Text

NA-2022-6 SRO as Given Written
	ML22214A027
Person / Time
Site:	North Anna
Issue date:	08/02/2022
From:	; NRC/RGN-II
To:	;
References
	Download: ML22214A027 (162)
	v • d • e

Form 4.2-4 Senior Reactor Operator Written Examination Cover Sheet U.S. Nuclear Regulatory Commission Senior Reactor Operator Written Examination Applicant Information Name:

Date:

Facility/Unit Region:

I II III IV Reactor Vendor/Type:

Start Time:

Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80 percent overall, with 70 percent or better on the senior reactor operator (SRO)-only items if given in conjunction with the reactor operator (RO) exam; SRO-only exams given alone require a final grade of 80 percent to pass. You have 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months to complete the combined examination and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months if you are only taking the SRO-only portion.

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

Applicants Signature Results RO/SRO-Only/Total Examination Points ______ / ______ / ______ Points Applicants Points

______ / ______ / ______ Points Applicants Grade

______ / ______ / ______ Percent ML22214A027

2022 NAPS ILT NRC EXAM SRO ANSWER SHEET NAME:_________________________________

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2022 NAPS ILT NRC EXAM SRO ANSWER SHEET NAME:_________________________________

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Page 1 of 100 Given Conditions:

U1 tripped from 20% following a Refueling Outage

RCS Tave is below program

All S/G Narrow Range Levels are above program

The crew is currently at step 3 (Check Feedwater Status) of 1-ES-0.1 (Reactor Trip Response)

The BOP observes the indications below:

To control the excessive cooldown and S/G Level, the BOP will have to manually shut ___(1)___

by depressing the ___(2)___ pushbutton for each Valve's Controller.

A.

(1) both the MFW Regulating Valves and MFW Regulating Bypass Valves (2) down B.

(1) both the MFW Regulating Valves and MFW Regulating Bypass Valves (2) up C.

(1) the MFW Regulating Bypass Valves only (2) down D.

(1) the MFW Regulating Bypass Valves only (2) up Question:

(1 point) 1 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION

Given Conditions:

U1 has experienced a Small Break LOCA

The crew is performing 1-FR-C.2 (Response to Degraded Core Cooling) and identifies the following parameters:

o RCS Pressure is 2100 psig o

All RCP's are stopped o

Normal SI Header Flow is 0 gpm o

Alternate SI Header Flow is 0 gpm Which one of the following actions is required as the mitigating strategy per 1-FR-C.2?

A.

Perform a max rate cooldown.

B.

Reduce RCS Pressure by opening one PORV.

C.

Start a RCP.

D.

Establish HHSI flow.

Question:

(1 point) 2 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 2 of 100

Given Conditions:

U1 experienced a Large Break LOCA

1-ES-1.3 (Transfer to Cold Leg Recirculation) is the procedure in effect

The BOP observes the following RWST Level Indications:

1-QS-LI-100A-16%

1-QS-LI-100B-17%

1-QS-LI-100C-17%

1-QS-LI-100D-17%

Based on the above, which of the choices below completes the following statements?

An automatic swap over to the Containment Sump ___(1)___ be occurring.

In accordance with 1-ES-1.3, a manual swap over is required at ___(2)___ RWST Level.

A.

(1) should not (2) less than 15%

B.

(1) should not (2) 16%

C.

(1) should (2) less than 15%

D.

(1) should (2) 16%

Question:

(1 point) 3 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 3 of 100

Which of the choices below completes the following statements?

The RCP Flywheels cause the RCPs to coast down ___(1)___ to ensure Actual Heat Flux is maintained less than Critical Heat Flux during a Loss of Offsite Power event.

Without the RCP Flywheels, the RCS would likely exceed ___(2)___.

A.

(1) slower (2) DNBR B.

(1) slower (2) QPTR C.

(1) faster (2) DNBR D.

(1) faster (2) QPTR Question:

(1 point) 4 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 4 of 100

Given Conditions:

U1 is at 100% power

The RO observes the indications below:

Based on the indications above, which of the choices below completes the following statements?

___(1)___ is failing high.

In accordance with 1-AR-C-C5, the RO should ___(2)___ to control Pressurizer Level.

A.

(1) 1-CH-FCV-1122 Controller (2) isolate Letdown, then isolate Charging by closing 1-CH-MOV-1289A B.

(1) 1-CH-FCV-1122 Controller (2) place 1-CH-FCV-1122 Controller in MAN and lower demand C.

(1) 1-CH-FT-1122 (2) isolate Letdown, then isolate Charging by closing 1-CH-MOV-1289A D.

(1) 1-CH-FT-1122 (2) place 1-CH-FCV-1122 Controller in MAN and lower demand 1C-C5 Question:

(1 point) 5 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 5 of 100

Initial Conditions:

U1 is in a Refueling Outage

Fuel movement is in progress Current Conditions:

The crew has entered 1-AP-11 (Loss of RHR) due to no running RHR Pumps

The crew is currently performing Attachment 11 (Containment Closure) o While placing the Containment Purge Exhaust Fans through the Iodine Filters, ONLY one of the two switches for 1-HV-AOD-104-1,2,3,4 dampers could be placed to the FILTER position o

The designated OPs Personnel Hatch Operator can ONLY close the Inner Personnel Hatch door o

All other actions of 1-AP-11 Attachment 11 are complete Which of the choices below completes the following statements regarding Attachment 11 of 1-AP-11?

The Containment Purge Exhaust Fans ___(1)___ re-align to the Iodine Filters.

Based on the Personnel Hatch status, Containment Closure ___(2)___ met.

A.

(1) will (2) is not B.

(1) will not (2) is not C.

(1) will (2) is D.

(1) will not (2) is Question:

(1 point) 6 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 6 of 100

Initial Conditions:

A Loss of Component Cooling occurred

Per 0-AP-27 (Malfunction of the Spent Fuel Pit System), Service Water was aligned to the Spent Fuel Pit Coolers Current Conditions:

Component Cooling has been returned to service

Component Cooling Water through the SFP Heat Exchanger has been restored using (CC Restoration to SFP Cooling Heat Exchangers) of 0-AP-27 Based on the above and in accordance with Attachment 2, which of the choices below completes the following statement?

Fuel Pit Temperature will be controlled by throttling the Component Cooling ___(1)___ Valves to the SFP Heat Exchangers.

Throttling will occur locally in the ___(2)___.

A.

(1) Inlet (2) Demin Alley B.

(1) Inlet (2) Fuel Building Basement C.

(1) Outlet (2) Demin Alley D.

(1) Outlet (2) Fuel Building Basement Question:

(1 point) 7 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 7 of 100

Initial Conditions:

U1 is in Mode 1

1-RC-PCV-1456 (Pressurizer PORV) fails open

RCS Pressure is 2200 psig and lowering Current Conditions:

The crew has entered 1-AP-44 (Loss of Reactor Coolant System Pressure)

The crew has placed the control switch for 1-RC-PCV-1456 to CLOSE o

1-RC-PCV-1456 remains OPEN Based on the above and in accordance with 1-AP-44, which of the choices below completes the following statements?

The next required action is to ___(1)___ for 1-RC-PCV-1456.

If unable to perform this action, the Unit ___(2)___ remain in Mode 1.

A.

(1) close the Block Valve (2) can B.

(1) close the Block Valve (2) cannot C.

(1) place the Nitrogen and PRZR PWR RV Appendix R Isolation Switches to DISABLE (2) can D.

(1) place the Nitrogen and PRZR PWR RV Appendix R Isolation Switches to DISABLE (2) cannot Question:

(1 point) 8 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 8 of 100

Initial Conditions:

U2 entered 2-FR-S.1 (Response to Nuclear Power Generation/ATWS) from 2-E-0 (Reactor Trip or Safety Injection)

Controls Rods are inserting in Auto at max speed

Neutron flux is lowering Current Conditions

The crew is currently at step 5 (Initiate Emergency Boration of RCS)

2-CH-MOV-2350 (U2 Emergency Borate Valve) failed to open In accordance with 2-FR-S.1, the negative reactivity insertion is ___(1)___ and the crew will ___(2)___.

A.

(1) adequate (2) transition back to 2-E-0 B.

(1) adequate (2) continue in 2-FR-S.1 C.

(1) inadequate (2) manually initiate Safety Injection D.

(1) inadequate (2) align the BIT for injection Question:

(1 point) 9 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 9 of 100

Given Conditions:

U1 is at 100% power

The crew has entered 1-AP-2.2 (Fast Load Reduction) and will be ramping the unit offline Based on the above, which of the choices below completes the following statement?

In accordance with 1-AR-K-G6 (N-16 RAD DET), the N-16 Detectors are declared non-functional at ______% Reactor Power.

A.

35 B.

30 C.

25 D.

20 Question:

(1 point) 10 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 10 of 100

Given Conditions:

The crew just entered 1-ECA-2.1 (Uncontrolled Depressurization of All Steam Generators)

Containment Pressure is 50 psia and raising rapidly

All S/G Pressures are 200 psig and lowering rapidly

All S/G Narrow Range Levels are off scale low Which of the choices below completes the following statements?

In accordance with 1-ECA-2.1, the BOP will throttle Auxiliary Feedwater Flow to ___(1)___ in order to ___(2)___.

A.

(1) 100 gpm per S/G (2) maintain the minimum heat sink requirements B.

(1) 100 gpm per S/G (2) prevent a S/G from drying out C.

(1) greater than or equal to 340 gpm total (2) maintain the minimum heat sink requirements D.

(1) greater than or equal to 340 gpm total (2) prevent a S/G from drying out Question:

(1 point) 11 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 11 of 100

Given Conditions:

Crew initiated RCS Bleed and Feed in 1-FR-H.1 (Response to Loss of Secondary Heat Sink)

All S/G WR Levels are <12%

Hot Leg Temperatures are approximately 560°F and stable

The Operators restore a Feedwater source

SRO directs the BOP to feed only one S/G rather than all three S/Gs The reason for feeding only one S/G under these conditions is to A.

prevent a Pressurized Thermal Shock condition B.

demonstrate the reliability of the Feedwater source C.

limit a potential fault or rupture to one S/G D.

determine if one S/G is capable of maintaining a heat sink Question:

(1 point) 12 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 12 of 100

Initial Conditions:

U1 is at 75% power following a controlled ramp

Rods are in Manual

Tave is 568°F and stable

Tref is 571°F and stable Current Conditions:

AC Vital Bus 1-II de-energizes

The crew has completed 1-AP-3 (Loss of Vital Instrumentation)

1-AP-4.3 (Loss of Power Range NIs) is the procedure in effect o

SRO desires Tave/Tref to be matched with Rods Based on the above, which of the choices below completes the following statement?

In accordance with 1-AP-4.3, Rods currently can't be withdrawn until the ___(1)___ Protection Interlock is cleared for ___(2)___ Power Range NI.

A.

(1) C-2 (2) N-44 B.

(1) C-2 (2) N-42 C.

(1) C-4 (2) N-44 D.

(1) C-4 (2) N-42 Question:

(1 point) 13 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 13 of 100

Initial Conditions:

U1 is at 100% power Current Conditions:

The BOP observes the indications below:

Which of the choices below completes the following statements?

The Unit has experienced a loss of 125VDC Vital Bus (1)

Due to the loss of power, the Main Steam Trip Valves are (2)

A.

1) 1-I

2) open B.

1) 1-I

2) closed C.

1) 1-III

2) open D.

1) 1-III

2) closed Question:

(1 point) 14 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 14 of 100

Initial Conditions:

Both Units are at 100% power

Both U1 Service Water Pumps are running

An Earthquake occurred Current Conditions:

The following MCR indications are observed:

o All MCR Aux. Building Sump Indicators are off-scale high o

Both U1 SW Pump Amps are off-scale high o

Both U1 SW Pump Discharge Pressures are off-scale low

The crew entered 0-AP-12 (Loss of Service Water)

In accordance with 0-AP-12, the crew will trip ___(1)___ and align ___(2)___ to the U1 Main Control Room Chillers.

A.

(1) U1 only (2) Bearing Cooling B.

(1) U1 only (2) Fire Protection C.

(1) both Units (2) Bearing Cooling D.

(1) both Units (2) Fire Protection Question:

(1 point) 15 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 15 of 100

Given Conditions:

Chemistry is sampling 1-SI-TK-1A ("A" Safety Injection Accumulator)

CH-11.401 (Remote Sampling of 1-SI-TK-1A From Safeguards) is the procedure in effect

The following valves have been OPENED from the MCR o

1-SI-TV-1842 (SI Accumulator Test Line Inside Isolation Valve) o 1-SI-TV-1859 (SI Accumulator Test Line Outside Isolation Valve) o 1-SI-HCV-1850A ("A" SI Accumulator Sample Isolation Valve)

The Chemistry Technician is purging the Sample Line Which of the following choices below completes the following statements?

On a loss of Instrument Air, 1-SI-TK-1A Level will ___(1)___.

T.S. 3.5.1 (Accumulators) requires 1-SI-TK-1A Boron Concentration to be ___(2)___ ppm.

A.

(1) continue to lower (2) 2600 to 2800 B.

(1) stabilize (2) 2600 to 2800 C.

(1) continue to lower (2) 2500 to 2800 D.

(1) stabilize (2) 2500 to 2800 Question:

(1 point) 16 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 16 of 100

1-ECA-1.2 (LOCA Outside Containment) directs actions to verify valve positions for only ONE system. This system is the most likely location for a LOCA Outside Containment.

Which system is addressed by 1-ECA-1.2?

A.

CVCS Charging.

B.

CVCS Letdown.

C.

Low Head Safety Injection.

D.

High Head Safety Injection.

Question:

(1 point) 17 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 17 of 100

Why is monitoring for Bleed and Feed criteria a continuous action step in 1-FR-H.1 (Response to Loss of Secondary Heat Sink)?

A.

To maintain S/G Integrity.

B.

To ensure adequate ECCS flow to remove decay heat.

C.

To extend the time allowed to restore Feedwater.

D.

To minimize the heat input into Containment while on Bleed and Feed.

Question:

(1 point) 18 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 18 of 100

Initial Conditions:

U1 is at 95% power

All Rods Out position is 228 steps

"D" Bank Control Rods are at 217 steps

Control Rods are in AUTO Current Conditions:

1-RC-TI-1408A (Median Hi Tave) fails low to 530°F

1-AR-B-A7 (MEDIAN/HI TAVG < > TREF DEVIATION) is lit

The crew has entered 1-AP-1.1 (Continuous Uncontrolled Rod Motion) o Rods have been placed in MANUAL o

"D" Bank Control Rods are at 225 steps and still stepping out Based on the above, which of the choices below completes the following statements?

In accordance with 1-AP-1.1, the crew is required to ___(1)___

The Reactor is tripped due to a ___(2)___.

A.

(1) trip the Reactor and enter 1-E-0 immediately.

(2) loss of reactivity control B.

(1) trip the Reactor and enter 1-E-0 immediately.

(2) RCS Loop Average Temperature < 541°F C.

(1) verify Rods stop at C-11 ("D" Bank at All Rods Out). If not, trip the Reactor and enter 1-E-0.

(2) loss of reactivity control D.

(1) verify Rods stop at C-11 ("D" Bank at All Rods Out). If not, trip the Reactor and enter 1-E-0.

(2) RCS Loop Average Temperature < 541°F Question:

(1 point) 19 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 19 of 100

Given Conditions:

U1 is in Mode 2

A Reactor startup is in progress following a Refueling Outage As power is increased, maintaining the limits of T.S. 3.2.4 (Quadrant Power Tilt Ratio) applies in Mode 1 when rated thermal power is ___(1)___ than 50% and ensures that the gross ___(2)___

power distribution remains within design values used in the safety analysis.

A.

(1) less (2) axial B.

(1) less (2) radial C.

(1) greater (2) axial D.

(1) greater (2) radial Question:

(1 point) 20 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 20 of 100

Initial Conditions:

U1 is at 100% power

Normal Letdown has been removed from service to perform work on 1-CH-TV-1204B (Letdown Isolation Valve)

Excess Letdown is in service Current Conditions:

1-CC-RM-120 (CCHX Radiation Monitor) Hi and Hi-Hi alarms are lit

CC Head Tank Level is trending up Which of the choices below completes the following statements?

The leak is occurring in the ___(1)___ Heat Exchanger.

In accordance 1-AP-49.1 (Loss of Normal and Excess Letdown), the sum of Charging and RCP Seal Injection Flow must be maintained ___(2)___ than Emergency Boration Flow.

A.

(1) Excess Letdown (2) greater B.

(1) Excess Letdown (2) less C.

(1) Seal Water (2) greater D.

(1) Seal Water (2) less Question:

(1 point) 21 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 21 of 100

Initial Conditions:

U1 is at 100% power

Pressurizer Level Channel Defeat Switch is selected to 459/460

1-RC-LT-1461 (Channel III Pressurizer Level Transmitter) failed low and all associated bistables have been placed in trip by I&C Current Conditions

The Reference Leg for 1-RC-LT-1460 (Channel II Pressurizer Level Transmitter) ruptures Which one of the following describes the plant response? (Assume no Operator actions)

A.

Letdown isolates.

B.

Charging Flow reduces to minimum.

C.

Reactor Trip signal is initiated.

D.

Backup Heaters energize.

Question:

(1 point) 22 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 22 of 100

Initial Conditions:

U1 is in a Refueling Outage

Fuel Movement is in progress

All 125VAC Vital Busses are energized

Both Source Range Instruments are Operable Current Conditions:

125VAC Vital Bus 1-I is de-energized Based on the above, which of the choices below completes the following statements?

Per T.S. 3.9.3 (Nuclear Instrumentation), fuel movement ___(1)___ continue.

Per 1-AP-4.1 (Malfunction of Nuclear Instrumentation-Source Range), Channel Selector Switch on the Audible Count Rate Drawer must be selected to the Operable Source Range Channel ___(2)___ to ensure Source Range Counts are heard in Containment.

A.

(1) can (2) N-31 B.

(1) can (2) N-32 C.

(1) cannot (2) N-31 D.

(1) cannot (2) N-32 Question:

(1 point) 23 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 23 of 100

Initial Conditions:

U1 is at 100% power

A S/G Tube Leak has developed on the "B" S/G

1-SV-RM-121 (Condenser Air Ejector Radiation Monitor) High alarm is lit

The crew has entered 1-AP-24 (Steam Generator Tube Leak)

Current Conditions

While performing 1-AP-24, the Tube Leak worsens as noted by the "B" S/G Main Feedwater Flow parameters Which of the choices below completes the following statements?

The BOP will observe a ___(1)___ trend of the "B" S/G Main Feedwater Flow.

The crew will ___(2)___.

A.

(1) lowering (2) remain in 1-AP-24 only B.

(1) lowering (2) perform 1-AP-24 and 1-E-0 (Reactor Trip or Safety Injection) concurrently C.

(1) rising (2) remain in 1-AP-24 only D.

(1) rising (2) perform 1-AP-24 and 1-E-0 (Reactor Trip or Safety Injection) concurrently Question:

(1 point) 24 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 24 of 100

Which of the choices below completes the following statements?

1-FR-C.1 (Response to Inadequate Core Cooling) is required to be entered when CETCs FIRST exceed ___(1)___.

Normal support conditions ___(2)___ required to start the RCPs in step 21 (Check if RCPs Should Be Started) of 1-FR-C.1.

A.

(1) 700°F with 50% RVLIS Full Range (2) are B.

(1) 1200°F (2) are C.

(1) 700°F with 50% RVLIS Full Range (2) are not D.

(1) 1200°F (2) are not Question:

(1 point) 25 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 25 of 100

In a comparison between a delayed neutron and prompt neutron produced from the same fission event, the prompt neutron is more likely to

. (Assume that both neutrons remain in the North Anna Core)

A.

cause fast fission of a U-238 Nucleus B.

be captured by a U-238 Nucleus at a resonance energy between 1 eV and 1000 eV C.

be captured by a Xe-135 Nucleus D.

cause thermal fission of a U-235 Nucleus Question:

(1 point) 26 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 26 of 100

What is the mitigation strategy for reducing radiation levels in Containment in accordance with 1-FR-Z.3 (Response to High Containment Radiation)?

A.

Initiate the Quench Spray System.

B.

Initiate the Recirculation Spray System.

C.

Place the Containment Iodine Filtration Fans in service.

D.

Place the Containment Purge Exhaust Fans through the Charcoal Filters.

Question:

(1 point) 27 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 27 of 100

Based on the above indications, Control Power to the "A" RCP Bearing Lift Pump ___(1)___

available and Bearing Oil Lift Pressure ___(2)___ adequate.

A.

1) is

2) is B.

1) is

2) is not C.

1) is not

2) is D.

1) is not

2) is not Question:

(1 point) 28 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 28 of 100

With LTOPS in service, LCO 3.4.12 (Low Temperature Overpressure Protection System) states that two Charging Pumps can be made capable of injecting for less than or equal to ___(1)___ minutes for

___(2)___.

A.

(1) 60 (2) pump swapping B.

(1) 60 (2) surveillance testing C.

(1) 15 (2) pump swapping D.

(1) 15 (2) surveillance testing Question:

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If 1-RH-MOV-1720A (Residual Heat Removal Outlet Isolation Valve to B Cold Leg) is energized and opened at power, level will INITIALLY rise in the ___(1)___.

A.

"B" Safety Injection Accumulator B.

Pressurizer Relief Tank C.

Primary Drains Transfer Tank D.

Containment Sump Question:

(1 point) 30 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 30 of 100

Initial Conditions:

U1 is at 100% power Current Conditions:

U1 has been tripped due to a Small Break LOCA inside Containment

1H Bus is de-energized

The crew is performing 1-ES-1.2 (Post-LOCA Cooldown and Depressurization)

Safety Injection Accumulator Isolation is required Which of the choices below completes the following statements?

