140 87 83 79 75 72 65 57 50 42 35 28 150 88 84 81 77 73 66 58 51 43 36 29 160 90 86 82 78 74 67 59 52 44 37 29 170 91 87 84 80 76 68 60 53 45 37 30 180 93 89 85 81 77 69 61 53 46 38 30 190 94 90 86 82 78 70 62 54 46 38 30 200 96 91 87 83 79 71 63 55 47 39 31 210 97 93 89 84 80 72 64 56 48 39 31 220 98 94 90 86 81 73 65 56 48 40 32 230 99 95 91 87 82 74 66 57 49 40 32 240 100 96 92 88 83 75 66 58 49 41 33 250 102 97 93 89 84 76 67 58 50 41 33 260 103 98 94 90 85 77 68 59 50 42 33 270 104 99 95 91 86 77 68 60 51 42 34 Table continues on next page.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 6 of 8 34 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 280 105 100 96 91 87 78 69 60 52 43 34 290 106 101 97 92 88 79 70 61 52 43 34 300 107 102 98 93 89 80 71 62 53 44 35 310 108 103 99 94 90 80 71 62 53 44 35 320 109 104 100 95 90 81 72 63 53 44 35 330 110 105 100 96 91 82 72 63 54 45 36 340 111 106 101 97 92 83 73 64 54 45 36 350 112 107 102 97 93 83 74 64 55 45 36 360 112 108 103 98 93 84 74 65 55 46 36 370 113 108 104 99 94 85 75 65 56 46 37 380 114 109 104 100 95 85 75 66 56 47 37 390 115 110 105 100 95 86 76 66 56 47 37 400 116 111 106 101 96 86 76 67 57 47 38 410 117 112 107 102 97 87 77 67 57 48 38 420 117 112 107 102 97 88 77 68 58 48 38 430 118 113 108 103 98 88 78 68 58 48 38 Table continues on next page.

PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 7 of 8 35 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Initial Spent Fuel Pool Temperature (°F)

Time (hours) to reach 200°F during Modes 1, 2, 3, 4, or 5 Time After Shutdown (days) 80 85 90 95 100 110 120 130 140 150 160 440 119 114 109 104 99 89 78 68 58 48 39 450 120 115 109 104 99 89 79 69 59 49 39 460 120 115 110 105 100 90 79 69 59 49 39 470 121 116 111 106 100 90 80 70 59 49 39 480 122 117 111 106 101 91 80 70 60 50 39 490 122 117 112 107 102 91 81 70 60 50 40 500 123 118 113 107 102 92 81 71 61 50 40 510 124 119 113 108 103 92 82 71 61 50 40 520 125 119 114 109 103 93 82 72 61 51 40 530 125 120 115 109 104 93 83 72 62 51 41 540 126 120 115 110 104 94 83 72 62 51 41 550 126 121 116 110 105 94 83 73 62 52 41 560 127 122 116 111 105 95 84 73 62 52 41 570 128 122 117 111 106 95 84 74 63 52 41 580 128 123 117 112 106 96 85 74 63 52 42 590 129 123 118 112 107 96 85 74 63 53 42 600 130 124 119 113 107 97 88 75 64 53 42 PAGE NO.

CNS LOSS OF SPENT FUEL COOLING OR LEVEL - Page 8 of 8 36 of 100 AP/1/A/5500/041 Revision 18 Time for Spent Fuel Pool To Reach 200°F

3. (Continued)

Completion Times 1.3 1.3 Completion Times (continued)

Catawba Units 1 and 2 1.3-7 Amendment Nos. 298/294 EXAMPLES EXAMPLE 1.3-3 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One Function X train inoperable.

A.1 Restore Function X train to OPERABLE status.

7 days AND 10 days from discovery of failure to meet the LCO B.

One Function Y train inoperable.

B.1 Restore Function Y train to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND 10 days from discovery of failure to meet the LCO C.

One Function X train inoperable.

AND One Function Y train inoperable.

C.1 Restore Function X train to OPERABLE status.

OR C.2 Restore Function Y train to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 72 hours

Prepared for Duke Energy by: Operations Support Services, Inc. - www.ossi-net.com (2/18/16)

Modes:

1 Power Operation Defueled DEF 2

Startup 5

Cold Shutdown 3

Hot Standby 4

Hot Shutdown Catawba Nuclear Station CSD-EP-CNS-0101-02 Rev 001 Irradiated Fuel Event RU1.1 Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x SLC/TS limits for 60 min. (Notes 1, 2)

RU1.2 Reading on any Table R-1 effluent radiation monitor

> column UE for 60 min. (Notes 1, 2, 3)

RA1.1 Dose assessment using actual meteorology indicates doses

> 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

RS1.1 Reading on any Table R-1 effluent radiation monitor > column SAE for 15 min. (Notes 1, 2, 3, 4)

