ML21137A299

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Proposed Written Exam
ML21137A299
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 02/18/2021
From: Gregory Roach
NRC/RGN-III/DRS/OLB
To:
FirstEnergy Nuclear Operating Co
Shared Package
ML20136A254 List:
References
Download: ML21137A299 (63)
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Davis Besse 1LOT21 NRC Written Exam Rev. 1

1. Initial Plant Conditions:
  • DB-OP-02000, Section 4.0 Supplemental Actions are in progress Current Plant Conditions:
  • Security Reports steam issuing from a MSSV located on the West side of the Aux Building roof
  • NNI X AC Power Available Light is OFF
  • Control Rod 3-1 has indications of being stuck out at 100%

Based on Current Plant Conditions, which of the following actions will be completed NEXT?

A. Emergency Borate the Reactor Coolant System refer to DB-OP-02516 CRD Malfunctions B. Place #1 Atmospheric Vent Valve in HAND and lower Steam Generator 1 Pressure refer to DB-OP-02525 Steam Leaks C. Initiate AND Isolate Steam Feed Rupture Control System (SFRCS) refer to DB-OP-02532 Loss of NNI/ICS Power D. Start the STBY Makeup Pump refer to DB-OP-02512 Makeup and Purification System Malfunctions

Davis Besse 1LOT21 NRC Written Exam Rev. 1

2. Initial Conditions:
  • 100% power
  • Makeup Pump 1 is in service
  • Component Cooling Water Pump 1 is in service A plant transient occurs which causes the following Annunciators:
  • 6-5-C SEAL INJ LO
  • 6-6-C SEAL INJ TOTAL FLOW
  • 1-3-D BUS C1 L/O
  • 4-2-E PZR LVL LO All systems responded as expected and all Immediate Actions have been completed.

Which of the following actions is required NEXT IAW DB-OP-02512 Makeup and Purification System Malfunctions to mitigate this event?

A. Monitor RCP Seal performance. REFER TO DB-OP-02515 Reactor Coolant Pump and Motor Abnormal Operation B. Verify CCW is being supplied to the RCPs. REFER TO Attachment 1 Verification Of CCW Flow To Reactor Coolant Pumps C. Trip the Reactor, GO TO DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture D. Trip the Reactor, Stop ALL RCPs, GO TO DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture

Davis Besse 1LOT21 NRC Written Exam Rev. 1

3. Following an ATWS event the Reactor was shutdown by momentarily Deenergizing 480-volt Unit Substations E2 AND F2.

Based on the above event, complete the following statements:

This event will be classified as an ____(1)_____.

After the rods drop into the core, Absolute Position Indication for Groups 1 through 7 will ____(2)____ when power is restored to E2 and F2.

REFERENCE PROVIDED A. (1) Unusual Event (2) indicate 0%

B. (1) Unusual Event (2) NOT be available C. (1) Alert (2) indicate 0%

D. (1) Alert (2) NOT be available

Davis Besse 1LOT21 NRC Written Exam Rev. 1

4. The plant is operating at 100% power.

The breaker for MS107 Steam Generator 2 to Auxiliary Feed Pump Turbine 2 trips open and cannot be reset.

Which of the following describes the required action?

The __(1)__ Limiting Condition for Operation must be restored within __(2)__.

A. (1) Steam and Feed Rupture Control System Actuation Logic (2) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. (1) Emergency Feedwater System (2) 7 days C. (1) Steam and Feed Rupture Control System Actuation Logic (2) 7 days D. (1) Emergency Feedwater System (2) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

Davis Besse 1LOT21 NRC Written Exam Rev. 1

5. Initial conditions
  • Plant shutdown in progress
  • Reactor Power is 12%
  • Loss of Offsite Power
  • AFW Pump 1 cannot be restored and is still unavailable
  • Steam Generator 1 Tube to Shell differential temperatures are within limits
  • Station Blackout Diesel Generator has been started

(1) Which Section of Attachment 5 will the Command SRO direct the Reactor Operator to perform?