At 100% power, 1-SI-MOV-1865A/B/C (Safety Injection Accumulators Isolation MOVs) are (1)

In accordance with 1-ES-1.2, the crew can isolate (2)

A.

(1) energized (2) 1-SI-MOV-1865B and 1-SI-MOV-1865C B.

(1) energized (2) 1-SI-MOV-1865C only C.

(1) de-energized (2) 1-SI-MOV-1865B and 1-SI-MOV-1865C D.

(1) de-energized (2) 1-SI-MOV-1865C only Question:

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Given Conditions:

U1 is at 100% power

A Large Break LOCA coincident with a Loss of Offsite Power occurred

Two SI Accumulators failed to inject into the Core Based on the above, the effect of the two SI Accumulators failing to inject is ______.

A.

less Nitrogen is injected to block Natural Circulation B.

Containment Sump Inventory will be insufficient for long-term Core Cooling C.

Containment Recirculation Sump pH will be lower than required D.

insufficient water will be available during the Blowdown and Refill Phases of the LOCA Question:

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Given Conditions

U1 is tripped

Safety Injection is in progress

Annunciator Status o

B-G1 (PRZ RELIEF TANK HI-LO LEVEL) is locked in o

B-H1 (PRZ RELIEF TANK HI TEMP) is locked in o

B-F1 (PRZ RELIEF TK HI PRESS) just cleared

Containment Sump Level is rising

Containment Pressure is rising

Containment Radiation is rising Assuming no Operator action, the failure of which of the following would result in these conditions?

A.

Rx Head Vent Valves Open.

B.

1-RC-PCV-1456 (Pressurizer PORV) Open.

C.

Inner Rx Vessel Flange O-Ring.

D.

CRDM Pressure Boundary Housing.

Question:

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Initial Conditions:

U1 is at 100% power

1-CC-P-1A (U1 "A" Component Cooling Water Pump) is running

2-CC-P-1A (U2 "A" Component Cooling Water Pump) is running Current Conditions:

1-EE-BKR-15J11 (4160V 1J Bus Normal Feeder Breaker) spuriously trips open

1J EDG starts and loads onto the 1J 4160V Emergency Bus Based on the conditions above, 1-CC-P-1B (U1 "B" Component Cooling Water Pump) will start automatically ______.

A.

with no time delay after the 1J EDG loads onto the Bus B.

15 seconds after the 1J EDG loads onto the Bus C.

with no time delay after the Stub Bus Breaker is reset manually D.

15 seconds after the Stub Bus Breaker is reset manually Question:

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Initial Conditions:

U1 is in Mode 5

RCS is solid

Lowest RCS Tc <180°F

RHR is returning to the "C" RCS Loop

1-RH-FCV-1605 (RHR Heat Exchanger Bypass Flow) is in Automatic Current Conditions:

1-RH-FT-1605 (RHR Heat Exchangers Outlet Header Flow Transmitter) fails low Based on the conditions above, Pressurizer Pressure will ______.

A.

lower until 1-RH-FCV-1605 is throttled open B.

rise and stabilize at approximately 530 psig C.

lower until charging flow is manually raised D.

rise and stabilize at approximately 365 psig Question:

(1 point) 35 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 35 of 100

Given Conditions:

U2 experienced a Safety Injection coincident with a Loss of Offsite Power from 100%

2-ECA-2.1 (Uncontrolled Depressurization of All Steam Generators) has been entered The crew is required to go to 2-FR-P.1 (Response to Imminent Pressurized Thermal Shock) if any Cold Leg Temperature drops below ___(1)___°F and will depressurize the RCS using ___(2)___.

A.

(1) 285 (2) a Pressurizer PORV B.

(1) 285 (2) Auxiliary Spray C.

(1) 315 (2) a Pressurizer PORV D.

(1) 315 (2) Auxiliary Spray Question:

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Initial Conditions:

U1 is stable at 70% power

ICP-RC-1-T-1412 (Delta T/Tave Protection Channel I (1-RC-T-1412) Calibration) is in progress o

The Delta T and Tave Defeat Switches have been placed in Defeat Loop "A" o

All associated bistables have been placed in TEST Current Conditions:

1-RC-TE-1422E (Loop "B" Tcold Temperature Element) is failing low Which of the choices below completes the following statements?

The RO will observe a ___(1)___ trend on Loop "B" T.

The Unit ___(2)___ automatically trip.

A.

(1) rising (2) will B.

(1) lowering (2) will C.

(1) rising (2) will not D.

(1) lowering (2) will not Question:

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Initial Conditions:

1-E-0 (Rx Trip or Safety Injection) was entered o

The Reactor did not trip Current Conditions:

The crew is currently performing 1-FR-S.1 (Response to Nuclear Power Generation/ATWS)

An Operator has NOT been dispatched to trip the Unit locally

While inserting Control Rods in MANUAL, the RO observes the following:

o 1-AR-E-D4 (AMSAC INITIATED) locked in o

All Rod Bottom Lights just illuminated Based on the conditions above, what is the status of the breakers listed below?

(1) MG Set Input Breakers (2) Reactor Trip Breakers A.

(1) open (2) closed B.

(1) open (2) open C.

(1) closed (2) closed D.

(1) closed (2) open Question:

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Given Conditions:

A Large Break LOCA is in progress on U1

RWST Level is lowering 2%/min Which of the choices below completes the following statements?

1-QS-LI-100A, 1-QS-LI-100C, and ___(1)___ input into the auto start logic for the Recirculation Spray Pumps.

Recirculation Spray Flow will be at MAX rate in ___(2)___ minutes.

A.

(1) 1-QS-LI-100B (2) 21 B.

(1) 1-QS-LI-100B (2) 19 C.

(1) 1-QS LI-100D (2) 21 D.

(1) 1-QS-LI-100D (2) 19 Question:

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Initial Conditions:

U1 is at 100% power

U1 Steam Chiller is currently in an Idled condition

Shift Order priorities for U1 are to remove the Mechanical Chiller from service and place the U1 Steam Chiller in service Current Conditions

1-CC-TV-115C (Chilled Water Return From Recirculation Air Coolers) spuriously shut and the Mechanical Chiller tripped

1-CC-TV-115C cannot be re-opened

The crew has entered 1-AP-35 (Loss of Containment Air Recirculation Cooling)

Containment Average Temperature is rising Which of the choices below completes the following statements?

LCO 3.6.5 (Containment Air Temperature) will NOT be met when Containment Average Temperature FIRST exceeds ___(1)___°F.

In accordance with 1-AP-35, the crew will ___(2)___.

A.

(1) 105 (2) swap to the U1 Steam Chiller B.

(1) 105 (2) swap to Service Water C.

(1) 115 (2) swap to the U1 Steam Chiller D.

(1) 115 (2) swap to Service Water Question:

(1 point) 40 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 40 of 100

Given Conditions:

U1 CDA is in progress Quench Spray Water Alkalinity is ensured by ______.

A.

1-QS-P-3 (Refueling Water Chemical Addition Pump)

B.

the Casing Cooling Pumps C.

the RWST Suction Weir D.

the QS Spray Ring Nozzle size Question:

(1 point) 41 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 41 of 100

Adequate Net Positive Suction Head and long term Recirculation Spray System operation is ensured when level is at a MINIMUM Sump Level of ___(1)___ on the ___(2)___ Range Sump Level Indicators.

A.

(1) 4'10" (2) Narrow B.

(1) 4'10" (2) Wide C.

(1) 11'0" (2) Narrow D.

(1) 11'0" (2) Wide Question:

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Given Conditions:

The crew has entered 0-FCA-1 (MCR Fire)

MCR has been evacuated 2 (U1 Dedicated Flow Path Alignment) is being performed In accordance with Attachment 22, the U1 S/G PORVs will be closed by ___(1)___ to EMERG CLOSE in the Unit 1 ___(2)___.

A.

(1) placing the switch handles (2) Emergency Switchgear B.

(1) installing the keys and turning the key switches (2) Emergency Switchgear C.

(1) placing the switch handles (2) Cable Vault and Tunnel D.

(1) installing the keys and turning the key switches (2) Cable Vault and Tunnel Question:

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Initial Condition:

U1 is at 70% power

All plant equipment is available

1-SD-P-1A (U1 "A" High Pressure Heater Drain Pump) tripped Which of the choices below completes the following statements?

1-SD-P-1A discharges into the suction of the Main ___(1)___ Pumps.

1-SD-P-1A tripping ___(2)___ an entry condition for 1-AP-31 (Loss of Main Feedwater).

A.

(1) Feedwater (2) is B.

(1) Feedwater (2) is not C.

(1) Condensate (2) is D.

(1) Condensate (2) is not Question:

(1 point) 44 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 44 of 100

Which of the choices below completes the following statement?

The MCC power supply for 1-FW-MOV-100A (Auxiliary Feedwater MOV Header to "A" S/G) is ______.

A.

1-EP-MCC-1J1-2N B.

1-EP-MCC-1A2-2 C.

1-EP-MCC-1A1-2 D.

1-EP-MCC-1B1-3 Question:

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Initial Conditions:

U2 is at 100% power Current Conditions

Safety Injection occurs on U2

Safeguards Operator reports that there is an unisolable COOLING Water Pipe break upstream of the Lube Oil Cooler for 2-FW-P-2 (Turbine Driven Auxiliary Feedwater Pump)

Which of the choices below completes the following statements?

With no Operator action, the Emergency Condensate Storage Tank Level will lower based off of

___(1)___.

In order to secure 2-FW-P-2, the RO will ___(2)___.

A.

(1) Auxiliary Feedwater Flow to the S/Gs only (2) close 2-MS-TV-111A and 2-MS-TV-111B (2-FW-P-2 Steam Supply Valves) only B.

(1) Auxiliary Feedwater Flow to the S/Gs AND the pipe break flow (2) close 2-MS-TV-111A and 2-MS-TV-111B (2-FW-P-2 Steam Supply Valves) only C.

(1) Auxiliary Feedwater Flow to the S/Gs only (2) reset SI, then close 2-MS-TV-111A and 2-MS-TV-111B (2-FW-P-2 Steam Supply Valves)

D.

(1) Auxiliary Feedwater Flow to the S/Gs AND the pipe break flow (2) reset SI, then close 2-MS-TV-111A and 2-MS-TV-111B (2-FW-P-2 Steam Supply Valves)

Question:

(1 point) 46 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 46 of 100

Initial Conditions:

U1 is at 25% power

The Unit is being shut down for a Refueling Outage

The Switchyard is in the normal electrical configuration

U1 RO completed transferring the U1 Station Service Busses to the Reserve Station Service Transformers IAW 1-OP-26.1 (Transferring 4160V Busses)

Current Conditions:

Switchyard Transformer #1 experiences an internal fault Which of the choices below completes the following statement?

The ___(1)___ Station Service Bus will de-energize and the Unit ___(2)___ trip on RCS Single Loop Loss of Flow.

A.

(1) "B" (2) will B.

(1) "B" (2) will not C.

(1) "C" (2) will D.

(1) "C" (2) will not Question:

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Given Conditions:

U1 is at 100% power

1-FW-P-1A ("A" Main Feedwater Pump) and 1-FW-P-1C ("C" Main Feedwater Pump) are in service

The RO discovers the closed (running light) is not lit for 1-FW-P-1A1 (Main Feedwater Pump 1A1)

Turbine Building Operator reports the same indication locally at the breaker for 1-FW-P-1A1 in U1 307' Switchgear

Both sets of light bulbs are NOT burned out Based on the conditions above, ______.

A.

the breaker for 1-FW-P-1A1 CANNOT be remotely opened with the control switch B.

a Train "A" Safety Injection WILL cause the breaker for 1-FW-P-1A1 to trip open C.

a Breaker Overload condition WILL cause the breaker for 1-FW-P-1A1 to trip open D.

a Bus Undervoltage condition WILL cause the breaker for 1-FW-P-1A1 to trip open Question:

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Initial Conditions:

U2 is at 100% power

All equipment is available

2-PT-82J (2J EDG Slow Start PT) is in progress o

EDG Load reading is 2550 KW o

EDG KVARs reading is 50KVARS out Current Conditions:

Spurious U2 Safety Injection occurs Which of the choices below completes the following statements?

During the transient, the 2J Bus Frequency will ___(1)___.

In order to open 2-EE-BKR-25J2 (2J EDG Output Breaker), the Backboards Operator will ___(2)___.

A.

(1) remain stable (2) reset Safety Injection, then lower load and open 2-EE-BKR-25J2 B.

(1) remain stable (2) lower load and open 2-EE-BKR-25J2 C.

(1) lower as equipment loads on the bus (2) reset Safety Injection, then lower load and open 2-EE-BKR-25J2 D.

(1) lower as equipment loads on the bus (2) lower load and open 2-EE-BKR-25J2 Question:

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Given Conditions:

U1 is at 100% power

1-PT-82H (1H EDG Slow Start Test) is in progress o

EDG Voltage (incoming) is 124VAC o

Bus Voltage (running) is 122VAC

Backboards Operator is at the step to close 1-EE-BKR-15H2 (1H EDG Output Breaker)

Which of the choices below completes the following statements?

After the EDG Output Breaker is closed, the KVAR indication will be ___(1)___.

In order to return KVAR indication to 0 using the Exciter Voltage Control Switch, the Backboards Operator will place the switch to the ___(2)___ position.

A.

(1) KVARs out (2) LOWER B.

(1) KVARs out (2) RAISE C.

(1) KVARs in (2) LOWER D.

(1) KVARs in (2) RAISE Question:

(1 point) 50 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 50 of 100

Given Conditions:

Both Units are at 100% power

1-SV-RM-121 (Condenser Air Ejectors Discharge Radiation Monitor) reading is 40 cpm and stable

No alarms are lit on 1-SV-RM-121 If the Instrument Power Fuses for 1-SV-RM-121 blow, the meter indication will go off-scale ___(1)___

and the Air Ejector's Discharge ___(2)___ align to Containment.

A.

(1) low (2) will B.

(1) low (2) will not C.

(1) high (2) will D.

(1) high (2) will not Question:

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Given Conditions:

U1 is at 100% power

U2 just entered Mode 4

CC System is split out Which of the choices below completes the following statements as U2 RHR is placed in service?

The temperature on the "A" (#4) Service Water Return Header will be ___(1)___.

The temperature on the "B" (#3) Service Water Return Header will be ___(2)___.

A.

(1) stable (2) stable B.

(1) stable (2) rising C.

(1) rising (2) stable D.

(1) rising (2) rising Question:

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Given Conditions:

U1 "C" S/G Faulted inside Containment

Containment Pressure is 35 psia and lowering

The BOP observes the following indications Which of the choices below completes the following statements?

The ___(1)___ Signal must be RESET in order to reposition 1-SW-MOV-108A and 1-SW-MOV-108B.

Assuming the appropriate ESFAS Signal is RESET, the valves ___(2)___ fully OPEN if the OPEN pushbuttons are depressed and released.

A.

(1) Safety Injection (SI)

(2) will B.

(1) Containment Depressurization Actuation (CDA)

(2) will C.

(1) Safety Injection (SI)

(2) will not D.

(1) Containment Depressurization Actuation (CDA)

(2) will not Question:

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Initial Conditions:

Both Units are at 100% power

1-IA-C-1 (U1 Instrument Air Compressor) and 1-SA-C-1 (U1 Service Air Compressor) are running in HAND

2-IA-D-1 (U2 Instrument Air Dryer) and 2-IA-D-7 (Instrument Air Dryer 7) are in service

Security reports loud air flow coming from the Muffler on 2-IA-D-1 Current Conditions

Instrument Air Pressure is 94 psig and lowering

2-IA-C-1 (U2 Instrument Air Compressor) and 2-SA-C-1 (U2 Service Air Compressor) are running in AUTO

Both Units entered the respective Unit's AP-28 (Loss of Instrument Air)

Which of the choices below completes the following statements?

In accordance with AP-28, ___(1)___ will be placed in HAND.

In accordance with Attachment 10 (Turbine/Auxiliary Building Corrective Actions), the Operator will

___(2)___ 2-IA-D-1.

A.

(1) all available Compressors.

(2) isolate B.

(1) all available Compressors.

(2) bypass and isolate C.

(1) both Unit's Containment Instrument Air Compressors only (2) isolate D.

(1) both Unit's Containment Instrument Air Compressors only (2) bypass and isolate Question:

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Initial Conditions:

U1 is at 100% power

1-CH-P-1B (U1 "B" Charging Pump) is running

No equipment is out of service Current Conditions:

U1 just entered 1-E-0 (Reactor Trip or Safety Injection)

RO is performing the IOAs

BOP observes the following o

1P-H2 (Safety Injection Initiated) Annunciator is flashing o

1-CH-P-1B is the only Charging Pump running Based on the above, select the correct responses regarding the Phase A Containment Isolation Valves:

(1) Expected valve position for Train A Valves (2) Expected valve position for Train B Valves A.

(1) All closed (2) Some open, some closed B.

(1) Some open, some closed (2) All closed C.

(1) All closed (2) All closed D.

(1) Some open, some closed (2) Some open, some closed Question:

(1 point) 55 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 55 of 100

Given Conditions:

U1 is at 96% power

"D" Bank IRPI indications are 225 steps EXCEPT:

o B8 is 228 steps o

K10 is 214 steps

The Bank Overlap Counter reading in the Logic Cabinet is 599 steps Which of the choices below completes the following statements?

Based on the Bank Overlap Counter, the expected "D" Bank Group Step Counter reading is ___(1)___

steps.

The LCO for T.S. 3.1.4 (Rod Group Alignment Limits) is ___(2)___.

A.

(1) 215 (2) met B.

(1) 215 (2) not met C.

(1) 225 (2) met D.

(1) 225 (2) not met Question:

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Given Conditions:

U1 is stable at 90% power

The Main Turbine is in OPEN LOOP

Rods are in Manual

The 1A Feedwater Heater is being removed from service Which of the choices below completes the following statements?

As the Feedwater Heater is being removed from service, Tave will ___(1)___.

PRNI indicated power will be ___(2)___ than actual power.

A.

(1) rise (2) greater B.

(1) rise (2) less C.

(1) lower (2) greater D.

(1) lower (2) less Question:

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Given Conditions:

A small fire occurred in the MCR

The crew is performing actions of 1-AP-20 (Operation From the Auxiliary Shutdown Panel)

The RO and SRO have manned the Auxiliary Shutdown Panel (ASDP) and are transferring control of equipment Which of the following describes parameters that are DIRECTLY available to be read within the ASDP in order to perform 1-AP-20 actions?

1.

Wide Range Steam Generator Water Level 2.

Refueling Water Storage Tank Level 3.

Auxiliary Feedwater Pump Discharge Pressures 4.

Subcooling Margin A.

1 & 3 B.

2 & 4 C.

2 & 3 D.

1, 3, & 4 Question:

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Which of the choices below completes the following statements?

During a Reactor startup using 1-OP-1.7 (Unit Startup From Mode 3 to Mode 2 Following Refueling),

critical data will be taken at ___(1)___ amps on the Intermediate Range Detectors.

When the RO uses Control Rods to stabilize power and obtain critical data, the RO will observe a prompt ___(2)___ in startup rate.

A.

(1) 5x10

-7 (2) drop B.

(1) 1x10

-8 (2) drop C.

(1) 5x10

-7 (2) jump D.

(1) 1x10

-8 (2) jump Question:

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Given Conditions:

U1 is at 100% power

The Unit will be ramped offline for a Refueling Outage in accordance with 1-OP-3.7 (Unit Shutdown From Mode 1 to Mode 5 For Refueling)

Based on the above, which of the choices below completes the following statements?

A more linear response during the ramp is ensured by transferring Main Turbine Controls to

___(1)___ when less than 98% power.

During the ramp, the BOP will observe the ___(2)___ Valve Position lowering.

A.

(1) OPEN LOOP (2) Throttle B.

(1) OPEN LOOP (2) Governor C.

(1) IMP-IN (2) Throttle D.

(1) IMP-IN (2) Governor Question:

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Initial Conditions:

U1 is at 100% power

1-CN-P-1A ("A" Main Condensate Pump) and 1-CN-P-1C ("C" Main Condensate Pump) are running

MCR Condenser Pressure indication is 2 in. HgA Current Conditions:

1-AR-Q-E1 (FEEDWATER HEATER 5A HI-LO LEVEL) is locked in

The Turbine 1 Operator reports the following:

o 1-FW-E-5A (5A Feedwater Heater) High Level Divert Valve is fully open o

1-FW-E-5A Feedwater Heater Level is rising Assuming no Operator action, which of the choices below completes the following statements?

As level continues to rise in 1-FW-E-5A, 1-CN-P-1A and 1-CN-P-1C will ___(1)___.

The Shell Side Pressure in 1-FW-E-1A (1A Feedwater Heater) will ___(2)___.

A.

(1) remain running (2) remain stable B.

(1) remain running (2) lower C.

(1) trip (2) remain stable D.

(1) trip (2) lower Question:

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Given Conditions:

1-GW-O2R-102 (WGDT Outlet O2 Analyzer) is aligned to 1-GW-TK-1A ("A" Waste Gas Decay Tank)

1-AR-Y-E4 (Gaseous Waste Decay Tank Analyzer Trouble) is lit Which of the choices below completes the following statements?

1-AR-Y-E4 will illuminate when Oxygen Concentration is greater than or equal to ___(1)___%.

Oxygen Concentration is required to be reduced to ___(2)___.

A.

(1) 2 (2) prevent a potentially flammable/explosive gas mixture B.

(1) 2 (2) minimize oxidation (rusting) of the Carbon Steel Inner Tank C.

(1) 4 (2) prevent a potentially flammable/explosive gas mixture D.

(1) 4 (2) minimize oxidation (rusting) of the Carbon Steel Inner Tank Question:

(1 point) 62 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 62 of 100

The U1 Circulating Water Pumps are normally powered from the ___(1)___ through Feeder Breaker

___(2)___.

A.