RS1.2 Dose assessment using actual meteorology indicates doses

> 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

RA1.2 Reading on any Table R-1 effluent radiation monitor

> column ALERT for 15 min. (Notes 1, 2, 3, 4)

RA1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

RU2.1 UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:

- 1EMF15 (2EMF4) Spent Fuel Building Refueling Bridge

- 1EMF17 (2EMF2) Reactor Building Refueling Bridge RA2.2 Damage to irradiated fuel resulting in a release of radioactivity AND A Trip 2 radiation alarm on any of the following radiation monitor indications:

- 1EMF15 (2EMF4) Spent Fuel Building Refueling Bridge

- 1EMF17 (2EMF2) Reactor Building Refueling Bridge

- 1EMF42 (2EMF42) Spent Fuel Pool Ventilation

- 1EMF39L (2EMF39L) Containment Noble Gas RA3.1 Dose rates > 15 mR/hr in EITHER of the following areas:

Control Room (EMF12)

OR Central Alarm Station (by survey)

Release of gaseous or liquid radioactivity greater than 2 times the SLC/TS limits for 60 minutes or longer Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE Unplanned loss of water level above irradiated fuel Significant lowering of water level above, or damage to, irradiated fuel RA2.1 Uncovery of irradiated fuel in the REFUELING PATHWAY Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown RS1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

- Closed window dose rates > 100 mR/hr expected to continue for 60 min.

- Analyses of field survey samples indicate thyroid CDE

> 500 mrem for 60 min. of inhalation.

(Notes 1, 2)

Abnorm.

Rad Levels

/ Rad Effluent R

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT 2

Rad Effluent 1

None None None None HS1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision HA1.1 HU2.1 Seismic event > OBE as indicated by OBE EXCEEDED alarm on 1AD-4, B/8 HU4.1 A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

- Report from the field (i.e., visual observation)

- Receipt of multiple (more than 1) fire alarms or indications

- Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)

Seismic event greater than OBE level HU4.2 Receipt of a single fire alarm (i.e., no other indications of a FIRE)

AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)

HU3.2 Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode FIRE potentially degrading the level of safety of the plant Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown Confirmed SECURITY CONDITION or threat HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes HOSTILE ACTION within the PROTECTED AREA HU7.1 Other conditions existing that in the judgment of the Site Emergency Coordinator warrant declaration of a UE Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress 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.

An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility HA6.1 Control Room evacuation resulting in transfer of plant control to alternate locations HS6.1 An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1):

- Reactivity (Modes 1, 2 and 3 only)

- Core Cooling

- NCS heat removal Inability to control a key safety function from outside the Control Room Other conditions exist which, in the judgment of the Emergency Coordinator, indicate that events are in progress 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.

HA7.1 Other conditions exist that in the judgment of the Site Emergency Coordinator warrant declaration of an Alert HS7.1 Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress 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 result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

Other conditions existing that in the judgment of the Site Emergency Coordinator warrant declaration of a Site Area Emergency HG7.1 Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress 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 Other conditions exist which in the judgment of the Site Emergency Coordinator warrant declaration of a General Emergency None Hazards H

Seismic Event Fire Control Room Evacuation Hazardous Gases Security EC Judgment 2

4 5

1 6

7 None HU3.3 Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Table H-1 Fire Areas

- Reactor Building (Containment)

- Auxiliary Building

- Diesel Generator Rooms

- RN Pump House

- Dog Houses

- Standby Shutdown Facility (SSF)

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT None None None ISFSI EU1.1 Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any Table E-1 dose limit Damage to a loaded cask CONFINEMENT BOUNDARY E

Area Rad Levels 3

RA1.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

- Closed window dose rates > 10 mR/hr expected to continue for 60 min.

- Analyses of field survey samples indicate thyroid CDE

> 50 mrem for 60 min. of inhalation.

(Notes 1, 2)

RG1.1 Reading on any Table R-1 effluent radiation monitor > column GE for 15 min. (Notes 1, 2, 3, 4)

RG1.2 Dose assessment using actual meteorology indicates doses

> 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

RG1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

- Closed window dose rates > 1000 mR/hr expected to continue for 60 min.

- Analyses of field survey samples indicate thyroid CDE

> 5000 mrem for 60 min. of inhalation.

(Notes 1, 2)

RA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 rooms or areas (Note 5)

HU3.1 A tornado strike within the PROTECTED AREA Hazardous event HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Natural or Tech.