AND (2) Which of the following actions will be required to restore flow to Steam Generator 1?

A. (1) Section A: Motor Driven Feedwater Pump in the Main Feedwater Mode (2) Block AND Reset SP7B using both HIS SP7AB and HIS SP7CB on panel C5792N, MSIV/MFW Control Valve Reset Switch Panel B. (1) Section A: Motor Driven Feedwater Pump in the Main Feedwater Mode (2) Block AND Reset SP7B using both HIS SP7AB and HIS SP7CB on panel C5712, CTRM Right Console (MFW Control) Panel C. (1) Section B: Emergency Feedwater Pump via the Auxiliary Feedwater header (2) Throttle EFW Flow to Steam Generator 1 using HCEF8-2, EFWP DISCHARGE FLOW VALVE on panel C5732, EFW Control Panel D. (1) Section B: Emergency Feedwater Pump via the Auxiliary Feedwater header (2) Throttle EFW Flow to Steam Generator 1 using HCEF8-2, EFWP DISCHARGE FLOW VALVE on panel C5706, CTRM Center Console (AFW Control) Panel

Davis Besse 1LOT21 NRC Written Exam Rev. 1

6. Plant Conditions:
  • The Reactor tripped due to a loss of Off-site power
  • Both Essential Busses C1 AND D1 remain de-energized AND attempts to start the SBODG and Both EDGs IAW Specific Rule 6 actions have failed Based on the Plant Conditions, what procedure will be utilized and how will Steam Generator 2 level be initially controlled?

A. DB-OP-02521 Loss of AC Bus Power Sources will direct Steam Generator 2 to be controlled at the required level by adjusting HIS521A AFPT 2 GOVERNOR.

B. DB-OP-02521 Loss of AC Bus Power Sources will direct closing AF599 AFW TO SG 2 LINE STOP to terminate an AFW Overcooling event.

C. DB-OP-02704 Extended Loss of AC Power DC Load Management will direct Steam Generator 2 to be controlled at the required level by adjusting HIS521A AFPT 2 GOVERNOR.

D. DB-OP-02704 Extended Loss of AC Power DC Load Management will direct closing AF599 AFW TO SG 2 LINE STOP to terminate an AFW Overcooling event.

Davis Besse 1LOT21 NRC Written Exam Rev. 1

7. A planned power reduction from 100% power, with ICS in full Automatic was in progress to remove #1 Condensate Pump from service for maintenance. When at 95%

power, the following indications were received:

  • (5-1-E) CRD SYSTEM FAULT
  • (5-2-E) CRD ASYMMETRIC ROD
  • Tave lowered to approximately 579°F
  • NI 5 indicates 89%
  • NI 6 indicates 96%
  • NI 7 indicates 88%
  • NI 8 indicates 95%

Based on the indications above, complete the following statements.

DB-OP-02516, CRD MALFUNCTIONS will be entered for indications of a (1)___.

IAW DB-OP-002516, CRD MALFUNCTIONS, reactor power will be lowered to ___(2)___ RTP.

A. (1) Stuck Rod (2) 60%

B. (1) Stuck Rod (2) 50%

C. (1) Dropped Rod (2) 60%

D. (1) Dropped Rod (2) 50%

Davis Besse 1LOT21 NRC Written Exam Rev. 1

8. A plant startup-up is in progress with the following conditions:
  • Reactor power, as indicated on NI 3 and NI 4 (intermediate range detectors),

is 1 X 10-8 amps

  • All systems are in normal alignment for this condition A fuse in the power supply to the NI 3 detector blows (detector supply voltage is zero).

(1) How will the NI 1 and NI 2 (source range detectors) respond to this blown fuse?

(2) IAW Technical Specifications, what actions are required?