(1) "B" RSST (2) 1-EP-BKR-15G1 B.

(1) "B" RSST (2) 1-EP-BKR-15G10 C.

(1) "C" RSST (2) 1-EP-BKR-15G1 D.

(1) "C" RSST (2) 1-EP-BKR-15G10 Question:

(1 point) 63 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 63 of 100

Initial Conditions:

Unit 1 is at 100% power

The 480V "J" busses are de-energized

CO2 spuriously actuated in the U2 Cable Spreading Room

U1 has entered 1-AP-20 (Operation From The Auxiliary Shutdown Panel) due to excessive CO2 Monitor readings in the MCR Current Conditions:

The U1 Auxiliary Shutdown Panel is manned Based on the conditions above, which one of the following evolutions can be performed using controls WITHIN the U1 Auxiliary Shutdown Panel?

A.

Increase Pressurizer Pressure using Group 4 Heaters.

B.

Throttle AFW Flow to the "B" S/G.

C.

Control RCS Temperature using Steam Dumps.

D.

Initiate an Emergency Boration using 1-CH-P-2B ("B" Boric Acid Storage Tank Pump).

Question:

(1 point) 64 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 64 of 100

Given Conditions:

U1 is shutdown for a Refueling Outage

Core offload is in progress

Fuel Building Radiation Automatic Interlock switch is in ENABLE

A Fuel Handling Accident occurs in the Fuel Building

High and Hi-Hi alarms are lit on 1-RM-RMS-153 (Fuel Pit Bridge Radiation Monitor)

Which of the choices below completes the following statements?

MCR Bottled Air will dump automatically ___(1)___.

In accordance with 0-AP-5.1 (Common Unit Radiation Monitoring System), the BOP will ___(2)___.

A.

(1) with no time delay (2) place Fuel Building Exhaust Fans through the Charcoal Filters B.

(1) with no time delay (2) secure the Fuel Building Exhaust Fans C.

(1) following a 2-minute time delay (2) place Fuel Building Exhaust Fans through the Charcoal Filters D.

(1) following a 2-minute time delay (2) secure the Fuel Building Exhaust Fans Question:

(1 point) 65 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 65 of 100

Which of the following completes both statements regarding plant announcements in accordance with OP-AP-100 (Conduct of Operations)?

When starting and stopping equipment from the Control Room, Operations Personnel are expected to make a plant announcement for loads with a MINIMUM Voltage of ___(1)___ Volts.

At a minimum, the plant announcement must include the planned activity and direction for plant personnel to stand clear of the ___(2)___.

A.

(1) 4160 (2) component being started/stopped ONLY B.

(1) 4160 (2) component being started/stopped, including its associated Electrical Switchgear C.

(1) 480 (2) component being started/stopped ONLY D.

(1) 480 (2) component being started/stopped, including its associated Electrical Switchgear Question:

(1 point) 66 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 66 of 100

Which of the choices below completes the following statements regarding board hours in accordance with OP-AA-103 (Operator Qualifications)?

An Operator must stand ___(1)___ 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months watches per calendar quarter to maintain an active license.

The official record for maintaining an active license is recorded ___(2)___.

A.

(1) 5 (2) in Emp Center B.

(1) 5 (2) on the OP-AA-103 Proficiency Sheet C.

(1) 7 (2) in Emp Center D.

(1) 7 (2) on the OP-AA-103 Proficiency Sheet Question:

(1 point) 67 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 67 of 100

Which of the choices below completes the following statements in accordance with Attachment 17 (Feedback Incorporation Process-FIP) of AD-AA-100 (Technical Procedure Process Control)?

A FIP ___(1)___ require an independent review prior to the cognizant management review and approval.

A FIP ___(2)___ be used to perform work in the field.

A.

(1) does not (2) cannot B.

(1) does not (2) can C.

(1) does (2) cannot D.

(1) does (2) can Question:

(1 point) 68 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 68 of 100

Given Conditions:

A tagout for 1-CH-FL-4A (U1 "A" Seal Injection Filter) is in progress

Operator performing the tagout reports the following:

o The Inlet Valve Reach Rod Shear Pin broke o

The Reach Rod will no longer operate the valve o

The valve has been manually closed from inside the Seal Injection Cubicle Which of the choices below completes the following statements regarding tagging the Inlet Valve in accordance with OP-AA-200 (Equipment Clearance)?

Inside the Seal Injection Cubicle, the Operator will hang a ___(1)___ tag.

In the Charging Pump Alleyway, the Operator will hang a ___(2)___ tag.

A.

(1) Danger (2) Caution B.

(1) Danger (2) Danger C.

(1) Caution (2) Danger D.

(1) Caution (2) Caution Question:

(1 point) 69 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 69 of 100

If an alarm is received when attempting to leave the Protected Area through the Main Gate, the posted Station Standard requires an individual to step back from the Monitor and ______

A.

immediately notify HP of the alarm.

B.

attempt a second exit through the Personnel Monitor. If it alarms, back out of the monitor and immediately notify HP of the alarms.

C.

attempt a second exit through the Personnel Monitor. If it alarms, back out and go to the HP office.

D.

perform a whole body frisk using the Hand Held Monitor. If clean, leave the Protected Area.

Question:

(1 point) 70 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 70 of 100

Differential Boron Worth (K/K/ppm) will become ___(1)___ negative as Moderator Temperature increases because, at higher temperatures, a 1 ppm increase in Reactor Coolant Boron Concentration will add ___(2)___ Boron Atoms to the Core.

A.

(1) more (2) more B.

(1) more (2) fewer C.

(1) less (2) more D.

(1) less (2) fewer Question:

(1 point) 71 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 71 of 100

Given Conditions:

U1 is at 100% power

All Steam Flow, Feed Flow, and First-Stage Pressure Selector Switches are on Channel IV

1-MS-PT-1476 ("A" S/G Channel IV Steam Pressure Transmitter) fails high Which of the choices below completes the following statements?

1-MS-FI-1475 ("A" S/G Channel IV Steam Flow Indicator) will ___(1)___.

Assuming no Operator action, "A" S/G Level will ___(2)___.

A.

(1) fail high (2) rise B.

(1) fail low (2) lower C.

(1) remain the same (2) remain the same D.

(1) fail low (2) rise Question:

(1 point) 72 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 72 of 100

Given Conditions:

U1 is at 100% power

All S/G Pressures are 810 psig

Safeguards Operator reports steam is issuing from the Tailpipe of 1-MS-PCV-101A

("A" S/G PORV)

Based on the above, select the approximate Tailpipe Exit Temperature of the steam.

A.

212°F B.

310°F C.

380°F D.

530°F Question:

(1 point) 73 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 73 of 100

Given Conditions

Hot work is going to be performed in the U1 Cable Vault and Tunnel

CO2 lock-out is required Based on the above, which of the choices completes the following statements?

The U1 Cable Vault and Tunnel is provided CO2 from the ___(1)___-ton CO2 Tank In accordance with 0-OP-52.4 (Low Pressure Carbon Dioxide System B), CO2 lockout will be performed ___(2)___.

A.

(1) 6 (2) locally in the U1 Emergency Switchgear Room B.

(1) 6 (2) remotely from the Main Control Room C.

(1) 17 (2) locally in the U1 Emergency Switchgear Room D.

(1) 17 (2) remotely from the Main Control Room Question:

(1 point) 74 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 74 of 100

Given Conditions:

U1 is at 60% power

S/G Feed Flows are as follows:

o "A" S/G: 2.5 x 10 6 lbm/hr o

"B" S/G: 2.5 x 10 6 lbm/hr 6

o "C" S/G: 2.5 x 10 lbm/hr

S/G Pressures are as follows:

o "A" S/G: 840 psig o

"B" S/G: 840 psig o

"C" S/G: 840 psig

Feedwater Temperature entering each S/G is 390°F

Steam Quality is 100%

Using a simplified heat balance, what is the approximate Reactor Core Thermal Power?

A.

588 MWt B.

611 MWt C.

1764 MWt D.

1833 MWt Question:

(1 point) 75 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 75 of 100

Given Conditions:

U1 is in Mode 5

The Reactor Vessel Head is on

All Pressurizer Safety Valves are installed

The following annunciators are locked in o

1-AR-C-F4 (RCS OVERPRESS) o 1-AR-C-D1 (PRESSURIZER SAFETY VALVE OR PORV OPEN)

1-RC-PT-1403 (RCS Wide Range Pressure Transmitter) is failed high Which of the choices below completes the following statements?

In accordance with 1-AR-C-F4, manual control of 1-RC-PCV-1456 (Pressurizer PORV) ___(1)___

required.

Per T.S. 3.4.12 (LTOPS: Low Temperature Overpressure Protection System), the associated PORV must be restored to OPERABLE status within ___(2)___.

A.

(1) is not (2) 7 days B.

(1) is not (2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C.

(1) is (2) 7 days D.

(1) is (2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Question:

(1 point) 76 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 76 of 100

Given Conditions:

U1 is at 100% power

1-AP-44 (Loss of RCS Pressure) IOAs are complete

RCS Pressure continues to lower

The following indications are observed Based on the conditions above, ______ when 1-AP-44 is complete.

A.

all RCPs are running B.

all RCPs are secured C.

1-RC-P-1A ("A" Reactor Coolant Pump) is the only RCP running D.

1-RC-P-1B ("B" Reactor Coolant Pump) is the only RCP running Question:

(1 point) 77 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 77 of 100

Initial Conditions:

U1 is at 100% power

1H EDG is tagged out for 6 year PMs Current Conditions:

Time 1130:

o Loss of Offsite Power occurred

Time 1135:

o 1-AR-H-H2 (EMER DIESEL GEN #1J DIFFERENTL) is locked in o

Safeguards operator dispatched from the MCR with a copy of 0-OP-6.4 (Operation of the SBO-SBO Event)

Which of the choices below completes the following statements?

The HIGHEST Emergency Classification is ___(1)___.

The MAXIMUM time allowed to make an initial report to State and Local Governments following EAL Declaration is ___(2)___ minutes.

REFERENCE PROVIDED A.

(1) an Alert (2) 15 B.

(1) an Alert (2) 30 C.

(1) a Site Area Emergency (2) 15 D.

(1) a Site Area Emergency (2) 30 Question:

(1 point) 78 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 78 of 100

Given Conditions:

A Loss of Offsite Power has occurred

The 1H EDG is running and loaded on the 1H Bus

The 1J EDG tripped on 87X Generator Differential Which of the choices below completes the following statements?

The 1H EDG Loads will be configured using ___(1)___.

Prior to configuring loads, the 1H EDG ___(2)___ Operable.

A.

(1) 1-ECA-0.0 (Loss of All AC Power)

(2) is not B.

(1) 1-ECA-0.0 (Loss of All AC Power)

(2) is C.

(1) 0-AP-10 (Loss of Electrical Power)

(2) is not D.

(1) 0-AP-10 (Loss of Electrical Power)

(2) is Question:

(1 point) 79 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 79 of 100

Initial Conditions:

Both Units are at 100% power

SBO Diesel is tagged out for maintenance

At time 1100 on 9/1/21, 1-SI-P-1A ("A" Low Head Safety Injection Pump) Shaft sheared during 1-PT-57.1A (Emergency Core Cooling Subsystem-Low Head Safety Injection Pump)

1-SI-P-1A has been tagged out Current Conditions:

Time is 2300 on 9/1/21

The crew has entered 0-AP-8 (Response to Grid Instability) due to a Severe Grid Disturbance

500KV Switchyard Voltage is 497KV and stable Which of the choices below completes the following statements?

1-SI-P-1B ("B" Low Head Safety Injection Pump) ___(1)___.

The most limiting action will be to ___(2)___.

REFERENCE PROVIDED A.

(1) is currently inoperable (2) restore one Offsite Circuit to operable status by 2300 on 9/2/21 B.

(1) is currently inoperable (2) enter LCO 3.0.3 immediately C.

(1) will be inoperable at 1100 on 9/2/21 (2) restore one Offsite Circuit to operable status by 2300 on 9/2/21 D.

(1) will be inoperable at 1100 on 9/2/21 (2) enter LCO 3.0.3 at 1100 on 9/2/21 Question:

(1 point) 80 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 80 of 100

Initial Conditions:

2-FW-P-3A ("A" Motor Driven Auxiliary Feedwater Pump) is tagged out

U2 has experienced a Large Break LOCA o

2-FW-P-3B ("B" Motor Driven Auxiliary Feedwater Pump) tripped on motor overload o

The crew has completed step 9 (Align SI System for Cold Leg Recirculation) of 2-ES-1.3 (Transfer to Cold Leg Recirculation)

Current Conditions:

2-AR-F-D8 (TURBINE DRIVEN AFW PUMP TROUBLE OR LUBE OIL TRBL) is lit o

2-FW-P-2 (Turbine Driven Auxiliary Feedwater Pump) has severe mechanical damage

The crew has implemented Attachment 3 (Containment Sump Strainer Blockage or Loss of Suction) of 2-ES-1.3 o

2-SI-P-1A and 2-SI-P-1B have both been secured Which of the choices below completes the following statements?

___(1)___ required both Low Head Safety Injection Pumps to be secured.

The correct procedure to transition to is ___(2)___.

A.

(1) Oscillating Amps and Flow (2) 2-ECA-1.1 (Loss of Emergency Coolant Recirculation)

B.

(1) Oscillating Amps and Flow (2) 2-FR-H.1 (Loss of Secondary Heat Sink)

C.

(1) RWST Level of 8%

(2) 2-ECA-1.1 (Loss of Emergency Coolant Recirculation)

D.

(1) RWST Level of 8%

(2) 2-FR-H.1 (Loss of Secondary Heat Sink)

Question:

(1 point) 81 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 81 of 100

Given Conditions:

U1 is currently at 97% power

The crew is in 1-AP-1.3 (Control Rod Out of Alignment) due to rod H14 in Control Bank "D" being misaligned

Current QPTR value is 1.0700 Which one of the following describes the maximum allowed reactor power level if the QPTR cannot be restored within the required limits of T.S. 3.2.4 (QPTR)?

A.

85%

B.

82%

C.

79%

D.

76%

Question:

(1 point) 82 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 82 of 100

Initial Conditions:

U1 was at 3% power following a Refueling Outage

N35 (Channel I Intermediate Range Detector) failed low and the crew entered 1-AP-4.2 (Malfunction of Intermediate Range Nuclear Instrumentation )

Decision was made to continue ramping to > P-10 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per T.S. 3.3.1F.2 and 1-AP-4.2

N35 Operation Selector Switch is in the 1x 10

-9 Amps position

N35 Level Trip Switch is in BYPASS Current Conditions:

U1 is at 8% power and ramping to >P-10

N-36 (Channel II Intermediate Range Detector) fails low Based on the conditions above, the Reactor ___(1)___ automatically trip and a ___(2)___

hour notification to the NRC is required in accordance with VPAP-2802 (Notifications and Reports).

REFERENCE PROVIDED A.

(1) did (2) 8 B.

(1) did (2) 4 C.

(1) did not (2) 8 D.

(1) did not (2) 4 Question:

(1 point) 83 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 83 of 100

Assuming NO additional actions, which one of the following will result in a required Technical Specification shutdown within the next 31 days?

A.

1-BD-TV-100B ("A" S/G Blowdown Inside Containment Isolation Valve) failed in mid position.

B.

1-VG-TV-100B (Containment Gaseous Vent Header Inside Containment Isolation Valve) failed in mid position with 1-VG-TV-100A (Containment Gaseous Vent Header Outside Containment Isolation Valve) closed and de-activated.

C.

Both Personnel Hatch Airlock Doors shut and locked with the Outer Door Seals failed.

D.

Both Personnel Hatch Airlock Doors shut and locked with the Airlock Door Interlock Mechanism inoperable.

Question:

(1 point) 84 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 84 of 100

Given Conditions:

The crew has entered 1-FR-H.2 (Response To Steam Generator Overpressure)

The SRO is reading the NOTE before step 1

The Main Steam Trip Valves are closed

Secondary Radiation is normal

The S/G Narrow Range Levels are as follows:

o "A": 92% and rising o

"B": 60% and stable o

"C": 58% and stable

The S/G Pressures are as follows o

"A": 1210 psig and rising o

"B": 1130 psig and stable o

"C": 1125 psig and stable Which of the choices below completes the following statements?

The SRO will ___(1)___.

In accordance with OP-AP-104 (Emergency and Abnormal Operating Procedures), Yellow Path Procedures

___(2)___.

A. (1) transition to 1-FR-H.3 (Response to Steam Generator High Level) and initiate blowdown from the "A" S/G (2) are entered at the discretion of the SRO B. (1) remain in 1-FR-H.2 and use manual control of the "A" S/G PORV to reduce pressure to less than S/G Safety Valve setpoints (2) are entered at the discretion of the SRO C. (1) transition to 1-FR-H.3 (Response to Steam Generator High Level) and initiate blowdown from the "A" S/G (2) must be entered provided a Red or Orange Path doesnt exist D.

(1) remain in 1-FR-H.2 and use manual control of the "A" S/G PORV to reduce pressure to less than S/G Safety Valve setpoints (2) must be entered provided a Red or Orange Path doesnt exist Question:

(1 point) 85 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 85 of 100

Initial Conditions:

U1 is in Mode 5

RCS is at 110°F and 350 psig

All S/G Narrow Range Levels are off-scale low

All Secondary Side Manways have been removed

All RCPs are available

1-RH-P-1A ("A" Residual Heat Removal Pump) is in service with normal RHR Flows Current Conditions:

1-RH-MOV-1701 (RHR Suction Isolation Valve) Disc separated from the Stem and has fallen into the flow stream

Crew entered 1-AP-11 (Loss of RHR) and secured 1-RH-P-1A

RCS Temperature is rising Based on the conditions above and in accordance with 1-AP-11, the crew will A.

start at least one RCP and establish Forced Circulation B.

go to attachment 10 and establish Natural Circulation C.

go to attachment 8 and establish Reflux Boiling D.

go to attachment 6 and establish Forced Feed and Spill Question:

(1 point) 86 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 86 of 100

Given Conditions:

U1 is in a RFO

Operators are clearing a tagout on the PRT o

PRT was completely drained for internal maintenance

PRT is desired to be filled to 75% on board indication 1-RC-LI-1470 (PRT Level Indicator)

Which of the choices below completes the following statements?

To obtain 75% indicated level, the crew will add ___(1)___ gallons of PG.

The PRT is designed to accept a steam discharge from the Pressurizer equal to ___(2)___ of the Pressurizer Steam Volume at full power.

REFERENCE PROVIDED A.

(1) 7600 (2) 100%

B.

(1) 7600 (2) 110%

C.

(1) 7823 (2) 100%

D.

(1) 7823 (2) 110%

Question:

(1 point) 87 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 87 of 100

Initial Conditions:

Both Units are at 100% power

Service Water is aligned to both Unit's CARFs

The Mechanical Chiller is tagged out Current Conditions:

A Spurious Unit 2 Safety Injection signal occurs Which of the choices below completes the following statements?

On Unit 1, Containment Temperature will ___(1)___.

To restore Unit 1 Containment Temperature, the SRO will direct the crew to ___(2)___.

A.

(1) lower (2) re-align Service Water IAW 0-OP-49.1 (Service Water System Normal Operations)

B.

(1) lower (2) close 1-SW-TV-101A&B (SW Supply and Return to CARFs) IAW 1-OP-21.1 (Containment Ventilation)

C.

(1) rise (2) re-align Service Water IAW 0-OP-49.1 (Service Water System Normal Operations)

D.

(1) rise (2) open 1-SW-TV-101A&B (SW Supply and Return to CARFs) IAW 1-OP-21.1 (Containment Ventilation)

Question:

(1 point) 88 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 88 of 100

Given Conditions:

U1 is at 100% power

Maintenance is required on 1-BY-C-02 (Normal Battery Charger 1-I)

1-BY-C-03 (Swing Battery Charger 1C-I) is placed in service Which of the choices below completes the following statements?

1-BY-C-03 ___(1)___ be used to satisfy the LCO for T.S. 3.8.4 (DC Sources-Operating).

In accordance with T.S. 3.8.4 Bases, each Station Battery has adequate storage capacity to carry the required load continuously for at least ___(2)___ hours.

A.

(1) cannot (2) 2 B.

(1) cannot (2) 8 C.

(1) can (2) 2 D.

(1) can (2) 8 Question:

(1 point) 89 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 89 of 100

Given Conditions:

Both Units are at 100% power

2-IA-C-1 (U2 Instrument Air Compressor) is tagged out

A rupture has occurred on 2-SA-TK-2 (Turbine Building Service Air Receiver)

Instrument Air Pressure is rapidly lowering In accordance with 1-AP-28 (Loss of Instrument Air), the Unit Supervisor will direct the crew to trip the Reactor and enter 1-E-0 (Reactor Trip or Safety Injection) at ___(1)___ psig to prevent "one at a time" closure of the Main Steam Trip Valves, which would cause a High Steam Line ___(2)___ Safety Injection.

A.

(1) 70 (2) Delta P B.

(1) 75 (2) Delta P C.

(1) 70 (2) Flow D.

(1) 75 (2) Flow Question:

(1 point) 90 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 90 of 100

Initial Conditions:

Both Units are at 100% power

1-AR-E-C6 (SPENT FUEL PIT LO LEVEL) is locked in

The AB Operator reports the SFP level is at -6" and lowering on the local gauge

Crew enters 0-AP-27 (Malfunction of Spent Fuel Pit System)

Current Conditions:

All attempts to makeup to the SFP from either Unit's Blender are unsuccessful Which of the choices below completes the following statements?

The order for using additional makeup sources to the SFP is ___(1)___.

Permission from the ___(2)___ is required to use these sources.

A.

(1) Fire Protection, then Service Water (2) Shift Manager B.