Hazard 3

HU4.3 A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

HU4.4 A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish None None None None None None None Table F-1 Fission Product Barrier Threshold Matrix Containment (CMT) Barrier Fuel Clad (FC) Barrier Reactor Coolant System (NCS) Barrier Loss Potential Loss Loss Loss Potential Loss Potential Loss A. NCS or SG Tube Leakage B. Inadequate Heat Removal C. CMT Radiation /

NCS Activity D. CMT Integrity or Bypass None None None None None

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier

1. An automatic or manual ECCS (SI) actuation required by EITHER:

UNISOLABLE NCS leakage

SG tube RUPTURE

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the NCS barrier

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the NCS barrier

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier None None None E. EC Judgment

1. CSFST Integrity-RED PATH conditions met

1. A leaking or RUPTURED SG is FAULTED outside of containment

1. CSFST Core Cooling-RED PATH conditions met

1. CSFST Core Cooling-ORANGE PATH conditions met

2. CSFST Heat Sink-RED PATH conditions met AND Heat sink is required None

1. CSFST Heat Sink-RED PATH conditions met AND Heat sink is required

1. CSFST Core Cooling-RED PATH conditions met AND Restoration procedures not effective within 15 min. (Note 1)

1. EMF53A/B > Table F-2 column FC Loss

2. Dose equivalent I-131 coolant activity > 300 µCi/gm

1. EMF53A/B > Table F-2 column NCS Loss

1. EMF53A/B > Table F-2 column CMT Potential Loss None None None

1. Containment isolation is required AND EITHER Containment integrity has been lost based on Emergency Coordinator judgment UNISOLABLE pathway from Containment to the environment exists

2. Indications of NCS leakage outside of containment

1. CSFST Containment-RED PATH conditions met

2. Containment hydrogen concentration > 6%

3. Containment pressure > 3 psig with < one full train of containment cooling operating per design for > 15 min. (Notes 1, 10)

System Malfunct.

SA1.1 AC power capability, Table S-1, to essential 4160V buses 1(2)ETA and 1(2)ETB reduced to a single power source for 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS SG1.1 None None Fission Product Barriers FS1.1 Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1)

Loss or potential loss of any two barriers (Table F-1)

FA1.1 Any loss or any potential loss of either Fuel Clad or NCS (Table F-1)

FG1.1 1

2 3

4 1

2 3

4 1

2 3

4 None SS1.1 Loss of Essential AC Power Loss of all offsite AC power capability to essential buses for 15 minutes or longer SU1.1 Loss of all offsite AC power capability, Table S-1, to essential 4160V buses 1(2)ETA and 1(2)ETB for 15 min.

(Note 1)

Loss of all but one AC power source to essential buses for 15 minutes or longer Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer Prolonged loss of all offsite and all onsite AC power to essential buses Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses EDA, EDC, EDB and EDD for 15 min. (Note 1)

SS2.1 Loss of all vital DC power for 15 minutes or longer SA6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5%

AND Manual trip actions taken at the reactor control console (manual reactor trip switches or turbine manual trip) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8)

An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5%

AND All actions to shut down the reactor are not successful as indicated by reactor power > 5%

AND EITHER:

Core Cooling RED PATH conditions met Heat Sink RED PATH conditions met SS6.1 Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor Inability to shut down the reactor causing a challenge to core cooling or NCS heat removal Loss of all onsite or offsite communications capabilities SU7.1 Loss of CR Indications UNPLANNED loss of Control Room indications for 15 minutes or longer SU3.1 An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

SA3.1 An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

AND Any significant transient is in progress, Table S-3 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode SA9.1 NCS activity greater than Technical Specification allowable limits SU4.1 NCS activity > Technical Specification 3.4.16 limits or Facility Operating License limits (151/159), whichever is more restrictive NCS leakage for 15 minutes or longer SU5.1 NCS unidentified or pressure boundary leakage

> 10 gpm for 15 min.

OR NCS identified leakage > 25 gpm for 15 min.

OR Leakage from the NCS to a location outside containment

> 25 gpm for 15 min.

(Note 1)

Automatic or manual trip fails to shut down the reactor SU6.1 None Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1(2)ETB for 15 min.

(Note 1)

Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1(2)ETB AND SSF fails to supply NC pump seal injection OR CA supply to SGs AND EITHER:

- Restoration of at least one essential bus in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely (Note 1)

- Core Cooling RED PATH conditions met F

S 1

3 9

Loss of Comm 7

An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is successful in shutting down the reactor as indicated by reactor power < 5%

(Note 8)

None Table S-4 Communication Methods Onsite ORO System PA System Radios (on-site)

Private Branch Exchange (PBX)

Cellular Phones Satellite Phones Business Line DEMNET X

X X

X NRC X

X X

X None None None Loss of Vital DC Power 2

MODES 1, 2, 3 & 4 SG1.2 Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1(2)ETB for 15 min.

AND Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses EDA, EDC, EDB and EDD for 15 min.