REFERENCE PROVIDED A. (1) re-energize (2) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, Reduce neutron flux to 1E-10 amp B. (1) re-energize (2) once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, Verify SDM is within the limits specified in the COLR C. (1) remain de-energized (2) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, Reduce neutron flux to 1E-10 amp D. (1) remain de-energized (2) once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, Verify SDM is within the limits specified in the COLR

Davis Besse 1LOT21 NRC Written Exam Rev. 1

9. Given the following conditions:
  • The crew has entered DB-OP-02518 High Condenser Pressure
  • Reactor power is at 34%
  • Condenser Pressure is at 5.6 inches HgA and rising slowly
  • Turbine Load is at 235 MWe IAW DB-OP-02518 High Condenser Pressure which ONE of the following actions is required next?

A. Reduce Reactor power to maintain Condenser pressure less than or equal to 5.0 inches HgA. REFER TO DB-OP-02504 Rapid Shutdown.

B. Trip the Turbine. REFER TO DB-OP-02500 Turbine Trip.

C. Trip the Reactor AND GO TO DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture.

D. Trip the Reactor AND Initiate AND Isolate SFRCS THEN GO TO DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture.

Davis Besse 1LOT21 NRC Written Exam Rev. 1

10. The plant is operating at 100% power in a normal system alignment.

The following plant conditions are noted:

  • The plant remains steady at 100% power.
  • It is determined that Ch 2 SCM Meter will NOT respond as required.

Which of the following is required due to a failed RC System Subcooling Margin Monitor?

REFERENCE PROVIDED A. Only TNC 8.3.7, Post Accident Monitoring (PAM) Instrumentation is NOT met, comply with Nonconformance A.

B. Only LCO 3.3.17, Post Accident Monitoring (PAM) Instrumentation is NOT met, comply with Condition A.

C. Both TNC 8.3.7 and LCO 3.3.17 are NOT met, comply with TNC 8.3.7 Nonconformance A and LCO 3.3.17 Condition A.

D. Both TNC 8.3.7, Post Accident Monitoring (PAM) Instrumentation and LCO 3.3.17, Post Accident Monitoring (PAM) Instrumentation are met, no action is required.

Davis Besse 1LOT21 NRC Written Exam Rev. 1

11. The plant is operating at 100% power with all systems in a normal alignment.

At 0800, a reactor trip occurs. SFAS Actuates on Low RCS Pressure, Low-Low RCS Pressure and High Containment Pressure.

At 0815, The Zone 3 Equipment Operator reports HPI Pump 1 discharge pressure is approximately 150 psig and has an abnormal running noise.

At 0830, BWST level is 39 feet and lowering and level will reach 9 feet at 1630.

At 0845, LPI Train 1 AND 2 indicate 0 gallons per minute.

At 0900, Incore temperatures have stabilized at approximately 480 ºF with RCS pressure at 500 psig.

Which ONE (1) of the following DB-OP-02000 Attachments provides the required actions that mitigate these plant events?

A. Attachment 11, HPI Flow Balancing.

B. Attachment 12, Establishing Long Term Boron Dilution.

C. Attachment 14, Establishing HPI Alternate Minimum Recirc Flowpath.

D. Attachment 22, Cross Connect LPI Pump Discharge.

Davis Besse 1LOT21 NRC Written Exam Rev. 1

12. Which of the following transients or failures is the Reactor Protection System credited for to ensure the Reactor Coolant System Pressure Safety Limit is not exceeded, per Technical Specification Bases?

A. Make-up flow transient B. Pressurizer Spray Valve failure C. Pressurizer Pilot Operated Relief Valve (PORV) failure D. Rod withdrawal transient from subcritical condition

Davis Besse 1LOT21 NRC Written Exam Rev. 1

13. Reactor Coolant Pump 1-1 was stopped at 70% power IAW Attachment 1 of DB-OP-02515 Reactor Coolant Pump and Motor Abnormal Operation due to high vibration.