(1) Fire Protection, then Service Water (2) Manager Nuclear Operations or Operations Manager on Call C.

(1) Service Water, then Fire Protection (2) Shift Manager D.

(1) Service Water, then Fire Protection (2) Manager Nuclear Operations or Operations Manager on Call Question:

(1 point) 91 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 91 of 100 preferred

Given Conditions:

U1 is in a Refueling Outage

Control Rod Drive Shaft (CRDS) Unlatching is in progress

The Shift Manager notifies the Refueling SRO of a death in their immediate family

The Refueling SRO immediately leaves the Manipulator Crane and exits the RCA Which of the choices below completes the following statements?

CRDS Unlatching ___(1)___.

If the CRDSs are not unlatched, the ___(2)___ Lift cannot be performed.

A.

(1) can continue (2) Reactor Vessel Head B.

(1) can continue (2) Upper Internals C.

(1) must be secured until another Refueling SRO shows up (2) Reactor Vessel Head D.

(1) must be secured until another Refueling SRO shows up (2) Upper Internals Question:

(1 point) 92 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 92 of 100

Which of the choices below completes the following statements?

In accordance with T.R. 3.3.7 (Radiation Monitoring Instrumentation) Surveillance Requirement, 1-RM-RMS-152 (New Fuel Storage Radiation Monitor) must be channel checked once every ___(1)___

hours.

In accordance with 0-LOG-6A (Backboards Logs) Channel Check requirements, if no upscale deflection is evident and net activity is zero, the Operator is required to place the Operation Selector Switch to the

___(2)___ position and verify the needle is free to move.

A.

(1) 6 (2) Check Source B.

(1) 6 (2) Pulse Cal C.

(1) 12 (2) Check Source D.

(1) 12 (2) Pulse Cal Question:

(1 point) 93 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 93 of 100

Which of the choices below completes the following statement regarding Standing Orders that are created due to an inadequate Technical Specification requirement?

In accordance with OP-AA-100 (Conduct of Operations), the Standing Order can be more restrictive and must be approved by the ______.

A.

Operations Manager B.

Plant Manager C.

Site Vice President D.

Facility Safety Review Committee (FSRC)

Question:

(1 point) 94 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 94 of 100

Given Conditions:

U2 is in a Refueling Outage with Core Offload in progress

The Manipulator Crane is in transit to the Upender with a Fuel Assembly

Refueling Cavity Level is lowering in an uncontrolled manner

The crew enters 2-AP-52 (Loss of Refueling Cavity Level During Refueling)

Which of the choices below completes the following statement?

In accordance with 2-AP-52, the expected crew response is to ______.

A.

place the Fuel Assembly in the Containment Upender with the frame down B.

position the Manipulator Crane over the deepest part of the Fuel Transfer Canal and leave the Fuel Assembly suspended C.

quickly transfer the Fuel Assembly to the Spent Fuel Pit D.

return the Fuel Assembly back in the Reactor Vessel Question:

(1 point) 95 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 95 of 100

Given Conditions:

U1 is in Mode 3 following a Refueling Outage

1-PT-46.21 (RCS Pressure Boundary Components Affected by Boric Acid Accumulation) walk down identified RX Head Vent Tailpipe leakage of 60 dpm o

Leakage is not contained by existing Splatter Shield o

Leakage target is the Reactor Head

Maintenance has constructed a Splatter Shield Extension that will be attached to the existing Splatter Shield o

Shield will stay in place until next Refueling Outage Which of the choices below completes the following statements in accordance with CM-AA-TCC-204 (Temporary Configuration Changes)?

In accordance with CM-AA-TCC-204 (Temporary Configuration Changes), the Splatter Shield Extension is defined as a ___(1)___ Modification.

In accordance with OP-AA-107 (Subatmospheric Containment Entry), permission to leave the Splatter Shield Extension in Containment is required by ___(2)___.

A.

(1) Temporary (2) Station Engineering only B.

(1) Temporary (2) Station Engineering and the Facility Safety Review Committee (FSRC)

C.

(1) Procedurally Controlled Temporary (2) Station Engineering only D.

(1) Procedurally Controlled Temporary (2) Station Engineering and the Facility Safety Review Committee (FSRC)

Question:

(1 point) 96 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 96 of 100

Given Conditions:

U2 is at 100% power Which one of the emergent maintenance items below must be considered Priority 1 Maintenance in accordance with WM-AA-100 (Work Management)?

ASSUME ONLY THE STATED COMPONENT IS INOPERABLE IN EACH CASE A.

Replacement of the U2 "C" Charging Pump Motor.

B.

Replacement of the U2 "A" Service Water Pump Motor.

C.

Replacement of the U2 "A" Casing Cooling Pump Motor.

D.

Replacement of the U2 "A" Motor Driven Auxiliary Feedwater Pump Motor.

Question:

(1 point) 97 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 97 of 100

Which of the choices below completes the following statements for a Design Basis Steam Generator Tube Rupture?

In accordance with 10CFR50, Appendix A, General Design Criteria 19 (Control Room), the maximum expected dose to a MCR Operator is ___(1)___ Rem.

In accordance with T.S. 3.7.10 (Main Control Room/Emergency Switchgear Room Emergency Ventilation System), ___(2)___ (Emergency Ventilation Fan) cannot be used to satisfy the LCO.

A.

(1) 5 (2) 1-HV-F-41 B.

(1) 25 (2) 1-HV-F-41 C.

(1) 5 (2) 2-HV-F-41 D.

(1) 25 (2) 2-HV-F-41 Question:

(1 point) 98 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 98 of 100

Which ONE of the following responsibilities CAN the Interim Station Emergency Manager delegate?

A.

Protective Action Recommendation to the State B.

Declaration of the Emergency Classification upgrade C.

Authorization of emergency exposure to plant personnel D.

Initiation of EPIP-1.03 (Response to Alert)

Question:

(1 point) 99 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 99 of 100

Which of the choices below completes the following statements?

The LOWEST Emergency Action Level that requires a Protective Action Recommendation (PAR) is a

___(1)___ Emergency.

When completing Attachment 2 (Affected Sectors Map) of EPIP-1.06 (Protective Action Recommendation), the Average Wind Direction is recorded in degrees ___(2)___.

A.

(1) General (2) to B.

(1) Site Area (2) to C.

(1) General (2) from D.

(1) Site Area (2) from Question:

(1 point) 100 North Anna Power Station 2022 NAPS ILT NRC SRO EXAMINATION Page 100 of 100

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

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT Notification of UNUSUAL EVENT M

System Malfunct.

1 Loss of Emergency AC Power 6

RPS Failure 7

Loss of Commun.

MS6 Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal MU1.1 Loss of all offsite AC power capability, Table M-1, to Unit 1(2) 4160V emergency buses H and J for 15 min. (Note 1)

MA1.1 AC power capability, Table M-1, to Unit 1(2) 4160V emergency buses H and J 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 MS6.1 An automatic or manual trip did not shut down the reactor as indicated by reactor power 5%

AND All actions taken 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 MA6 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 MA6.1 An automatic or manual trip did not shut down the reactor as indicated by reactor power 5%

AND Subsequent automatic or manual trip actions (trip switches and manual turbine trip) are not successful in shutting down the reactor as indicated by reactor power 5% (Note 8)

MU7 Loss of all onsite or offsite communications capabilities MU7.1 Loss of all Table M-5 onsite communication methods OR Loss of all Table M-5 State and local agency communication methods OR Loss of all Table M-5 NRC communication methods MU4 Reactor coolant activity greater than Technical Specification allowable limits MU4.2 MU5 RCS leakage for 15 minutes or longer MU5.1 RCS unidentified or pressure boundary leakage

> 10 gpm for 15 min.

OR RCS identified leakage > 25 gpm for 15 min.

OR Leakage from the RCS to a location outside containment

> 25 gpm for 15 min.

(Note 1)

MU6 Automatic or manual trip fails to shut down the reactor MU6.1 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 (trip switches or manual turbine trip) are successful in shutting down the reactor as indicated by reactor power

< 5% (Note 8)

MG1 Prolonged loss of all offsite and all onsite AC power to emergency buses MS1 Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer MS1.1 Loss of all offsite and all onsite AC power to Unit 1(2) 4160V emergency buses H and J for 15 min. (Notes 1, 14)

MG1.1 Loss of all offsite and all onsite AC power to Unit 1(2) 4160V emergency buses H and J AND EITHER Long-term RCS heat removal capability is not likely to be established and maintained per procedure Core Cooling-RED Path conditions met MS2 Loss of all vital DC power for 15 minutes or longer MS2.1 Indicated voltage is < 105 VDC on all vital 125 VDC battery buses for 15 min. (Note 1)

MA3 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress MA3.1 MU3 UNPLANNED loss of Control Room indications for 15 minutes or longer MU3.1 Dose rate at 1 ft. from an unpressurized RCS sample Table M-4 None EAL MATRIX - HOT CONDITIONS (RCS > 200°F)

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

AND Any significant transient is in progress, Table M-3 MA1 Loss of all but one AC power source to emergency buses for 15 minutes or longer MU1 Loss of all offsite AC power capability to emergency buses for 15 minutes or longer None None 2

Loss of Vital DC Power 3

Loss of CR Indications None None MU4.3 MG2 Loss of all emergency AC and vital DC power sources for 15 minutes or longer MG2.1 Loss of all offsite and all onsite AC power to Unit 1(2) 4160V emergency buses H and J for 15 min. (Note 1)

AND Indicated voltage is < 105 VDC on all vital 125 VDC battery buses for 15 min. (Note 1)

None None 4

RCS Activity 5

RCS Leakage MA9 Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode MA9.1 9

Hazardous Event Affecting Safety Systems None None None None None None None None None 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 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 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 2

3 4

1 2

3 4

1 MU8 Failure to isolate containment or loss of containment pressure control MU8.1 Any penetration is not closed within 15 min. of a VALID Phase A or B isolation signal OR CTMT pressure > 28 psia with < one full train of CTMT depressurization equipment (Note 11) operating per design for 15 min. (Note 1) 2 3

4 1

2 3

4 1

None None None 8

CTMT Failure The occurrence of any Table M-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)

NAPS MU6.2 A manual trip did not shut down the reactor as indicated by reactor power 5%

AND A subsequent manual trip (trip switches or manual turbine trip) OR automatic trip is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8)

FG1.1 1

2 3

4 1

2 3

4 Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1)

FS1.1 1

2 3

4 1

2 3

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

FA1.1 1

2 3

4 1

2 3

4 Any loss or any potential loss of EITHER Fuel Clad or RCS (Table F-1)

F Fission Product Barrier Degradation None Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CTMT) Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. Core Cooling-RED Path conditions met

2. CTMT High Range Radiation Monitor RM-RMS-165/166(265/

266) reading > Table F-2 column Fuel Clad Loss

3. Coolant activity > 300 Ci/gm DEI-131

2. CTMT High Range Radiation Monitor RM-RMS-165/166(265/266) reading > Table F-2 column RCS Loss

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

UNISOLABLE RCS leakage

SG tube RUPTURE

3. Any condition in the opinion of the SEM that indicates loss of the RCS barrier

2. CTMT isolation (Phase A or B) is required AND EITHER:

CTMT integrity has been lost based on SEM judgment UNISOLABLE pathway from CTMT atmosphere to the environment exists

4. Any condition in the opinion of the SEM that indicates loss of the CTMT barrier

3. Containment RED Path conditions met

2. CTMT High Range Radiation Monitor RM-RMS-165/166(265/

266) reading > Table F-2 column CTMT Potential Loss

6. Any condition in the opinion of the SEM that indicates potential loss of the CTMT barrier A

RCS or SG Tube Leakage B

Inadequate Heat Removal C

CTMT Radiation /

RCS Activity D

CTMT Integrity or Bypass E

SEM Judgment None None None None None

1. A leaking or RUPTURED SG is FAULTED outside of CTMT None None None

5. CTMT pressure > 28 psia with

< one full train of CTMT depressurization equipment (Note 11) operating per design for 15 min. (Note 1)

4. CTMT hydrogen concentration 4%

None None

1. Core Cooling-ORANGE Path conditions met

2. Heat Sink-RED Path conditions met AND Heat sink is required

3. Indications of UNISOLABLE RCS leakage outside of CTMT

7. Any condition in the opinion of the SEM that indicates loss of the Fuel Clad barrier

3. Any condition in the opinion of the SEM that indicates potential loss of the Fuel Clad barrier

4. Any condition in the opinion of the SEM that indicates potential loss of the RCS barrier FPB Category

3. Heat Sink-RED Path conditions met AND Heat sink is required

2. Integrity-RED Path conditions met

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

4. Dose rate at 1 ft. from an unpressurized RCS sample Table F-3 None

5. Sample line dose rate threshold Table F-4

1. UNISOLABLE RCS or SG tube leakage > 150 gpm 1

1 1

1 Table M-5 Communications Methods System Onsite State/

Local NRC Radio Communications System X

Public Address and Intercom System X

Private Branch Telephone Exchange (PBX)

X X

X Sound Powered Telephone System X

Commercial Telephone System X

X Enterprise Transport Network X

DEENS X

Dedicated NRC Communications X

Table M-5 Communications Methods System Onsite State/

Local NRC Radio Communications System X

Public Address and Intercom System X

Private Branch Telephone Exchange (PBX)

X X

X Sound Powered Telephone System X

Commercial Telephone System X

X Enterprise Transport Network X

DEENS X

Dedicated NRC Communications X

MU4.1 With letdown in service, Reactor Coolant Letdown Radiation Monitor 1(2)CH-RI-128(228) > 2.4E+04 mrem/hr 2

3 4

1 2

3 4

1

6. With letdown in service, Reactor Coolant Letdown Radiation Monitor CH-RI-128(228) > 7.5E+04 mR/hr GENERAL EMERGENCY SITE AREA EMERGENCY ALERT R

Abnormal Rad Levels/

Rad Effluent 1

Rad Effluent 2

Irradiated Fuel Events 1

Security 2

Seismic Event 3

Natural or Tech.

Hazard 6

Control Room Evacuation 7

SEM Judgment RG1 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem adult thyroid CDE RS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem adult thyroid CDE RA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem adult thyroid CDE RU1 Release of gaseous or liquid radioactivity greater than 2 times the allocated ODCM limits for 60 minutes or longer RA2 Significant lowering of water level above, or damage to, irradiated fuel RU2 UNPLANNED loss of water level above irradiated fuel RU2.1 RA2.2 Damage to irradiated fuel resulting in a release of radioactivity AND

VALID Hi-Hi alarm on any of the following radiation monitors:

- RM-RMS-152 New Fuel Storage Areaa

- RM-RMS-153 Fuel Pit Bridge

- RM-RMS-162 (262) Manipulator Crane Area (Refueling Mode)

- RM-RMS-163 (263) Reactor Containment Area

- RM-RMS-159 (259) Containment Particulate

- RM-RMS-160 (260) Containment Area Gas

VALID high alarm on VG-RI-180-1 Vent Stack B Normal Range RA2.1 IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown RA3.1 Dose rates > 15 mR/hr in EITHER of the following:

Control Room Central Alarm Station (by survey)

HU2 Seismic event greater than OBE levels HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by NAPS Security Shift Supervisor OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat HA1.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by NAPS Security Shift Supervisor OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site HS1 HOSTILE ACTION within the PLANT PROTECTED AREA HS1.1 A HOSTILE ACTION is occurring or has occurred within the PLANT PROTECTED AREA as reported by NAPS Security Shift Supervisor HA6.1 An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel HS6 Inability to control a key safety function from outside the Control Room HS6.1 An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel AND Control of any of the following key safety functions is not re-established within 15 min. of the last licensed operator leaving the Control Room (Note 1):

Reactivity (modes 1, 2 and 3 only)

Core cooling RCS heat removal HU7.1 Other conditions exist which, in the judgment of the SEM, 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.

HA7.1 Other conditions exist which, in the judgment of the SEM, 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.

HS7.1 Other conditions exist which in the judgment of the SEM 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.

HG7.1 Other conditions exist which in the judgment of the SEM 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.

HG7 Other conditions exist which in the judgment of the SEM warrant declaration of a General Emergency HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes HU1 Confirmed SECURITY CONDITION or threat HA6 Control Room evacuation resulting in transfer of plant control to alternate locations HA7 Other conditions exist that in the judgment of the SEM warrant declaration of an Alert HU7 Other conditions existing that in the judgment of the SEM warrant declaration of a NOUE HS7 Other conditions existing that in the judgment of the SEM warrant declaration of a Site Area Emergency None None None None 3

Area Radiation Levels Modes:

None None RA1.1 RA1.3 Reading on any Table R-1 effluent radiation monitor

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

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 adult thyroid CDE > 50 mrem for 60 min. of inhalation.

(Notes 1, 2)

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 adult thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

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 adult thyroid CDE > 500 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 > 1,000 mrem TEDE or 5,000 mrem adult 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 > 1,000 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate adult thyroid CDE > 5,000 mrem for 60 min. of inhalation.

(Notes 1, 2)

Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses

> 10 mrem TEDE or 50 mrem adult thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

None None None HU2.1 1

Confinement Boundary RA1.2 Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem adult thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

RA2.3 Lowering of spent fuel pool level to 10 ft. (Level 2) on 1-FC-LI-105-1, 2 or 2A Spent Fuel Pit Wide Range Level RS2.1 Lowering of spent fuel pool level to 1 ft. (Level 3) on 1-FC-LI-105-1, 2 or 2A Spent Fuel Pit Wide Range Level RS2 Spent fuel pool level at the top of the fuel racks RG2.1 Spent fuel pool level cannot be restored to at least 1 ft.

(Level 3) on 1-FC-LI-105-1, 2 or 2A Spent Fuel Pit Wide Range Level for 60 min. (Note 1)

RG2 Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer RA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 room or area (Note 5)

HU3 Hazardous event HU3.1 A tornado strike within the PLANT PROTECTED AREA HU3.2 Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode HU3.3 Movement of personnel within the PLANT PROTECTED AREA is IMPEDED due to an event external to the PLANT PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

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)

HU4 FIRE potentially degrading the level of safety of the plant 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 in any Table H-1 area 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 (excluding Reactor Contaiment)

AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Notes 1, 13)

HU4.3 A FIRE within the plant PLANT PROTECTED AREA or ISFSI 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 or ISFSI Protected Area that requires an offsite fire department to assist with extinguishment 4

Fire HA5 Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) 5 Hazardous Gases None None None Prepared for Dominion Energy by Operations Support Services, Inc. 7-19-21 1

2 3

4 5

6 DEF Power Operation Defueled 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following:

Spent Fuel Pit Lo Level (1E-C6) alarm

Report of lowering level in refueling cavity or SFP

Loss of SFP Cooling suction flow AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:

RM-RMS-152 New Fuel Storage Area

RM-RMS-153 Fuel Pit Bridge

RM-RMS-162 (262) Manipulator Crane Area (Refueling Mode)

RM-RMS-163 (263) Reactor Containment Area 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

Seismic event > OBE (0.06g horizontal or 0.04g vertical) as indicated by OBE EXCEEDED indicator illuminated on the SYSCOM Network Control Center (NCC)

None 2

3 4

5 6

1 2

3 4

5 6

1 H

Hazards

[Refer to CA6.1 or MA9.1 for potential escalation due to a seismic event]

[Refer to CA6.1 or MA9.1 for potential escalation due to a hazardous event]

[Refer to CA6.1 or MA9.1 for potential escalation due to fire]

NAPS North Anna Power Station Emergency Action Level Matrix Revision 10 E

ISFSI EU1 Damage to a loaded cask CONFINEMENT BOUNDARY EU1.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any Table E-1 limit 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

NAPS 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

Startup Cold Shutdown 200°F Refueling Hot Standby 350°F Hot Shutdown

< 350°F RU1.1 Reading on SW-RM-130(230) CW Discharge Tunnel radiation monitor > 2 x the Hi-Hi setpoint for 60 min.