(Note 1)

None NCS Activity 4

RPS Failure 6

NCS Leakage 5

None None None SU6.2 A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8)

Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 ORO communication methods OR Loss of all Table S-4 NRC communication methods Hazardous Event Affecting Safety Systems Table S-2 Safety System Parameters

- Reactor power

- NCS level

- NCS pressure

- In-core T/C temperature

- Level in at least one S/G

- Auxiliary or emergency feed flow in at least one S/G None RA2.3 Lowering of spent fuel pool level to 24.5 ft. (Level 2) on 1(2)KFP5780 or 1(2)NVP8790 RS2.1 Lowering of spent fuel pool level to 15.5 ft. on 1(2)KFP5780 or 1(2)NVP8790 RG2.1 Spent fuel pool level at the top of the fuel racks Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer Spent fuel pool level cannot be restored to at least 15.5 ft.

on 1(2)KFP5780 or 1(2)NVP8790 for 60 min.

(Note 1) 5 6

1 2

3 4

DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 5

6 1

2 3

4 DEF 6

Refuel 1

2 3

4 1

1 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 1

1 2

3 4

Table S-1 AC Power Sources Offsite

- ATC (Train A)

- SATA (Train A) (if already aligned)

- ATD (Train B)

- SATB (Train B) (if already aligned)

Onsite

- D/G A (Train A)

- D/G B (Train B)

None Failure to isolate containment or loss of containment pressure control SU8.1 EITHER:

Any penetration is not isolated within 15 min. of a VALID containment isolation signal (Note 1)

OR Containment pressure > 3 psig with < one full train of containment cooling operating per design for > 15 min.

(Notes 1, 10) 1 2

3 4

1 2

3 4

Table S-5 Hazardous Events Seismic event (earthquake)

Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager CMT Failure 8

None Table F-2 Containment Radiation - R/hr (EMF53A/B)

NCS Loss Time After S/D 0 - 1 1 - 2 2 - 8

>8 8.8 8.4 7.0 6.2 FC Loss 550 400 160 100 CMT Potential Loss 5500 4000 1600 1000 Loss of all essential AC and vital DC power sources for 15 minutes or longer 1

2 3

4 Unit Vent Noble Gas Low Unit Vent Noble Gas High Liquid Waste Effluent Line Monitor Tank Discharge Gaseous Liquid 2.21E+4 cpm 1/2EMF36L 1/2EMF36H 0EMF49L 0EMF57L Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 2.22E+3 cpm 4.18E+6 cpm 2.42E+2 cpm 5.75E+3 cpm 4.50E+6 cpm 4.97E+5 cpm 5

6 1

2 3

4 DEF

[Refer to EAL CA6.1 OR SA9.1 for escalation due to seismic event]

[Refer to EAL CA6.1 OR SA9.1 for escalation due to natural or technological hazard]

[Refer to EAL CA6.1 OR SA9.1 for escalation due to FIRE]

Sequential number within subcategory/classification Subcategory number (1 if no subcategory)

XXX.X Category (R, H, E, C, S, F)

Emergency classification (G, S, A, U)

EAL Identifier Bldg. Elevation Mode Auxiliary 577' 4

4 4

Unit 1 Room/Area Unit 2 Room/Area Rm 478 (1EMXA)

Rm 496 (1ETA)

Rm 496 (1EMXS)

AB-577', JJ-57 (1MXK)

Rm 330 (1EMXJ)

Rm 372 (1ETB)

Rm 372 (1EMXD)

Auxiliary 560' Table H-2 Safe Operation & Shutdown Rooms/Areas Rm 469 (2EMXA)

Rm 486 (2ETA)

Rm 486 (2EMXS)

AB-577', JJ-57 (2MXK)

Rm 320 (2EMXJ)

Rm 362 (2ETB)

Rm 362 (2EMXD)

Bldg. Elevation Mode Auxiliary 577' 4

4 4

Unit 1 Room/Area Unit 2 Room/Area Rm 478 (1EMXA)

Rm 496 (1ETA)

Rm 496 (1EMXS)

AB-577', JJ-57 (1MXK)

Rm 330 (1EMXJ)

Rm 372 (1ETB)

Rm 372 (1EMXD)

Auxiliary 560' Table R-2 Safe Operation & Shutdown Rooms/Areas Rm 469 (2EMXA)

Rm 486 (2ETA)

Rm 486 (2EMXS)

AB-577', JJ-57 (2MXK)

Rm 320 (2EMXJ)

Rm 362 (2ETB)

Rm 362 (2EMXD) 4 5

6 1

2 3

4 HA1.2 A validated notification from NRC of an aircraft attack threat within 30 min. of the site HU1.2 Notification of a credible security threat directed at the site HU1.3 A validated notification from the NRC providing information of an aircraft threat Table S-3 Significant Transients

- Reactor trip

- Runback > 25% thermal power

- Electrical load rejection > 25% electrical load

- Safety injection actuation None Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Note 9: In the absence of reliable NCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data when in Mode 5 and 6.

Note 10: If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes.

Note 11: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.

Note 12: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

NOTES Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.