(Notes 1, 2, 3) 2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

RU1.2 Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x the allocated ODCM limits for 60 min. (Notes 1, 2) 2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

Cable Vaults & Tunnels Emergency Switchgear Rooms Emergency Diesel Generator Rooms Reactor Containment Quench Spray Pump Houses Safeguards Area Main Steam Valve House Cable Spreading Rooms Control Room CR Chiller Rooms Auxiliary / Fuel / Decontamination Buildings Fuel Oil Pump House Room A or B Service Water Pump House and Valve House Intake Structure Control House Auxiliary Service Water Pump House Auxiliary Feedwater Pump House Turbine Building Table H-1 NAPS Fire Areas Cable Vaults & Tunnels Emergency Switchgear Rooms Emergency Diesel Generator Rooms Reactor Containment Quench Spray Pump Houses Safeguards Area Main Steam Valve House Cable Spreading Rooms Control Room CR Chiller Rooms Auxiliary / Fuel / Decontamination Buildings Fuel Oil Pump House Room A or B Service Water Pump House and Valve House Intake Structure Control House Auxiliary Service Water Pump House Auxiliary Feedwater Pump House Turbine Building Table H-1 NAPS Fire Areas Table R-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building El 274' Instrument Rack Rooms Cable Vault & Tunnels 1, 2, 3, 4 4

Table R-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building El 274' Instrument Rack Rooms Cable Vault & Tunnels 1, 2, 3, 4 4

Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building El 274' Instrument Rack Rooms Cable Vault & Tunnels 1, 2, 3, 4 4

RU1.3 Reading on any Table R-1 effluent radiation monitor

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

4 5

6 DEF 1

2 3

4 5

6 DEF 1

Notification of UNUSUAL EVENT North Anna Power Station Emergency Action Level Matrix Revision 10 NAPS 2

> 2 - 8

> 8 Table M-4 Tech. Spec. Coolant Activity Dose Rates Time > Shutdown (hrs) mR/hr/ml 0.80 0.50 0.30 2

> 2 - 8

> 8 Table M-4 Tech. Spec. Coolant Activity Dose Rates Time > Shutdown (hrs) mR/hr/ml 0.80 0.50 0.30 2

> 2 - 4

> 4 - 8

> 8 - 14

> 14 Table F-2 CTMT High Range Radiation Monitor Barrier Thresholds RM-RMS-165/166(265/266)

Time > Shutdown (hrs)

Fuel Clad Loss (R/hr) 125 85 45 20 10 RCS Loss (R/hr) 5 5

5 5

5 CTMT Potential Loss (R/hr) 500 340 180 80 40 2

> 2 - 4

> 4 - 8

> 8 - 14

> 14 Table F-2 CTMT High Range Radiation Monitor Barrier Thresholds RM-RMS-165/166(265/266)

Time > Shutdown (hrs)

Fuel Clad Loss (R/hr) 125 85 45 20 10 RCS Loss (R/hr) 5 5

5 5

5 CTMT Potential Loss (R/hr) 500 340 180 80 40 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 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/SEM Table M-6 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/SEM Table M-6 Hazardous Events Automatic turbine runback > 25% thermal reactor power Electrical load rejection > 25% full electrical load Reactor Trip SI Actuation Table M-3 Significant Transients Automatic turbine runback > 25% thermal reactor power Electrical load rejection > 25% full electrical load Reactor Trip SI Actuation Table M-3 Significant Transients Reactor power RCS level RCS pressure Core exit TC temperature Level in at least one SG Auxiliary feedwater flow to at least one SG Table M-2 Safety System Parameters Reactor power RCS level RCS pressure Core exit TC temperature Level in at least one SG Auxiliary feedwater flow to at least one SG Table M-2 Safety System Parameters Table M-1 AC Power Supplies Offsite:

Unit 1 Transfer Bus D Transfer Bus F Station Bus 1B Station Bus 2B Unit 2 Transfer Bus E Transfer Bus F Station Bus 2C Station Bus 1A Onsite:

1(2)H EDG 1(2)J EDG AAC (SBO) Diesel Generator (if already aligned)

Table M-1 AC Power Supplies Offsite:

Unit 1 Transfer Bus D Transfer Bus F Station Bus 1B Station Bus 2B Unit 2 Transfer Bus E Transfer Bus F Station Bus 2C Station Bus 1A Onsite:

1(2)H EDG 1(2)J EDG AAC (SBO) Diesel Generator (if already aligned) 2

> 2 - 8

> 8 Table F-3 FC Loss Coolant Activity Dose Rates Time > Shutdown (hrs) mR/hr/ml 15.0 8.0 3.0 2

> 2 - 8

> 8 Table F-3 FC Loss Coolant Activity Dose Rates Time > Shutdown (hrs) mR/hr/ml 15.0 8.0 3.0 2

> 2 - 8

> 8 Table F-4 FC Loss RCS Sample Line Dose Rates Time > Shutdown (hrs)

R/hr 4.0 2.0 1.0 2

> 2 - 8

> 8 Table F-4 FC Loss RCS Sample Line Dose Rates Time > Shutdown (hrs)

R/hr 4.0 2.0 1.0 None None Vent Stack A VG-RI-179-1 or 2 2.6E+08 µCi/sec 2.6E+07 µCi/sec 2.6E+06 µCi/sec 2.6E+05 µCi/sec Table R-1 Gaseous Effluent Monitor Classification Thresholds Release Point & Monitor GE SAE Alert NOUE Vent Stack B VG-RI-180-1 or 2 2.0E+08 µCi/sec 2.0E+07 µCi/sec 2.0E+06 µCi/sec 2.0E+05 µCi/sec Process Vent GW-RI-178-1 or 2 3.5E+08 µCi/sec 3.5E+07 µCi/sec 3.5E+06 µCi/sec 3.5E+05 µCi/sec Vent Stack A VG-RI-179-1 or 2 2.6E+08 µCi/sec 2.6E+07 µCi/sec 2.6E+06 µCi/sec 2.6E+05 µCi/sec Table R-1 Gaseous Effluent Monitor Classification Thresholds Release Point & Monitor GE SAE Alert NOUE Vent Stack B VG-RI-180-1 or 2 2.0E+08 µCi/sec 2.0E+07 µCi/sec 2.0E+06 µCi/sec 2.0E+05 µCi/sec Process Vent GW-RI-178-1 or 2 3.5E+08 µCi/sec 3.5E+07 µCi/sec 3.5E+06 µCi/sec 3.5E+05 µCi/sec Sample analysis indicates that a reactor coolant activity value is > any of the following Technical Specification 3.4.16 limits:

Dose equivalent I-131 > 1.0 µCi/gm for > 48 hrs

Dose equivalent I-131 > 60 µCi/gm

Dose equivalent Xe-133 > 197 µCi/gm for > 48 hrs NOTES Note 1:

The SEM should declare the event promptly upon determining that the 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

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 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 11: One full train of containment depressurization equipment consists of one Quench Spray (QS) System and one Recirculation Spray (RS) System from either train operating together.

Note 7:

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

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 8:

A manual trip action is any operator action, or set of actions from the control room, 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:

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

Note 10: 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.

Note 12: If an RCS heat removal system is in operation within the applicable Table C-5 heat-up duration and RCS temperature is being reduced, the EAL is not applicable.

Note 13: A Reactor Containment fire alarm is considered VALID upon receipt of multiple (more than one) fire zone alarms.

Note 14: For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition provided it can be aligned within the 15 minute classification criteria.

NOTES Note 1:

The SEM should declare the event promptly upon determining that the 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

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 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 11: One full train of containment depressurization equipment consists of one Quench Spray (QS) System and one Recirculation Spray (RS) System from either train operating together.

Note 7:

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

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 8:

A manual trip action is any operator action, or set of actions from the control room, 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:

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

Note 10: 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.

Note 12: If an RCS heat removal system is in operation within the applicable Table C-5 heat-up duration and RCS temperature is being reduced, the EAL is not applicable.

Note 13: A Reactor Containment fire alarm is considered VALID upon receipt of multiple (more than one) fire zone alarms.

Note 14: For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition provided it can be aligned within the 15 minute classification criteria.

Table E-1 ISFSI Cask Surface Dose Rate Limits TN-32

116 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask

1,600 mrem/hr at the front bird screen

4 mrem/hr at the door centerline

4 mrem/hr at the end shield wall exterior HSM-H TN-32B HBU

192 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask EOS-HSM

50 mrem/hr average over the front face

20 mrem/hr at the door centerline

10 mrem/hr average at the end shield wall exterior Table E-1 ISFSI Cask Surface Dose Rate Limits TN-32

116 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask

1,600 mrem/hr at the front bird screen

4 mrem/hr at the door centerline

4 mrem/hr at the end shield wall exterior HSM-H TN-32B HBU

192 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask EOS-HSM

50 mrem/hr average over the front face

20 mrem/hr at the door centerline

10 mrem/hr average at the end shield wall exterior

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT C

Cold S/D

/ Refuel System Malfunct.

1 RCS Level 6

Hazardous Event Affecting Safety Systems CU1.1 UNPLANNED loss of reactor coolant results in RCS water level < a required lower limit for 15 min. (Note 1)

CA1.1 RCS water level < minimum required for continued RHR pump operation CU5 Loss of all onsite or offsite communications capabilities CU5.1 Loss of all Table C-6 onsite communication methods OR Loss of all Table C-6 State and local agency communication methods OR Loss of all Table C-6 NRC communication methods CU3 UNPLANNED increase in RCS temperature CU3.1 CU4 Loss of vital DC power for 15 minutes or longer CU4.1 Indicated voltage is < 105 VDC on required vital 125 VDC battery buses for 15 min. (Note 1)

CG1 Loss of RCS inventory affecting fuel clad integrity with containment challenged CS1 Loss of RCS inventory affecting core decay heat removal capability CS1.1 CG1.1 CA2 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer CA2.1 CU2 Loss of all but one AC power source to emergency buses for 15 minutes or longer CU2.1 UNPLANNED increase in RCS temperature to > 200°F EAL MATRIX - COLD CONDITIONS (RCS 200°F)

AC power capability, Table C-4, to Unit 1(2) 4160V emergency buses H and J 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 Loss of all offsite and all onsite AC power to Unit 1(2) 4160V emergency buses H and J for 15 min. (Notes 1, 14)

CA1 Significant loss of RCS inventory CU1 UNPLANNED loss of RCS inventory Loss of all RCS temperature and RCS water level indication for 15 min. (Note 1)

None 2

Loss of Emergency AC Power 3

RCS Temp.

CU3.2 None None 4

Loss of Vital DC Power 5

Loss of Commun.

CA6 Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode CA6.1 The occurrence of any Table C-7 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode

Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)

None None None None CU1.2 RCS water level cannot be monitored AND EITHER:

UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory

Visual observation of UNISOLABLE RCS leakage CA1.2 RCS water level cannot be monitored for 15 min. (Note 1)

AND EITHER

UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory

Visual observation of UNISOLABLE RCS leakage With CONTAINMENT CLOSURE not established, any confirmed loss of inventory indication, Table C-2, with RVLIS full range < 62%

Any confirmed loss of inventory indication, Table C-2, with RVLIS full range < 61% for 30 min. (Note 1)

AND Any Containment Challenge indication, Table C-3 CA3 Inability to maintain plant in cold shutdown CA3.1 UNPLANNED increase in RCS temperature to > 200°F for

> Table C-5 duration (Notes 1, 12)

OR UNPLANNED RCS pressure increase > 10 psi (does not apply to solid plant conditions)

None None 5

6 5

6 DEF 5

6 5

6 DEF 5

6 5

6 NAPS CS1.3 RCS level cannot be monitored for 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

UNPLANNED rise in any Table C-1 sump or tank level of sufficient magnitude to indicate core uncovery

Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery

Any containment area radiation monitor reading

> 3 R/hr (Refueling Mode)

Erratic source range/excore monitor indications 5

6 5

6 CG1.2 RCS level cannot be monitored for 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

UNPLANNED rise in any Table C-1 sump or tank level of sufficient magnitude to indicate core uncovery

Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery

Any containment area radiation monitor reading

> 3 R/hr (Refueling Mode)

Erratic source range/excore monitor indications AND Any Containment Challenge indication, Table C-3 5

6 5

6 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT R

Abnormal Rad Levels/

Rad Effluent 1

Rad Effluent 2

Irradiated Fuel Events 1

Security 2

Seismic Event 3

Natural or Tech.

Hazard 6

Control Room Evacuation 7

SEM Judgment RG1 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem adult thyroid CDE RS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem adult thyroid CDE RA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem adult thyroid CDE RA2 Significant lowering of water level above, or damage to, irradiated fuel RU2 UNPLANNED loss of water level above irradiated fuel RU2.1 RA2.2 Damage to irradiated fuel resulting in a release of radioactivity AND

VALID Hi-Hi alarm on any of the following radiation monitors:

- RM-RMS-152 New Fuel Storage Areaa

- RM-RMS-153 Fuel Pit Bridge

- RM-RMS-162 (262) Manipulator Crane Area (Refueling Mode)

- RM-RMS-163 (263) Reactor Containment Area

- RM-RMS-159 (259) Containment Particulate

- RM-RMS-160 (260) Containment Area Gas

VALID high alarm on VG-RI-180-1 Vent Stack B Normal Range RA2.1 IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown RA3.1 Dose rates > 15 mR/hr in EITHER of the following:

Control Room Central Alarm Station (by survey)

HU2 Seismic event greater than OBE levels HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by NAPS Security Shift Supervisor OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat HA1.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by NAPS Security Shift Supervisor OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site HS1 HOSTILE ACTION within the PLANT PROTECTED AREA HS1.1 A HOSTILE ACTION is occurring or has occurred within the PLANT PROTECTED AREA as reported by NAPS Security Shift Supervisor HA6.1 An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel HS6 Inability to control a key safety function from outside the Control Room HS6.1 An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel AND Control of any of the following key safety functions is not re-established within 15 min. of the last licensed operator leaving the Control Room (Note 1):

Reactivity (modes 1, 2 and 3 only)

Core cooling RCS heat removal HU7.1 Other conditions exist which, in the judgment of the SEM, 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.

HA7.1 Other conditions exist which, in the judgment of the SEM, 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.

HS7.1 Other conditions exist which in the judgment of the SEM 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.

HG7.1 Other conditions exist which in the judgment of the SEM 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.

HG7 Other conditions exist which in the judgment of the SEM warrant declaration of a General Emergency HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes HU1 Confirmed SECURITY CONDITION or threat HA6 Control Room evacuation resulting in transfer of plant control to alternate locations HA7 Other conditions exist that in the judgment of the SEM warrant declaration of an Alert HU7 Other conditions existing that in the judgment of the SEM warrant declaration of a NOUE HS7 Other conditions existing that in the judgment of the SEM warrant declaration of a Site Area Emergency None None None None None 3

Area Radiation Levels Modes:

None None RA1.1 RA1.3 Reading on any Table R-1 effluent radiation monitor

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

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 adult thyroid CDE > 50 mrem for 60 min. of inhalation.

(Notes 1, 2)

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 adult thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

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 adult thyroid CDE > 500 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 > 1,000 mrem TEDE or 5,000 mrem adult 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 > 1,000 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate adult thyroid CDE > 5,000 mrem for 60 min. of inhalation.

(Notes 1, 2)

Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses

> 10 mrem TEDE or 50 mrem adult thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

None None None HU2.1 1

Confinement Boundary RA1.2 Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem adult thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

RA2.3 Lowering of spent fuel pool level to 10 ft. (Level 2) on 1-FC-LI-105-1, 2 or 2A Spent Fuel Pit Wide Range Level RS2.1 Lowering of spent fuel pool level to 1 ft. (Level 3) on 1-FC-LI-105-1, 2 or 2A Spent Fuel Pit Wide Range Level RS2 Spent fuel pool level at the top of the fuel racks RG2.1 Spent fuel pool level cannot be restored to at least 1 ft.

(Level 3) on 1-FC-LI-105-1, 2 or 2A Spent Fuel Pit Wide Range Level for 60 min. (Note 1)

RG2 Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer RA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 room or area (Note 5)

HU3 Hazardous event HU3.1 A tornado strike within the PLANT PROTECTED AREA HU3.2 Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode HU3.3 Movement of personnel within the PLANT PROTECTED AREA is IMPEDED due to an event external to the PLANT PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

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)

HU4 FIRE potentially degrading the level of safety of the plant 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 in any Table H-1 area 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 (excluding Reactor Containment)

AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Notes 1, 13)

HU4.3 A FIRE within the plant PLANT PROTECTED AREA or ISFSI 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 or ISFSI Protected Area that requires an offsite fire department to assist with extinguishment 4

Fire HA5 Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) 5 Hazardous Gases None None None None None Prepared for Dominion Energy by Operations Support Services, Inc. 7-18-21 1

2 3

4 5

6 DEF Power Operation Defueled 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following:

Spent Fuel Pit Lo Level (1E-C6) alarm

Report of lowering level in refueling cavity or SFP

Loss of SFP Cooling suction flow AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:

RM-RMS-152 New Fuel Storage Area

RM-RMS-153 Fuel Pit Bridge

RM-RMS-162 (262) Manipulator Crane Area (Refueling Mode)

RM-RMS-163 (263) Reactor Containment Area 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

Seismic event > OBE (0.06g horizontal or 0.04g vertical) as indicated by OBE EXCEEDED indicator illuminated on the SYSCOM Network Control Center (NCC)

None 2

3 4

5 6

1 2

3 4

5 6

1 H

Hazards

[Refer to CA6.1 or MA9.1 for potential escalation due to a seismic event]

[Refer to CA6.1 or MA9.1 for potential escalation due to a hazardous event]

[Refer to CA6.1 or MA9.1 for potential escalation due to fire]

NAPS North Anna Power Station Emergency Action Level Matrix Revision 10 None E

ISFSI EU1 Damage to a loaded cask CONFINEMENT BOUNDARY EU1.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any Table E-1 limit 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

NAPS 2

3 4

5 6

DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

Startup In service Standpipe and Ultrasonic level bottomed out Lowering RVLIS level trend RHR pump amp fluctuations Table C-2 Inventory Loss Confirmatory Indications In service Standpipe and Ultrasonic level bottomed out Lowering RVLIS level trend RHR pump amp fluctuations Table C-2 Inventory Loss Confirmatory Indications CS1.2 With CONTAINMENT CLOSURE established, any confirmed loss of inventory indication, Table C-2, with RVLIS full range < 61%

5 6

None None None Cold Shutdown 200°F Refueling Hot Standby 350°F Hot Shutdown

< 350°F Cable Vaults & Tunnels Emergency Switchgear Rooms Emergency Diesel Generator Rooms Reactor Containment Quench Spray Pump Houses Safeguards Area Main Steam Valve House Cable Spreading Rooms Control Room CR Chiller Rooms Auxiliary / Fuel / Decontamination Buildings Fuel Oil Pump House Room A or B Service Water Pump House and Valve House Intake Structure Control House Auxiliary Service Water Pump House Auxiliary Feedwater Pump House Turbine Building Table H-1 NAPS Fire Areas Cable Vaults & Tunnels Emergency Switchgear Rooms Emergency Diesel Generator Rooms Reactor Containment Quench Spray Pump Houses Safeguards Area Main Steam Valve House Cable Spreading Rooms Control Room CR Chiller Rooms Auxiliary / Fuel / Decontamination Buildings Fuel Oil Pump House Room A or B Service Water Pump House and Valve House Intake Structure Control House Auxiliary Service Water Pump House Auxiliary Feedwater Pump House Turbine Building Table H-1 NAPS Fire Areas Table R-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building El 274' Instrument Rack Rooms Cable Vault & Tunnels 1, 2, 3, 4 4

Table R-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building El 274' Instrument Rack Rooms Cable Vault & Tunnels 1, 2, 3, 4 4

Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building El 274' Instrument Rack Rooms Cable Vault & Tunnels 1, 2, 3, 4 4

None RU1 Release of gaseous or liquid radioactivity greater than 2 times the allocated ODCM limits for 60 minutes or longer RU1.1 Reading on SW-RM-130(230) CW Discharge Tunnel radiation monitor > 2 x the Hi-Hi setpoint for 60 min.

(Notes 1, 2, 3) 2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

RU1.2 Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x the allocated ODCM limits for 60 min. (Notes 1, 2) 2 3

4 5

6 DEF 1

2 3

4 5

6 DEF 1

RU1.3 Reading on any Table R-1 effluent radiation monitor

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

4 5

6 DEF 1

2 3

4 5

6 DEF 1

Notification of UNUSUAL EVENT Notification of UNUSUAL EVENT None North Anna Power Station Emergency Action Level Matrix Revision 10 NAPS 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 Reactor Containment Sump Pressurizer Relief Tank (PRT)

Primary Drain Transfer Tank (PDTT)

Component Cooling (CC) Surge Tank Refueling Water Storage Tank (RWST)

Table C-1 Sumps/Tanks Reactor Containment Sump Pressurizer Relief Tank (PRT)

Primary Drain Transfer Tank (PDTT)

Component Cooling (CC) Surge Tank Refueling Water Storage Tank (RWST)

Table C-1 Sumps/Tanks CONTAINMENT CLOSURE not established (Note 6)

CTMT hydrogen concentration 4%

UNPLANNED rise in CTMT pressure Table C-3 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6)

CTMT hydrogen concentration 4%

UNPLANNED rise in CTMT pressure Table C-3 Containment Challenge Indications Table C-4 AC Power Supplies Offsite:

Unit 1 Transfer Bus D Transfer Bus F Station Bus 1B Station Bus 2B Unit 2 Transfer Bus E Transfer Bus F Station Bus 2C Station Bus 1A Onsite:

1(2)H EDG 1(2)J EDG AAC (SBO) Diesel Generator (if already aligned)

Table C-4 AC Power Supplies Offsite:

Unit 1 Transfer Bus D Transfer Bus F Station Bus 1B Station Bus 2B Unit 2 Transfer Bus E Transfer Bus F Station Bus 2C Station Bus 1A Onsite:

1(2)H EDG 1(2)J EDG AAC (SBO) Diesel Generator (if already aligned)

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/SEM Table C-7 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/SEM Table C-7 Hazardous Events Vent Stack A VG-RI-179-1 or 2 2.6E+08 µCi/sec 2.6E+07 µCi/sec 2.6E+06 µCi/sec 2.6E+05 µCi/sec Table R-1 Gaseous Effluent Monitor Classification Thresholds Release Point & Monitor GE SAE Alert NOUE Vent Stack B VG-RI-180-1 or 2 2.0E+08 µCi/sec 2.0E+07 µCi/sec 2.0E+06 µCi/sec 2.0E+05 µCi/sec Process Vent GW-RI-178-1 or 2 3.5E+08 µCi/sec 3.5E+07 µCi/sec 3.5E+06 µCi/sec 3.5E+05 µCi/sec Vent Stack A VG-RI-179-1 or 2 2.6E+08 µCi/sec 2.6E+07 µCi/sec 2.6E+06 µCi/sec 2.6E+05 µCi/sec Table R-1 Gaseous Effluent Monitor Classification Thresholds Release Point & Monitor GE SAE Alert NOUE Vent Stack B VG-RI-180-1 or 2 2.0E+08 µCi/sec 2.0E+07 µCi/sec 2.0E+06 µCi/sec 2.0E+05 µCi/sec Process Vent GW-RI-178-1 or 2 3.5E+08 µCi/sec 3.5E+07 µCi/sec 3.5E+06 µCi/sec 3.5E+05 µCi/sec Intact AND not Reduced (decreased) Iinventory Table C-5 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-Up Duration Not intact OR Reduced (deceased) Inventory Established Not Established 60 minutes 20 minutes 0 minutes Intact AND not Reduced (decreased) Iinventory Table C-5 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-Up Duration Not intact OR Reduced (deceased) Inventory Established Not Established 60 minutes 20 minutes 0 minutes Table C-6 Communications Methods System Onsite State/

Local NRC Radio Communications System X

Public Address and Intercom System X

Private Branch Telephone Exchange (PBX)

X X

X Sound Powered Telephone System X

Commercial Telephone System X

X Enterprise Transport Network X

DEENS X

Dedicated NRC Communications X

Table C-6 Communications Methods System Onsite State/

Local NRC Radio Communications System X

Public Address and Intercom System X

Private Branch Telephone Exchange (PBX)

X X

X Sound Powered Telephone System X

Commercial Telephone System X

X Enterprise Transport Network X

DEENS X

Dedicated NRC Communications X

NOTES Note 1:

The SEM should declare the event promptly upon determining that the 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

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 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 11: One full train of containment depressurization equipment consists of one Quench Spray (QS) System and one Recirculation Spray (RS) System from either train operating together.

Note 7:

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

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 8:

A manual trip action is any operator action, or set of actions from the control room, 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:

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

Note 10: 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.

Note 12: If an RCS heat removal system is in operation within the applicable Table C-5 heat-up duration and RCS temperature is being reduced, the EAL is not applicable.

Note 13: A Reactor Containment fire alarm is considered VALID upon receipt of multiple (more than one) fire zone alarms.

Note 14: For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition provided it can be aligned within the 15 minute classification criteria.

NOTES Note 1:

The SEM should declare the event promptly upon determining that the 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 due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

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 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 11: One full train of containment depressurization equipment consists of one Quench Spray (QS) System and one Recirculation Spray (RS) System from either train operating together.

Note 7:

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

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 8:

A manual trip action is any operator action, or set of actions from the control room, 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:

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

Note 10: 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.

Note 12: If an RCS heat removal system is in operation within the applicable Table C-5 heat-up duration and RCS temperature is being reduced, the EAL is not applicable.

Note 13: A Reactor Containment fire alarm is considered VALID upon receipt of multiple (more than one) fire zone alarms.

Note 14: For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition provided it can be aligned within the 15 minute classification criteria.

Table E-1 ISFSI Cask Surface Dose Rate Limits TN-32

116 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask

1,600 mrem/hr at the front bird screen

4 mrem/hr at the door centerline

4 mrem/hr at the end shield wall exterior HSM-H TN-32B HBU

192 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask EOS-HSM

50 mrem/hr average over the front face

20 mrem/hr at the door centerline

10 mrem/hr average at the end shield wall exterior Table E-1 ISFSI Cask Surface Dose Rate Limits TN-32

116 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask

1,600 mrem/hr at the front bird screen

4 mrem/hr at the door centerline

4 mrem/hr at the end shield wall exterior HSM-H TN-32B HBU

192 mrem/hr (neutron + gamma) average on top of the cask

436 mrem/hr (neutron + gamma) average on the side of the cask EOS-HSM

50 mrem/hr average over the front face

20 mrem/hr at the door centerline

10 mrem/hr average at the end shield wall exterior

AC SourcesOperating 3.8.1 North Anna Units 1 and 2 3.8.1-1 Amendments 231/212 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC SourcesOperating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System;

b. Two emergency diesel generators (EDGs) capable of supplying the onsite Class 1E power distribution subsystem(s);

c. One qualified circuit between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System and one EDG capable of supplying the onsite Class 1E AC power distribution subsystem on the other unit for each required shared component; and

d. Required sequencing timing relays.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One LCO 3.8.1.a offsite circuit inoperable.

A.1 Perform SR 3.8.1.1 for required OPERABLE offsite circuit(s).

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND (continued)

AC SourcesOperating 3.8.1 North Anna Units 1 and 2 3.8.1-2 Amendments 231/212 CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2 Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of no offsite power to one train concurrent with inoperability of redundant required feature(s)

AND A.3 Restore offsite circuit to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND 17 days from discovery of failure to meet LCO B.

One LCO 3.8.1.b EDG inoperable.

B.1 Perform SR 3.8.1.1 for the required offsite circuits.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required feature(s) supported by the inoperable EDG inoperable when its required redundant feature(s) is inoperable.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from discovery of Condition B concurrent with inoperability of redundant required feature(s)

AND (continued)

ACTIONS

AC SourcesOperating 3.8.1 North Anna Units 1 and 2 3.8.1-3 Amendments 231/212 CONDITION REQUIRED ACTION COMPLETION TIME B.

(continued)

B.3.1 Determine OPERABLE LCO 3.8.1.b EDG is not inoperable due to common cause failure.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR B.3.2 Perform SR 3.8.1.2 for OPERABLE LCO 3.8.1.b EDG.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND B.4 Restore EDG to OPERABLE status.

14 days AND 17 days from discovery of failure to meet LCO C.

NOTE----------

Only applicable if Alternate AC (AAC) diesel generator (DG) or one or more EDG on the other unit is inoperable.

One LCO 3.8.1.b EDG inoperable.

C.1.1 Restore inoperable AAC DG to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND C.1.2 Restore inoperable EDG(s) on other unit to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR C.2 Restore EDG to OPERABLE status.

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

AC SourcesOperating 3.8.1 North Anna Units 1 and 2 3.8.1-4 Amendments 231/212 CONDITION REQUIRED ACTION COMPLETION TIME D.

NOTE----------

Separate Condition entry is allowed for each offsite circuit.

One or more required LCO 3.8.1.c offsite circuit(s) inoperable.

D.1 Perform SR 3.8.1.1 for required OPERABLE offsite circuit(s).

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND D.2 Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of no offsite power to a train concurrent with inoperability of redundant required feature(s)

AND D.3 Declare associated shared component inoperable.

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

AC SourcesOperating 3.8.1 North Anna Units 1 and 2 3.8.1-5 Amendments 231/212 CONDITION REQUIRED ACTION COMPLETION TIME E.

One required LCO 3.8.1.c EDG inoperable.

E.1 Perform SR 3.8.1.1 for required offsite circuit(s).

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND E.2 Declare required feature(s) supported by the inoperable EDG inoperable when its redundant required feature(s) is inoperable.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from discovery of Condition E concurrent with inoperability of redundant required feature(s)

AND E.3 Declare associated shared component inoperable.

14 days F.

NOTE----------

Only applicable if one or more LCO 3.8.1.b EDG(s) or AAC DG is inoperable.

One required LCO 3.8.1.c EDG inoperable.

F.1.1 Restore inoperable AAC DG to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND F.1.2 Restore inoperable LCO 3.8.1.b EDG (s) to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR F.2 Declare associated shared component inoperable.

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

AC SourcesOperating 3.8.1 North Anna Units 1 and 2 3.8.1-6 Amendments 231/212 CONDITION REQUIRED ACTION COMPLETION TIME G.

Two LCO 3.8.1.a offsite circuits inoperable.

G.1 Declare required feature(s) inoperable when its redundant required feature(s) is inoperable.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from discovery of Condition G concurrent with inoperability of redundant required features AND G.2 Restore one offsite circuit to OPERABLE status.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months H.

One LCO 3.8.1.a offsite circuit inoperable.

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

Enter applicable Conditions and Required Actions of LCO 3.8.9, Distribution SystemsOperating, when Condition H is entered with no AC power source to any train.

AND One LCO 3.8.1.b EDG inoperable.

H.1 Restore offsite circuit to OPERABLE status.

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

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I.

Two LCO 3.8.1.b EDGs inoperable.

I.1 Restore one EDG to OPERABLE status.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months J.

Two required LCO 3.8.1.c EDGs inoperable.

J.1 Declare associated shared components inoperable.

Immediately ACTIONS

AC SourcesOperating 3.8.1 North Anna Units 1 and 2 3.8.1-7 Amendments 231/212 CONDITION REQUIRED ACTION COMPLETION TIME K.

NOTE----------

Separate Condition entry is allowed for each sequencing timing relay.

One or more required sequencing timing relay(s) inoperable.

K.1 Enter appropriate Conditions and Required Actions for any component made inoperable by inoperable sequencing timing relay(s).

Immediately AND K.2.1 Place the component(s) with the inoperable sequencing timing relay in a condition where it cannot be automatically loaded to associated emergency electrical bus.

Immediately OR K.2.2 Declare the associated EDG inoperable.

Immediately L.

Required Action and associated Completion Time of Condition A, B, C, G, H, or I not met.

L.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND L.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months M.

Three or more of LCO 3.8.1.a and LCO 3.8.1.b AC sources inoperable.

M.1 Enter LCO 3.0.3.

Immediately ACTIONS

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 57 OF 238 6.0 INSTRUCTIONS 6.1 General This Section presents required notifications and reports on the basis of initiating mechanisms.

Non-scheduled initiating mechanisms are those that cannot be, or are not easily, pre-scheduled.

Non-scheduled mechanisms are further classified according to event or condition, or according to time limitations for fulfilling the required action, or both. Scheduled reports are those whose completion can be pre-scheduled. Subsections 6.2, Non-Scheduled Notifications and Reports, and 6.4, Scheduled Reports, summarize requirements and implementation processes for both groups. Subsections 6.5 through 6.30 provide the details for each requirement.

NOTE: PI-AA-200, Corrective Action, establishes responsibilities and processing requirements for initiating and obtaining determinations of reportability for most non-periodic events. [Commitment 3.2.2]

6.1.1 Notifications

a. Voice or fax notifications or confirmations by dialable telephone, to individuals or organizations outside Dominion Energy, shall be to the numbers listed in the:

Applicable Emergency Plan Implementing Procedure

Emergency Telephone Directory Voice notification numbers that may not be included in the above listed documents are:

NRC Director, Spent Fuel Project Office(301) 415-8500

National Response Center (EPA and U.S. Coast Guard)(800) 424-8802

U.S. Coast Guard(804) 441-3314 (Surry)

FERC Regional Engineer(904) 415-6897

Department of Transportation (DOT)(800) 424-8802 or (202) 426-2675

Office of Pesticides & Toxic Substances(215) 597-8598

NERC EOP-004 Event Reporting(404) 446-9780

SERC EOP-004 Event Reporting(704) 405-8700

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 58 OF 238 6.1.1 Notifications (continued)

State Department of Environmental Quality Air/Water/Waste Regional Office (703) 583-3800 or (after hours) DEM (800) 468-8892 (North Anna)

Air/Water/Waste Regional Office (804) 527-5020 or (after hours) DEM (800) 468-8892 (Surry)

State Corporation Commission(804) 371-9611

Area Director of Occupational Safety and Health Administration (OSHA)

(804) 371-2327

Chemical Review Safety Board(201) 261-7600 or report@csb.gov

Nuclear Electric Insurance Limited (NEIL) (877) 634-5911

American Nuclear Insurers (ANI)(860)-682-1301

Local County Administrator

Louisa County(540) 967-0401

Surry County(757) 294-5271

State Department of Emergency Management(804) 674-2400, ask for EOC Duty Officer

Virginia Emergency Response Council (VERC) 800-468-8892

Louisa Co. Local Emergency Planning Committee (LEPC) - 540-967-1234

Surry Co. Local Emergency Planning Committee (LEPC) - 757-294-5271 Fax numbers that may not be included in the above listed documents are:

NRC Operations Center(301) 816-5151

NRC Regional Office(404) 997-4900

State Department of Environmental Quality

Air/Water/Waste/Pollution Response Regional Office-(804) 527-5106 (Surry)

Air/Water/Waste/Pollution Response Regional Office-(703) 583-3821 (North Anna)

Nuclear Electric Insurance Limited(302) 888-3008

American Nuclear Insurers(860) 659-0002

NERC EOP-004 Event Reporting(404) 446-9770

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 59 OF 238

b. Notifications to other departments inside Dominion Energy for consideration of additional action(s) to be taken include Electric Environmental Services.

6.1.2 Reports

a. Individuals or organizations responsible for preparing a report shall collect, interpret, and ensure the accuracy and validity of information required for a report in accordance with this procedure and with applicable implementing procedures.

b. Individuals or organizations responsible for reviewing a report shall conduct a technical, administrative, and regulatory review, as appropriate.

c. Documents to be submitted to NRC shall be sent to:

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

d. Documents to be submitted to the NRC Regional Office shall be sent to:

USNRC Region II Marquis One Tower 245 Peachtree Center Avenue, NE, Suite 1200 Atlanta, GA 30303-1257

e. Documents to be submitted to the REIRS Project Manager shall be sent to:

REIRS Project Manager Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

f. Documents to be submitted to the Office of Nuclear Material Safety and Safeguards shall be sent to:

Director, Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

g. Documents to be submitted to the Division of Low-Level Waste Management and Decommissioning shall be sent to:

Director, Division of Low-Level Waste Management and Decommissioning U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 60 OF 238

h. Documents to be submitted to the FERC Regional Office shall be sent to:

Federal Energy Regulatory Commission Atlanta Regional Office 3700 Crestwood Parkway, NW, Suite 950 Duluth, GA 30096

i. Documents to be submitted to the Virginia Department of Emergency Management shall be sent to:

Virginia Department of Emergency Management 10501 Trade Court Richmond, VA 23236-3713

j. Documents to be submitted to the State Department of Environmental Quality shall be sent to:

Air Northern Virginia Regional Office (NVRO) 13901 Crown Court Woodbridge, VA 22193 (North Anna)

State Department of Environmental Quality (Air)

Arboretum 5, Suite 250 9210 Aboretum Parkway Richmond, VA 23236 (Surry)

Water Northern Virginia Regional Office (NVRO) 13901 Crown Court Woodbridge, VA 22193 (North Anna)

Piedmont Regional Office 4949-A Cox Road Glen Allen, VA 2323060 (Surry)

k. Documents to be submitted to the Virginia Department of Health shall be sent to:

Thomas Jefferson Health District 1138 Rose Hill Drive P.O. Box 7546 Charlottesville, VA 22906 (North Anna)

Virginia Department of Health Southeast Virginia Regional Office 5700 Thurston, Suite 203 Virginia Beach, VA 23455 (Surry)

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 61 OF 238

l. Documents to be submitted to American Nuclear Insurers shall be sent to:

American Nuclear Insurers 95 Glastonbury Boulevard, Suite 300 Glastonbury, CT 06033

m. Documents to be submitted to Nuclear Electric Insurance Limited shall be sent to:

Nuclear Electric Insurance Limited Manufacturers Hanover Plaza 1201 Market Street, Suite 1200 Wilmington, DE 19801

n. Documents to be submitted to the South Carolina Department of Health and Environmental Control shall be sent to:

South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

o. Documents submitted to the local County Administrator shall be sent to:

Louisa County Administrator Surry County Administrator P.O. Box 160 P. O. Box 65 Louisa, VA 23093 Surry, VA 23883

p. Documents to be submitted to NERC for EOP-004, Event Reporting, shall be sent electronically to:

systemawareness@nerc.net

q. Documents to be submitted to SERC for EOP-004, Event Reporting, shall be sent electronically to:

Reporting_line_sit@list-serc1.org

r. Documents to be submitted to PJM for EOP-004, Event Reporting, shall be sent electronically to:

dispsup@pjm.com

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 62 OF 238 6.2 Non-Scheduled Notifications and Reports NOTE: Events involving Physical Security are evaluated for reportability by Protection Services in accordance with SY-AA-105, Security Safeguards/ Cyber Event Reportability.

NOTE: Reportability determinations for items included in Step 6.2.1 are initiated and processed in accordance with PI-AA-200, Corrective Action.

6.2.1 Critical, Significant, and Potentially Significant Events or Conditions NOTE: Notifications required by activation of the Emergency Action Plan for Lake Anna Dam are established and controlled by the Plan. However, see Step 6.3.4.a.5.

a. Emergency Plan ActivationSee Steps 6.3.2, 6.3.5, and 6.3.7.

NOTE: Operability/functionality (availability) is established by the controlling procedure (e.g., Technical Specifications, Station Administrative Procedure). Requirements in this procedure to report inoperable/nonfunctional systems or equipment generally rely on other procedures to establish the basis for determining operability/functionality.

b. Systems and Components

Reactor tripSee Steps 6.3.3, 6.3.4.a., 6.11.11, 6.28.1.a., and 6.28.2

Inoperable/nonfunctional (including unavailable or out of service) systems or componentsSee Steps 6.3.3, 6.7.2, 6.11.2, 6.11.11, 6.25.12.b., 6.25.15, 6.29.2, 6.29.3, 6.30.1, 6.30.2, 6.30.3, 6.30.4 and 6.30.6

Fire detection, suppression, or barrier inoperability/nonfunctionalitySee Steps 6.3.5.d., 6.3.6.a., 6.26.1 and 6.29.4

Defective systems or componentsSee Steps 6.3.4.1., 6.7.2, 6.11.2, 6.11.11 and 6.28.3.e.

Unacceptable containment leak rate test resultsSee Step 6.11.17

Significant changes in projected values of RTPTSSee Step 6.11.7

Reactor Vessel Overpressure Mitigating System is used to mitigate an RCS transientSee Step 6.25.12.a. (Surry)

Unscheduled outagesSee Steps 6.28.1.a. and 6.28.2

Dissolved gases in transformers exceed limitsSee Step 6.29.2

Conditions affecting the safety of Lake Anna Dam or its worksSee

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 63 OF 238 Steps 6.3.2.j. and 6.19.1

Planned removal from service, and restoration to service, of Lake Anna Dam safety devicesSee Step 6.19.8

c. Operating Limitations

Technical Specification safety limit exceededSee Steps 6.3.4.a.2.,

6.3.5.a.3., 6.11.2, 6.24.3 (North Anna), and 6.25.3 (Surry).

Limiting Condition for Operation not metSee Steps 6.3.4.1., 6.11.2, and 6.11.11.

Departure from license conditions or Technical Specifications permitted by 10 CFR 50.54(x)See Steps 6.3.3.a. and 6.11.11.

Excess oxygen in waste gas holdup systemSee Step 6.25.12.c. (Surry)

Excessive quadrant to average power tiltSee Step 6.25.12.d. (Surry)

d. Radiation or Exposure Events

Accidental criticalitySee Steps 6.3.3.b., 6.18.1 and 6.28.2

Personnel contaminationSee Steps 6.3.2.a.4., 6.6.4, 6.18.1 and 6.28.2

Radiation overexposuresSee Steps 6.3.2.a.4., 6.6.4, 6.18.1 and 6.28.2

Planned special exposuresSee Step 6.6.5

At receipt, contaminated or excessively radioactive packagesSee Step 6.3.2.c.

Radioactive effluent releasesSee Steps 6.3.2.a.4., 6.3.6.c., 6.6.4, 6.11.11, 6.11.16, 6.18.1, 6.27.2, 6.28.2, and 6.29.3

Radioactive materials transport incidentSee Steps 6.3.2.h. and 6.29.3

Twenty Four Hour NotificationSee Step 6.3.6.a.1.

Groundwater contaminationSee Step 6.3.4 and Subsection 6.32.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 64 OF 238

e. Security or Safeguards Events NOTE: Events involving Physical Security are evaluated for reportability by Protection Services in accordance with SY-AA-105, Security Safeguards/ Cyber Event Reportability.

Attempted or actual unauthorized entrySee Steps 6.3.3.f., 6.16.3 and 6.28.2

Acts, attempts, or threats to interrupt normal operationSee Steps 6.3.3.d., 6.16.3 and 6.28.2

Loss, theft, or attempted theft of special nuclear materialSee Steps 6.3.2.a.3.,

6.3.3.b., 6.3.3.d., 6.6.2.b., 6.16.3, 6.17.1 and 6.28.2

Involving byproduct, source, or special nuclear materialSee Steps 6.3.3.d.,

6.6.2, and 6.16.3

Attempted or actual introduction of contrabandSee Steps 6.3.3.h., 6.3.6.b., 6.8.1

Loss of shipment of special nuclear material or spent fuelSee Step 6.3.3.c.

Violations of requirements of NRC-approved physical security, guard training and qualification, and safeguard contingency plans

Cyber Security Event notificationsSee Steps 6.3.3, 6.3.4, 6.3.5, and 6.11.11.b.14.

f. Fitness for Duty Events

Significant Fitness for Duty eventsSee Steps 6.3.6.b., 6.8.1 and 6.28.2

NRC employee suspected to be unfit for dutySee Step 6.3.2.d.

Drug and Alcohol testing ErrorsSee Step 6.8.3.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 65 OF 238

g. Environmental Events

Toxic gas releasesSee Steps 6.3.6.c., 6.27.2.b. (North Anna) and 6.28.2.a.

Oil or hazardous material spills or releasesSee Steps 6.3.2.e., 6.3.2.f., 6.3.6.c.,

6.21.4, 6.27.2.b., 6.28.2.a., 6.28.3.l., and 6.28.3.n. (North Anna)

Smoke releases from StationSee Step 6.3.4.b.

Significant increase in nuisance organisms or conditions (North Anna)See Steps 6.3.6.c. and 6.27.2

Failure to comply with VPDES permit requirementsSee Steps 6.3.2.g., 6.3.6.f.,

6.3.6.e., 6.27.1.a. and 6.28.3.n.

Unplanned bypass of waste treatment facilitiesSee Steps 6.3.6.f. and 6.28.3.n.

Unpermitted, unusual, or extraordinary dischargeSee Steps 6.3.6.e.

and 6.28.3.n.

Unanticipated or emergency discharge of waste water or chemical substances See Steps 6.3.6.c. (North Anna), 6.3.6.e., 6.27.2.b. (North Anna), 6.28.2 and 6.28.3.n.

Bird of prey death or injury by electrocutionSee Step 6.23.4

Disturbance of an osprey nestSee Step 6.23.4

Excessive bird impactions (North Anna)See Step 6.27.2

Fish killsSee Steps 6.3.6.c., 6.27.2.b. and 6.28.2 (North Anna)

On-site plant or animal disease outbreaksSee Steps 6.3.6.c., 6.27.2.b.

and 6.28.2 (North Anna)

Mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973See Steps 6.3.6.c., 6.27.2 and 6.28.2.a. (North Anna)

Any air release, in any quantity, that results in one or more of the following outcomes:

Fatality - either to an employee or to general public Serious injury or illness resulting in death or impatient hospitalization to an employee or general public Substantial property damage or destruction to public or private property of at least $1,000,000. see Steps 6.3.2.k. and 6.3.5.e.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 66 OF 238

h. ISFSI-Unique Events

A defect in any spent fuel storage cask structure, system, or component important to safetySee Step 6.3.5.a.8.

A significant reduction in the effectiveness of any spent fuel storage cask confinement system during use of the storage caskSee Step 6.3.5.a.8.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 67 OF 238

i. Miscellaneous Events or Conditions

Special circumstances that may be considered media significantSee Steps 6.3.4.a.5., 6.12.3, and 6.28.2.a.

Unusual or unplanned occurrences that may be of concern to nearby residents See Step 6.28.2.a.

Station firesSee Steps 6.18.1, 6.28.2.a., 6.29.2 and 6.29.3

Demonstrations, picketing, civil disturbances, strikes, work stoppagesSee Steps 6.3.3.d., 6.3.4.a.5., and 6.28.2.a.

Earthquakes, storms, floods, forest or brush firesSee Steps 6.3.2.j., 6.11.11, 6.19.1 (North Anna), 6.28.2.a. and 6.29.3

Injuries or deathsSee Steps 6.3.2.h., 6.3.4.a.5., 6.3.5.c., 6.18.1 and 6.28.2.a.

Transportation of contaminated injured personSee Step 6.3.5.a.6.

Deaths or serious injuries at, or alleged to be related to, Lake Anna DamSee Steps 6.3.2.i., 6.3.5.c., 6.19.2 and 6.28.2.a. (North Anna)

Transport incidents involving radioactive or hazardous materialsSee Steps 6.3.2.h., 6.18.1, 6.22.2, and 6.29.3

Unanalyzed condition that significantly compromises Station safetySee Step 6.3.5.2.

Failure to notify NRC of planned removal or significant changes to equipment that controls amount of radioactivity in effluentsSee Step 6.3.6.d.

(North Anna)

Ambulance transport of personnel to an off-site medical facilitySee Step 6.28.2

Mishaps involving low-level waste formsSee Step 6.30.5

A failure to comply, potentially associated with a significant safety hazardSee Step 6.7.2

Nonreceipt of hazardous waste shipment manifest from receiverSee Steps 6.21.7.b. and 6.28.3.b.

Planned or emergency removal of asbestos or asbestos containing materialSee Step 6.28.3.b.

Actual or expected unavailability of licensed waste treatment operatorSee Step 6.28.3.m.

Pump and haul of bulk-storage-tank bottom watersSee Step 6.28.3.p. (Surry)

Operation of auxiliary boilerSee Step 6.28.3.f. (Surry)

Licensed material package effectiveness reduction or with safety-significant defectsSee Step 6.14.3

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 68 OF 238 6.2.2 Special Commitments; Administrative Matters

a. Outages and Refueling

OutagesSee Steps 6.28.1.a. and 6.28.1.b.

RefuelingSee Step 6.24.8

Removal of Reactor Vessel Material Surveillance Program couponsSee Step 6.11.15

Restart after refueling, fuel movement, license modification authorizing a power level increase, or Station modificationsSee Step 6.25.4 (Surry)

Inservice inspectionsSee Steps 6.24.4, 6.24.7.a. (North Anna), and 6.25.5 (Surry)

b. Legal & Commercial

1. Program & Procedure Changes

Changes to the security plans without prior NRC approvalSee Step 6.11.5.b.

Revisions to the Emergency Plan or implementing procedures without prior NRC approvalSee Step 6.11.5.c.

Changes to Chemical Test Program procedures.

Significant changes in the operation of equipment that controls the amount of radioactivity in effluents (North Anna)See Step 6.24.5.

Changes in discharge or management of pollutantsSee Step 6.28.3.j.

Significant changes from upstream or downstream conditions addressed in the North Anna Hydroelectric Project Emergency Action Plan (North Anna)

See Step 6.19.4

Decreased availability of private personnel or equipment to prevent or mitigate a worst-case oil releaseSee Step 6.28.3.l.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 69 OF 238

2. Station Changes

Major changes to radioactive liquid, gaseous, or solid waste treatment systemsSee Step 6.11.3

Introduction of an extremely hazardous substance in an amount greater than its threshold planning quantitySee Step 6.21.9

A change in type of product stored or handled at the Station for which an Safety Data Sheet (SDS) has not been submittedSee Step 6.28.3.l.

A substantial increase in the maximum oil storage capacity at the Station See Step 6.28.3.l.

3. Movement of Radioactive Materials

Shipment or receipt of SNMSee Steps 6.16.1, 6.16.2, 6.16.3, 6.16.4, and 6.17.3

First use of radioactive material packagingSee Step 6.14.2

Special Shipment of radioactive material - See Step 6.29.7

4. NRC Licences, Orders, & Inspections

Change in operator or senior operator statusSee Step 6.11.12

Receipt of NRC notices of violation that involve radiological working conditions, proposed impositions of civil penalty, orders for imposing requirements, orders modifying, suspending, revoking a license, orders imposing a civil penalty, and responses thereto. See Steps 6.5.1.e. and 6.5.1.f.

Issuance of an NRC shutdown orderSee Step 6.29.6

Issuance of Dominion Energy Annual ReportSee Steps 6.11.8 and 6.15.11

Five years before expiration of reactor operating licenseSee Step 6.11.5.f.

Three years before the predicted date that fracture toughness levels will no longer satisfy 10 CFR 50, App. G,Section IV.ASee Step 6.11.14.

Suspension or revocation of an NRC operating licenseSee Step 6.29.6

A change of licensee for the StationSee Step 6.28.3.l.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 70 OF 238

5. Permits, Orders, & Evaluations

Proposed changes to the VPDES permitSee Step 6.27.1.b. (North Anna) or Step 6.28.3.o. (Surry)

Changes or additions to the VPDES permit or State certificationSee Step 6.27.1.b.(North Anna) or Step 6.28.3.o. (Surry)

Stay of a VPDES permit or State certification appealSee Step 6.27.1.b.

(North Anna)

Modifications to Lake Anna Dam or its worksSee Step 6.19.3

Suspension from INPOSee Step 6.29.5

Classification as INPO Category 5See Step 6.29.5

6. Insurance & Financial

Material change in proof of financial protection or financial information previously filedSee Step 6.18.2

Expiration, renewal, or replacement of 10 CFR 140 financial protectionSee Step 6.18.3

Filing of Chapter 11 petition by or against any component of Dominion Energy ResourcesSee Step 6.11.5.g.

c. Individual Requests or Directives

Worker and former worker radiation exposure dataSee Steps 6.5.2 and 6.5.3.

Radiation overexposuresSee Step 6.5.4.

Terminating employees & workersSee Step 6.5.5.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 71 OF 238 6.3 Immediate to 72-Hour Notifications This subsection consolidates requirements for situations or events addressed by Subsections 6.5 through 6.30, for which notifications or reports are required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

6.3.1 General Requirements

a. When this subsection (Subsection 6.3) designates someone other than the Shift Manager or a member of Station management to notify a government agency, that person shall ensure the Shift Manager or a member of Station management is advised before making the notification. See also Step 6.3.4.a.5.

NOTE: Notifications for events that exceed an Emergency Action Level, as specified in EPIP-1.01, Emergency Manager Controlling Procedure, are controlled by EPIP-2.01, Notifications and Communications and EPIP-2.02, Notification of NRC. See also Steps 6.3.5 and 6.3.7. [10 CFR 50.72(a)(3), 10 CFR 50.72(c)(1), 10 CFR 50.72(c)(2)]

NOTE: When it is discovered that an event or condition had existed, but the basis for the emergency class no longer exists at the time of this discovery and no other reasons exist for an emergency declaration, then declaration of an emergency class is not required. See Step 6.3.3.i. for notification requirements.

NOTE: Environmental events occurring at Surry Dredged Material Management Area (DMMA) should be reported in a similar fashion as a station event.

b. For events reportable to the NRC Operations Center, the Shift Manager shall:

1. Complete NRC Form 361, Event Notification Worksheet.

2. Fax the Event Notification Worksheet to the NRC Operations Center. See Step 6.1.1.

3. Using the Emergency Notification System (ENS), verify that NRC received the fax.

4. Be prepared to read the entire contents of the Event Notification Worksheet to the NRC Operations Center officer.

5. Ensure the NRC Operations Center officer has a clear understanding of the issues, and that all questions regarding the notification have been answered.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 72 OF 238

6. If the ENS is nonfunctional, use commercial telephone service, other dedicated telephone service, or any other method that ensures the NRC Operations Center is notified as soon as practical. See Step 6.1.1. [10 CFR 50.72(a)(2)]

7. Maintain an open, continuous communications channel with the NRC Operations Center, when requested by NRC. [10 CFR 50.72(c)(3) & 10 CFR 73.71(a)(3)]

c. For events that are reportable in accordance with 10 CFR 50.72 and 10 CFR 72.75:

Immediately, the Shift Manager shall notify the Manager Nuclear Operations or the Operations Manager On Call, and the STA

Within one hour, the Manager Nuclear Operations or Operations Manager On Call shall notify the Site Vice President and the Plant Manager (Nuclear)

Within one hour, the STA shall notify the Director Nuclear Station Safety and Licensing

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the Director Nuclear Station Safety and Licensing (if absent, the Plant Manager (Nuclear)) shall notify the NRC Resident Inspector.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the Director Nuclear Station Safety and Licensing (if absent, the Plant Manager (Nuclear)) shall notify the Director, Nuclear Regulatory Affairs

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the Site Vice President, a Director, Manager Nuclear Operations, or Shift Manager shall notify the Senior Vice President Nuclear Operations

When notified, the Director, Nuclear Regulatory Affairs shall promptly notify appropriate corporate organizations, including Public Relations, Medical, Corporate Risk Management, and Power Supply, as applicable

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 73 OF 238 6.3.2 Immediate Notifications NOTE: Some conditions, indicated by See EPIP-1.01, may exceed an Emergency Action Level (EAL) as specified in EPIP-1.01, Emergency Manager Controlling Procedure.

If a condition exceeds an EAL, Emergency Plan Implementing Procedures (EPIPs) control State and Federal agency notifications. If an event or condition does not exceed an EAL, it may still be reportable in accordance with this procedure.

NOTE: Upon NRC request, the designated responsible person must maintain an open, continuous communications channel with the NRC Operations Center. [10 CFR 50.72(c)(3)]

a. The Shift Manager shall notify the NRC Operations Center via the ENS of:

1. Any further degradation in the level of safety of the plant or other worsening plant conditions, after telephone notifications to NRC as specified in Step 6.3.2 or Step 6.3.3. See EPIP-1.01. [10 CFR 50.72(c)(1)]

2. The results of ensuing evaluations or assessments of plant conditions, the effectiveness of response or protective measures taken, and information related to plant behavior that is not understood, after telephone notifications to NRC as specified in Step 6.3.2 or Step 6.3.3. [10 CFR 50.72(c)(2)]

3. Lost, stolen or missing licensed material in an aggregate quantity equal to or greater than 1,000 times the quantity specified in 10 CFR 20, Appendix C, under circumstances in which it appears persons in unrestricted areas could be exposed. See also Steps 6.6.2.b. and 6.6.2.c. [10 CFR 20.2201(a)(i)]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 74 OF 238 Table 1 Summary of Reporting Requirements for Non-Radiological Releases To the Environmenta

a. Step 6.3.2.f. explains releases to the environment when hazardous substances or hazardous wastes are involved.

For oil, releases includes spilling, leaking, pumping, pouring, emitting, emptying, discharging, injecting, escaping, leaching, or disposing. Oil releases solely within a workplace (an enclosed building with a concrete floor) are not subject to the RQ unless oil reaches a floor drain connected to a pathway to the environment. All outdoor releases are subject to the RQ.

Material Released To Amountb

b. RQ = reportable quantity as specified in VPAP-2202, Control of Chemicals and Hazardous Substances.

Report Toc, m

c. DEQ = State Department of Environmental Quality; NaRC = National Response Center*; LEPC = Local Emergency Planning Coordinator. *At Surry, if the NaRC is notified, the U.S. Coast Guard must also be notified. DEM (Department of Emergency Management (Emergency Operations Center)) is notified (instead of DEQ) on nights, weekends, and after hours. Environmental immediate notification is defined as soon as possible, but not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Phone numbers for the agencies are found in 6.1.1. Attachment 1, Oil or Hazardous Substance Release Report, should be used for spill information requested by the agencies. NRC notification is required within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of notifying any of these agencies.

See Oil Soil Outside containment facilitiesd

d. Oil tank dikes and transformer vaults are typical containment facilities. (See also Substep 6.3.2.e.)

> 25 gallons DEQ 6.3.2.e.

Solid surface Potential to reach soil or water Case basis State Waterse

e. Includes releases to storm drains or comparable conduits to state waters. See also 4.38, State Waters and Substep 6.3.2.f.

Any discernible amount NaRCl, DEQ Hazardous Substancef, g

f. Hazardous substances of concern to the Station are identified in VPAP-2202, Control of Chemicals and Hazardous Substances.

g. See Footnote 1. on page 78 if there is an on-site RQ release of a volatile substance.

Land Off-site RQ NaRC, DEQ, VERC, & LEPC 6.3.2.f.

6.3.2.h.h

h. 6.3.2.h. is applicable for transportation-related events.

Water On-sitei i.

An on-site hazardous material spill that, due to location, size, or substance properties, poses imminent or likely danger of an RQ release to the environment, must be reported to the same entities as offsite spills.

< RQ DEQ RQ NaRC, DEQ, & LEPC Off-site

< RQ DEQ RQ NaRC, DEQ Hazardous Wastej

j. Hazardous waste is defined in the Environmental Protection Plan.

k. If the oil spill that reaches navigable waters is: a) greater than 1000 gallons or b) the second spill that is greater than 42 gallons in a 12 month period, then contact Electric Environmental Services (EES). Additional reporting by EES to the EPA will be required. [Commitment 3.2.26]

l.

The National Response Center (NaRC) must be notified within 15 minutes of a discovery of an oil spill to state waters.

m. If any release occurs that causes radiation and/or gas, mist, solids, liquids to be released to the atmosphere which results in an injury to an employee or member of the public that requires hospitalization, or the death of an employee or member of the public, or property damage equal to or exceeding $1.0M; then a notification to the federal Chemical Safety Board (CSB) must be made. See also Step 6.3.2.k., Step 6.3.5.e., and ATTACHMENT 9.

Land On-sitei RQ NaRC & DEQ Off-site Any amount NaRC, DEQ, & LEPC Water On-sitei

< RQ DEQ RQ NaRC, DEQ, & LEPC Off-site Any amount NaRC, DEQ, & LEPC

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 75 OF 238 NOTE: The requirements of Step 6.3.2.a.4. do not apply to doses that result from planned special exposures, that are within the limits for planned special exposures, and that are reported in accordance with Step 6.6.5. [10 CFR 20.2202(e)]

4. Events that involve by-product, source, or special nuclear material possessed by Dominion Energy that may have caused or threatens to cause: [10 CFR 20.2202(a)]

An individual to receive:

- A total effective dose equivalent of 25 rems

- An eye dose equivalent of 75 rems

- A shallow-dose equivalent to the skin or extremities of 250 rads

Release of radioactive material inside or outside a restricted area, so that, if an individual had been present for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , they could have received an intake five times the occupational annual limit on intake If the event involves radiological overexposure, the DEM shall be notified as specified in Step 6.28.2. See also Step 6.6.3.c.

5. Upon declaration of an emergency as specified in the approved emergency plan regarding ISFSI events. [10 CFR 72.75(a)]

b. When informed by Security or Radiation Protection of events related to Category 1 and Category 2 radioactive material shipments (refer to Appendix A to 10 CFR Part 37 - Physical Protection of Category 1 and Category 2 Radioactive Materials), notify the NRC's Operations Center (301-816-5100) as soon as possible:

1. After notification of the local law enforcement agency (LLEA), upon discovery of any actual or attempted theft or diversion of a shipment, or any suspicious activity related to the shipment of Category 1 radioactive material. [10 CFR 37.81(c)]

2. Upon discovery of any actual or attempted theft or diversion of a shipment, or any suspicious activity related to the shipment, of a category 2 quantity of radioactive material. [10 CFR 37.81(d)]

3. Upon recovery of any lost or missing shipments of Category 1 or Category 2 quantities of radioactive material. [10 CFR 37(e) & (f)]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 76 OF 238

c. If:

Removable radioactive surface contamination exceeds the limits of 10 CFR 71.87(i) [10 CFR 20.1906(d)(1)]

or

External radiation levels exceed the limits of 10 CFR 71.47 [10 CFR 20.1906(d)(2)]

1. Radiological Protection shall notify Licensing Manager (Station) and the Shift Manager.

2. Radiological Protection or Licensing Manager (Station) shall notify (see Step 6.3.1.a.) the final delivering carrier and, by telephone and telegram, mailgram, or facsimile, the NRC Operations Center. See Step 6.1.1.

3. The notifier in Step 6.3.2.c.2. shall initiate a Condition Report as specified in PI-AA-200, including documentation of its notifications on the Condition Report.

d. If an NRC employee or NRC Contractor is believed to be under the influence of any substance or otherwise unfit for duty, the Fitness for Duty Administrator (Station) or a Station Management staff member shall notify (see Step 6.3.1.a.) the NRC Regional Administrator by telephone followed by written notification (e.g., e-mail or fax). If the Regional Administrator cannot be reach, notification shall be made to the NRC Operations Center. [10 CFR 26.77(c)]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 77 OF 238 NOTE: Use the online environmental incident flow charts for Virginia found at the following link: http://domnet.dominionnet.com/services/envsvc/Pages/Incident-Reporting.aspx in order to determine the reporting requirements. If the online flow charts are unavailable, then use Table 1, Summary of Reporting Requirements for Non-Radiological Releases To the Environment, to supplement 6.3.2.e. for reporting requirements. The Environmental Compliance Coordinator or Electric Environmental Services should be consulted when assessing oil or chemical release reportability.

e. If oil may have been released from the Station into state waters that:

Violates applicable water quality standards (i.e., any oil in water) [40 CFR 110.3]

Causes a film or sheen upon or discoloration of the surface of the water or adjoining shorelines [40 CFR 110.3]

Causes a sludge or emulsion to be deposited beneath the surface of the water or upon adjoining shorelines [40 CFR 110.3]

or If oil can reasonably be expected to enter, or there is a substantial threat that oil will enter, state waters or storm drains (Reference 3.1.8) or If more than 25 gallons1 of oil has been or can reasonably be expected to be released to soil, including a spill within containment facilities2 (Reference 3.1.8):

or If any spill from a UST or regulated AST3 reaches a solid surface, including surfaces inside secondary containment systems and inside buildings, and (1) if there is the potential for oil to reach surface water, and/or (2) if there is the potential for greater than 25 gallons of oil to reach soil

1. The individual who observes or suspects such an event or condition shall notify the Shift Manager.

2. The Shift Manager shall notify the Manager Nuclear Operations, the Environmental Compliance Coordinator, or Environmental Policy &

Compliance, as available.

1. Notice is considered to have been given to the State Water Control Board for oil releases to the ground up to 25 gallons if and only if the facility prepares and maintains a record of such oil releases for five years, and the oil is cleaned up within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of discovery.

2. Oil tank dikes and transformer vaults are typical containment facilities.

3. A regulated AST is an above ground oil tank which has a volume capacity greater than 660 gallons and is registered with the state of VA. All ASTs with a volume capacity greater than 660 gallons are required to be registered with the state.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 78 OF 238 NOTE: The National Response Center must be notified within 15 minutes of a discovery of an oil spill to state waters.

3. If the discharge is to a storm drain system4 or state waters, the Shift Manager shall notify the National Response Center, the State Department of Environmental Quality (Water) (DEQ), and (Surry) the U.S. Coast Guard. If the discharge is to land, DEQ shall be notified. Notifications shall be documented on Attachment 1, Oil or Hazardous Substance Release Report.

See 6.1.1.a. See also Steps 6.3.4.a.5., 6.21.4, and 6.28.3.n.3 NOTE: The Environmental Compliance Coordinator or Electric Environmental Services should be consulted when assessing hazardous material release reportability.

f. If a regulated, hazardous material release to the environment1 exceeds a reporting threshold as specified in Table 12:

1. The individual who becomes aware of the release or potential release shall notify the Shift Manager. See EPIP-1.01.

1. Reportable Quantity (RQ) is the amount of a regulated, hazardous material released to the environment during a 24-hour period that must be reported in accordance with federal agency requirements. RQ only applies to a release to the environment, so will not apply for every release of a regulated, hazardous material. For example, a hazardous substance spill that is contained entirely on-site, even if more than the RQ, is not reportable because it is not a release to the environment. However, if an RQ amount evaporates or is absorbed in soil, the spill has not been contained entirely on-site, and thereby becomes a reportable release to the environment.

If the VPDES or other permit authorizes discharge of a hazardous material, a discharge is not reportable as a release to the environment unless a discharge amount or concentration exceeds the permit-authorized limit or the discharge is via a pathway not specified during the permit application and approval process. Permit-authorized discharges are reportable only as required by the applicable permit (e.g., the monthly Discharge Monitoring Report, per Step 6.28.3.i., required by the VPDES permit).

If an amount or concentration does exceed a permit-authorized limit or is discharged via a pathway other than specified during the permit application and approval process, the RQ and associated reporting requirements will apply.

2. Table 1 does not mention PCBs because no PCBs are in use at the Station. The Environmental Compliance Coordinator or Electric Environmental Services should be contacted for further instructions if any question arises concerning PCBs being introduced on-site and any consequent reporting.

3. If the discharge occurs in the Main Switchyard the Dominion Energy System Operator Transmission shall be notified. If the discharge is from the transformer belonging to Rappahannock Electric Cooperative at the Dam then that company shall be notified (North Anna)

4. An oil spill to a storm drain which discharges to an oil/water separator or to the discharge canal is not considered a reportable oil spill unless the oil spill breeches the oil/water separator to state waters and/or the oil spill goes beyond the end of the discharge canal. See also Steps 4.38.1 and 4.38.2.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 79 OF 238

2. The Shift Manager shall notify the Manager Nuclear Operations, the Environmental Compliance Coordinator, or Electric Environmental Services, as available.

3. The Environmental Compliance Coordinator or Electric Environmental Services (see Step 6.3.1.a.) (North Anna) Shift Manager (Surry) shall notify the agencies listed in the Report To column of Table 1. If a reportable release involves off-site transportation (including storage incident to such transportation), the Shift Manager shall also notify the 911 operator, local and state police, and the National Response Center. Notifications shall be documented on Attachment 1, Oil or Hazardous Substance Release Report. See Step 6.1.1.a. See also Steps 6.3.2.h., 6.3.4.a.5., 6.23.3.b. and 6.28.3.n.

[CERCLA Sec. 304(b)(1); 40 CFR 302]

NOTE: Items marked with an asterisk(*) on Attachment 1 are required to be reported to the response agencies listed in block 10 of the attachment. (Reference 3.1.104)

4. Notifications shall include (to the extent known) [CERCLA Sec. 304(b)(2)]:

The chemical name or identity of the substance involved in the release

Whether the substance is on the list referred to in section 302(a) of CERCLA, 40 CFR 302.

An estimate of the quantity of substance released to the environment

The time and duration of the release

The medium or media into which the release occurred

Any known or anticipated acute or chronic health risks associated with the emergency and, where appropriate, advice regarding medical attention necessary for exposed individuals

Proper precautions to take as a result of the release, including evacuation

The name and telephone number of the Dominion Energy contact

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 80 OF 238

g. If the Station does not comply with one or more limitations, standards, monitoring, or management requirements specified in the VPDES permit (if oil is involved, go to Step 6.3.2.e.; if hazardous materials are involved, go to Step 6.3.2.f.) and such noncompliance:

May adversely affect State waters or

May endanger public health1 As soon as possible, the Environmental Compliance Coordinator or Electric Environmental Services shall notify (see Step 6.3.1.a.) the State Department of Environmental Quality (Water) by telephone with the following information

[VPDES Permit]:

A description and cause of noncompliance

The period of noncompliance, including exact dates and times or anticipated time when the noncompliance will cease

Actions taken or planned to reduce, eliminate, and prevent recurrence See also Steps 6.3.4.a.5., 6.28.2.a.1., and 6.28.3.n.

1. Applicable regulations use, but do not define, the terms adversely affect and endanger public health. These terms must be interpreted on a case-by-case basis by individuals with aquatic ecology expertise and thorough familiarity with current regulatory agency reporting and enforcement policy. Such individuals will also determine how soon a specific event must be reported to avoid enforcement (i.e., within minutes of an event, or some longer time within the not-to-exceed 24-hour limit established by the VPDES Permit).

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 81 OF 238

h. If an incident occurs during transport (including loading, unloading, and temporary storage) of:

Radioactive materials in which fire, breakage, spillage, or suspected radioactive contamination occurs (see also Step 6.29.3) [49 CFR 171.15(b)(2)]

Hazardous materials in which any of the following is a direct result of the hazardous materials: [49 CFR 171.15(b)(1)]

A person is killed A person requires hospitalization because of injuries An evacuation of the general public occurs lasting one or more hours One or more major transportation arteries or facilities are closed or shut down for one hour or more The operational flight pattern or routine of an aircraft is altered

A situation exists (e.g., a continuing danger to life exists at the scene of the incident) that, in the judgment of the carrier or Dominion Energy, should be reported even though it does not meet one of the previous criteria [49 CFR 171.15(b)(5)]

Licensing Manager (Station) shall notify (see Step 6.3.1.a.) DOT by telephone, or confirm carrier notification of DOT by telephone. See also Steps 6.3.2.f.

and 6.22.2. The notification shall include the [49 CFR 171.15(a)]:

Notifiers name

Name and address of carrier represented by the notifier

Phone number where the notifier can be contacted

Date, time, and location of incident

The extent of injuries, if any

Classification, name, and quantity of radioactive or hazardous materials involved, if available

Type of incident and nature of radioactive or hazardous material involvement and whether a continuing danger to life exists at the scene

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 82 OF 238

i. If a serious accident or a death occurs or a public safety rescue at or immediately above or below Lake Anna Dam or Dikes 3 and 61 or is alleged to be related to the existence or operation of the dam:

1. The Lake Anna Dam Operator shall notify the Shift Manager and provide information necessary to prepare Attachment 2, FERC Public Safety Database Report.

2. The Shift Manager shall initiate a Condition Report in accordance with PI-AA-200.

3. The Shift Manager should notify the FERC Regional Engineer of the condition by telephone and notify Station Licensing to follow-up with a summary e-mail to the FERC Regional Engineer. See Step 6.1.1.a. See also Steps 6.3.4.a.5., 6.3.5.c., and 6.19.2.b. (North Anna)

j. If a condition is identified that affects the safety of Lake Anna Dam or its associated works including failure of a spillway gate to operate (see Subsection 4.8), but does not require entry into the North Anna Hydroelectric Project Emergency Action Plan:

1. The Lake Anna Dam Operator shall notify the Shift Manager and provide relevant supporting information.

2. The Shift Manager shall notify, by telephone, the FERC Regional Engineer of the condition, initiate a Condition Report in accordance with PI-AA-200, and notify Station Licensing to follow-up with a summary e-mail to the FERC Regional Engineer. See Step 6.1.1.a. See also Steps 6.3.4.a.5. and 6.19.1.b.

(North Anna) [18 CFR 12.10(a)]

k. If any release occurs at the facility, at or above the RQ or notification limit, that causes radiation and/or gas, mist, solids, liquids to be released to the atmosphere, which results in an injury of an employee or member of the public that requires hospitalization, or the death of an employee or member of the public, or property damage equal to or exceeding $1.0M, then a notification to the federal Chemical Safety Board (CSB) must be made within 30 minutes of making a notification to the NaRC. See ATTACHMENT 9 for reporting details.

1. Incidents which involve other parts of the lake are excluded. [18 CFR 12.10(b)(4)]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 83 OF 238 6.3.3 One-hour Notifications NOTE: Some conditions, indicated by See EPIP-1.01, may exceed an Emergency Action Level (EAL) as specified in EPIP-1.01, Emergency Manager Controlling Procedure.

If a condition exceeds an EAL, EPIPs control State and Federal agency notifications.

If an event or condition does not exceed an EAL, it may still be reportable in accordance with this procedure.

As soon as practical, but within one hour, the Shift Manager, Station Emergency Manager, or Site Vice President shall notify the NRC Operations Center of:

a. Deviation from Technical Specifications (permitted by 10 CFR 50.54(x)) to protect the health and safety of the public, when no action consistent with license conditions and Technical Specifications can provide adequate or equivalent protection. [10 CFR 50.72(b)(1)]

NOTE: Notifications required by Steps 6.3.3.b., 6.3.3.c., and 6.3.3.d., are exempt from the requirement that Safeguards Information be transmitted only by protected telecommunications circuits approved by NRC.

NOTE: Events involving physical security are evaluated for Reportability by Protection Services in accordance with SY-AA-105.

b. An accidental criticality or loss of SNM. See EPIP-1.01.

[10 CFR 70.52 (a), 10 CFR 72.74(a), 10 CFR 74.11a]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 84 OF 238 NOTE: Step 6.3.3.c. notifications need not duplicate Step 6.3.3.d. notifications.

[10 CFR 74.11(c), 10 CFR 72.74(c)]

c. A loss of any [10 CFR 73.71(a)(1), 10 CFR 73.67(e)(3)(vii), 10 CFR 73.67(g)(3)(iii)]:

SNM shipment

Spent fuel shipment or Availability of supplemental information after initial notification. [10 CFR 73.71(a)(5)]

(See also Step 6.16.3.a.3.)

or Recovery of or accounting for such lost shipment.

See also Step 6.16.3.a.2. [10 CFR 73.71(a)(1), 10 CFR 73.67(e)(3)(vii), 10 CFR 73.67(g)(3)(iii)]

NOTE: Steps 6.3.3.d., 6.3.3.f., 6.3.3.g., 6.3.3.h.notifications need not duplicate Step 6.3.3.c.

or 10 CFR 50.72 notifications. [10 CFR 72.74(c), 10 CFR 73.71(e), 10 CFR 74.11(c)]

d. A reason to believe that a person has committed or caused, or attempted to commit or cause, or has made a credible threat to commit or cause (See also Step 6.16.3.b.2.).

[10 CFR 73.71(b)(1), 10 CFR 73 App. G.I, 10 CFR 70.52 (a), 10 CFR 72.74(a), 10 CFR 74.11(a)]:

Theft, loss, or unlawful diversion of SNM

Significant physical damage to the Station, nuclear fuel, or carrier of nuclear fuel

Interruption of normal operation through unauthorized use of or tampering with its machinery, components, or controls, including the security system

e. When informed by Security or Radiation Protection that a shipment of Category 1 quantities of radioactive material is lost or missing (refer to Appendix A to 10 CFR Part 37 - Physical Protection of Category 1 and Category 2 Radioactive Materials),

notify the NRC's Operations Center (301-816-5100) and Provide agreed upon updates to the NRC's Operations Center on the status of the investigation. [10 CFR 37.81(a)]

f. Actual entry of an unauthorized person into a protected area, material access area, controlled access area, vital area, or transport. (10 CFR 73, App. G (I)(b))

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 85 OF 238

g. Failure, degradation, or the discovered vulnerability in a safeguard system that could allow unauthorized or undetected access to a protected area, controlled access area, vital area, or transport for which compensatory measures have not been employed.

NOTE: Fitness-for-duty events are reported in accordance with 10 CFR 26 instead of 10 CFR 73.71. See Steps 6.3.6.b. and 6.8.1. [10 CFR 26.73(c)]

h. Actual or attempted introduction of contraband into a protected area, material access area, or transport (see SY-AA-105).

i. Discovery that an undeclared or misclassified event or condition met all the following criteria: [10 CFR 50.72(a)(1)(i)]

Exceeded an Emergency Action Level (EAL) as specified in EPIP-1.01, Emergency Manager Controlling Procedure

The basis for the emergency class no longer exists at the time of discovery

No other reasons exist for an emergency declaration In addition, the following shall be notified:

Department of Emergency Management (at approximately the same time)

Director Nuclear Protection Services

Louisa/Surry County Administrator

j. A cyber attack that adversely impacted safety-related or important to safety functions, security function, or emergency preparedness functions (including offsite communications); or that compromised support systems and equipment resulting in adverse impacts to safety, security, or emergency preparedness functions within the scope of 10 CFR 73.54. [10 CFR 73.77(a)(1)].

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 86 OF 238 6.3.4 Four-hour Notifications NOTE: Some conditions, indicated by See EPIP-1.01, may exceed an Emergency Action Level (EAL) as specified in EPIP-1.01, Emergency Manager Controlling Procedure.

If a condition exceeds an EAL, EPIPs control State and Federal agency notifications.

If an event or condition does not exceed an EAL, it may still be reportable in accordance with this procedure.

a. As soon as practical, but within four hours, the Shift Manager shall notify the NRC Operations Center via the ENS of:

NOTE: If a unit enters a limiting condition for operation (LCO) and a unit shutdown is started due to the LCO, the event is reportable even if shutdown is not completed. LCOs terminated by a unit shutdown for an unrelated reason are still reportable if the condition would not have been corrected within the LCO time limit for shutdown.

1. Initiation of plant shutdown (reduction of power or temperature) required by Technical Specifications. The initiation of plant shutdown does not include mode changes required by Technical Specifications if initiated after the plant is already in a shutdown condition. See EPIP-1.01. [10 CFR 50.72(b)(2)(i),

10 CFR 50.36 (c)(2)(i), NUREG 1022 Item 3.2.1]

2. Any event that results in a Technical Specifications safety limit violation and requires a reactor shut down shall be reported in accordance with 10 CFR 50.72(b)(2)(i); see also Steps 6.24.3 (North Anna) and 6.25.3 (Surry)

[10 CFR 50.36(c)(1)(i)(A)]

3. Any event that results or should have resulted in ECCS discharge into the RCS as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. [10 CFR 50.72(b)(2)(iv)(A)]

4. Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when actuation results from and is part of a pre-planned sequence during testing or reactor operation.

[10 CFR 50.72(b)(2)(iv)(B)]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 87 OF 238 NOTE: Notification to other government agencies has been or will be made is not necessarily an automatic notification to the NRC. Refer to NUREG - 1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, for discussions and examples (e.g., newsworthy events, environmental events, spurious, emergency siren actuations) or contact Station Licensing if clarification is needed. [NUREG-1022, Section 3.2.12]

5. Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned, or notification to other government agencies has been or will be made.

Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials. [Commitment 3.2.12] [10 CFR 50.72(b)(2)(xi)]

6. ISFSI Non-emergency Four-Hour Notifications shall include, if available at time of notification: [10 CFR 72.75(e)(3)]

The callers name and call back telephone number

A description of the event, including time and date

The exact location of the event

The quantities, and chemical and physical forms of the spent fuel, HLW or reactor related Greater than Class C (GTCC) waste involved

Any personnel radiation exposure data

7. An action taken in an emergency that departs from a license condition, technical specification, or certificate of compliance when the action is immediately needed to protect the public health and safety and no licensed action that provides adequate or equivalent protection is immediately apparentsee Step 6.15.7.f. [10 CFR 72.75(b)(1)]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 88 OF 238

8. Groundwater Protection Voluntary Communication Notifications to other government agencies may be reportable under 10 CFR 50.72(b)(2)(xi) requirement for a 4-hour notification to the NRC operations center based upon the following guidance:

If a licensee is notifying a local, state, or other federal agency in accordance with an existing law, regulation, or ordinance, then the licensee should make its notification to the NRC under the 50.72 notification requirement.

If a licensee is informally communicating with a local, state, or other federal agency (i.e., not under a specific law, regulation or ordinance), then the licensee has discretion as to whether to informally communicate with NRC (e.g., through the site resident inspector and/or regional NRC office) or formally through the 50.72 notification process. If due to the site-specific circumstances or heightened sensitivity to the issue at that site, the issue is likely to produce strong media interest, then the licensee should consider notifying NRC under the 50.72 requirement because this is actually the underlying intent of the regulation.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 89 OF 238 NOTE: During equipment start up or when adding an electrical load, diesels and boilers are NOT considered steady operating conditions.

b. Any person at the Station who observes unusual exhaust being released from stationary diesel equipment or the Auxiliary Boiler when at steady operating conditions shall notify the Shift Manager as soon as possible.

1. The Shift Manager shall notify the Environmental Compliance Coordinator (ECC), if available, of the unusual exhaust. IF the ECC is NOT available, THEN the Shift Manager may initiate action to determine the opacity, cause, and duration of the exhaust OR initiate shutdown of the diesel or boiler, if possible, before 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months duration of the unusual exhaust release is reached.

2. IF the exhaust opacity is 20 percent or more for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , THEN within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after discovery the Shift Manager or the ECC shall NOTIFY the Dominion Energy Environmental Services (DEES) to determine the potential reportability.

3. IF the unusual exhaust is reportable, THEN DEES will NOTIFY the appropriate DEQ regional office as soon as practical, but no later than 4 daytime business hours of the occurrence, with all of the pertinent information. (Refer to SPS and NAPS Air Permits.)

4. DEES will prepare and submit any written reports to the DEQ regional office regarding the release of unusual exhaust into the outdoor atmosphere within 14 days of the discovery. (Refer to SPS and NAPS Air Permits.)

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 90 OF 238

c. When informed by Security or Radiation Protection of events related to the shipment or onsite storage of Category 1 or Category 2 radioactive material (refer to Appendix A to 10 CFR Part 37 - Physical Protection of Category 1 and Category 2 Radioactive Materials), notify the NRC's Operations Center (301-816-5100) upon:

1. Determination that a shipment of Category 2 quantities of radioactive material is lost or missing [10 CFR 37.81(b)]

and If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months since the determination that the shipment was lost or missing and the radioactive material has not been located and secured, immediately the NRC's Operations Center

2. Discovery of any unauthorized entry that resulted in the actual or attempted threat, sabotage, or diversion of Category 1 or Category 2 quantity of radioactive material [10 CFR 37.57(a)]

3. Discovery of any suspicious activity related to the possible theft, sabotage, or diversion of Category 1 or Category 2 quantities of radioactive material

[10 CFR 37.57(b)]

d. When informed by Security of cyber attack that:

1. Could have caused an adverse impact to safety-related or important to safety functions, security function, or emergency preparedness functions (including offsite communications); or that could have compromised support systems and equipment, which if compromised, could have adversely impacted safety, security, or emergency preparedness functions within the scope of 10 CFR 73.54.

2. After discovery of a suspected or actual attack initiated by personnel with physical or electronic access to digital computer and communication system and networks within the scope of 10 CFR 73.54.

3. After notification of a local, State, or other Federal agency of an event related to implementation of the licensees cyber security program for digital computer and communication system and networks within the scope of 10 CFR 73.54.

[10 CFR 73.77(a)(2)]

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 91 OF 238 6.3.5 Eight-hour Notifications NOTE: Any event or condition that occurred within three years of the date of discovery.

Applicable to 6.3.5.a.1., 6.3.5.a.2., 6.3.5.a.5. and 6.3.5.a.7.

a. As soon as practical, but within eight hours, the Shift Manager shall notify the NRC Operations Center via the ENS of:

1. Any condition that results in the condition of the Station, including its principal safety barriers, being seriously degraded. [10 CFR 50.72(b)(3)(ii)(A)]

2. Any event or condition that results in the Station being in an unanalyzed condition that significantly degrades plant safety. [10 CFR 50.72(b)(3)(ii)(B)]

3. Any event that results in the Limiting Safety System settings for automatic protective devices to not function as required. [10 CFR 50.36(c)(1)(ii)(A)]

4. Any event or condition that results in valid actuation of any of the following systems, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation: [10 CFR 50.72(b)(3)(iv)(A)]

Reactor Protection System (RPS) - (RPS actuation with the reactor critical may be reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months under 10 CFR 50.72(b)(2)(iv)(B), see Step 6.3.4.a.4.)

General containment isolation signals affecting containment isolation valves in more than one system or multiple Main Steam Isolation Valves (MSIVs)

Emergency Core Cooling Systems (ECCS) including HHSI and LHSI (Actual discharges are reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months under 10 CFR 50.72(b)(2)(iv)(A), see Step 6.3.4.a.3.)

Auxiliary Feedwater System

Containment heat removal and depressurization systems including Containment spray and fan cooler systems

Emergency Diesel Generators (EDGs)

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 92 OF 238

5. Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:

Shut down the reactor and maintain it in a safe shutdown condition

Remove residual heat

Control the release of radioactive material; or

Mitigate the consequences of an accident. See EPIP-1.01. [10 CFR 50.72(b)(3)(v)]

6. Any event requiring the transport of a radioactively contaminated person to an off-site medical facility for treatment. See also Step 6.28.2.

[10 CFR 50.72 (b)(3)(xii) and 10 CFR 72.75 (c)(3)]

7. Any event that results in a major loss of emergency assessment capability, off-site response capability, or off-site communications capability, (e.g.,

significant portion of control room indication, Emergency Notification System, or offsite notification system). Equipment important to emergency response and emergency response facilities are listed in the attachments of EP-AA-303, Equipment Important to Emergency Response. See Attachment 3, Emergency Response Unavailability Reportable Actions Levels, for reportable action level criteria.

Emergency Assessment Capability

Offsite Response Capability

Offsite Communications Capability See EPIP-1.01. [10 CFR 50.72(b)(3)(xiii)]

8. Any instance of:

A defect in any spent fuel storage cask structure, system, or component that is important to safety [10 CFR 72.75(c)1]

or

A significant reduction in the effectiveness of any spent fuel storage cask confinement system during use of the storage cask [10 CFR 72.75(c)2 See EPIP-1.01.

DOMINION ENERGY VPAP-2802 REVISION 51 PAGE 93 OF 238

9. After receipt or collection of information regarding observed behavior, activities, or statements that may indicate intelligence gathering or pre-operational planning related to a cyber attack against digital computer and communication system and networks within the scope 10 CFR 73.54.

[10 CFR 73.77(a)(3)].

b. If an Alert, Site Area Emergency, or General Emergency is declared:

1. The Station Coordinator Emergency Preparedness shall prepare a Summary Report from information in completed Emergency Plan Implementing Procedures, Control Room logs, and interviews with persons involved with the declaration and response, as appropriate. See Attachment 6, Example DEM Summary Report.

2. The Site Vice President, Director Nuclear Station Safety and Licensing, or Plant Manager (Nuclear) shall approve the report.

3. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after termination of the event, Nuclear Emergency Preparedness shall ensure the report is delivered to the State Coordinator of the Virginia Department of Emergency Management. [NAEP 4.4; SEP 4.4]

c. If, on Dominion Energy property or at Lake Anna Dam, there is a Dominion Energy employee or contractor fatality including cardiac arrest or an event in which three or more Dominion Energy employees or contractors are hospitalized:

1. The Shift Manager shall notify Supervisor Nuclear Site Safety (Station) with the following information:

Number of fatalities

The employer of those killed

The circumstances of the event

The extent of injuries

2. Nuclear Site Safety (Station) shall notify OSHA as specified in Step 6.3.5.c.3.

ML22214A027

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