The BOP Reactor Operator checks for proper Feedwater (FW) flow ratios and reports SP6A SG 2 MFW Control Valve is not responding to FW Loop Demand.

Which of the following statements describes the impact of the failed MFW Control Valve and the appropriate direction from the CSRO?

A. SG2 will experience an overfeed condition, the CSRO will direct the BOP Reactor Operator to place Both FW Loop Demands, SP6A and SP7A in Manual to control FW Flow B. SG2 will experience an overfeed condition, the CSRO will direct the ATC Reactor Operator to Trip the Reactor and Initiate AND Isolate SFRCS C. SG2 will experience an underfeed condition, the CSRO will direct the BOP Reactor Operator to place Both FW Loop Demands, SP6A and SP7A in Manual to control FW Flow D. SG2 will experience an underfeed condition, the CSRO will direct the ATC Reactor Operator to Trip the Reactor and Initiate AND Isolate SFRCS

Davis Besse 1LOT21 NRC Written Exam Rev. 1

14. The following plant conditions exist:
  • The plant is DEFUELED
  • C1 electrical bus is tagged out for maintenance
  • The SBODG is functional Which one of the following conditions would require a declaration of an ALERT if the event was expected to last for more than 15 minutes?

REFERENCE PROVIDED A. A D1 electrical bus lockout B. A loss of the SBODG and EDG 2 C. A loss of offsite power concurrent with a loss of EDG 2 D. A loss of offsite power concurrent with a loss of the SBODG

Davis Besse 1LOT21 NRC Written Exam Rev. 1

15. Initial conditions:
  • Plant is at 100% Power
  • Loss of Offsite Power (LOOP)
  • EDG 2 trips Which of the following correctly states the response of EDG 1 to the LOOP and which procedure which will be entered NEXT?

A. EDG 1 output breaker, AC 101, will open immediately due to 13.8 Kv Bus Undervoltage Relay GO TO DB-OP-02521, Loss of AC Bus Power Sources B. EDG 1 will remain loaded until an undervoltage condition occurs and isolates the essential bus GO TO DB-OP-02521, Loss of AC Bus Power Sources C. EDG 1 output breaker AC 101, will open immediately due to 13.8 Kv Bus Undervoltage Relay GO TO DB-OP-02000, RPS, SFAS, SFRCS Trip OR SG Tube Rupture D. EDG 1 will remain loaded until an undervoltage condition occurs and isolates the essential bus GO TO DB-OP-02000, RPS, SFAS, SFRCS Trip OR SG Tube Rupture

Davis Besse 1LOT21 NRC Written Exam Rev. 1

16. The plant is operating at 100% power with all systems in normal alignment EXCEPT:

While maintenance is being performed in RPS CH 1, RPS CH 4 spuriously trips on High RCS Pressure and the reactor does NOT trip.

What is the Required Action for Limiting Conditions for Operations (LCO) 3.3.1 Reactor Protection System (RPS) Instrumentation?

A. Enter Condition A for RPS CH 4, a Separate Condition entry is allowed for each Function B. Enter Condition B for RPS CH1 and CH4, place CH 4 in Bypass within one hour C. Enter Condition B, for RPS CH1 and CH4, verify CH 4 is tripped within one hour D. Enter Condition A for RPS CH4 and Condition B for RPS CH 1 and CH 4, be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

Davis Besse 1LOT21 NRC Written Exam Rev. 1

17. Plant conditions:
  • A large break LOCA has occurred
  • DB-OP-02000 RPS, SFAS, SFRCS TRIP, or SG Tube Rupture Section 5.0 Lack of Adequate Subcooling Margin is in progress
  • PAM Channel 2 is NOT available Which of the following indications provide confirmation that routing to Section 9 Inadequate Core Cooling is required?

A. P/T plot is in Region 2 RCS Superheated and trending parallel to the saturation curve B. Two working incore detectors are displaying a NEG MARGIN C. P/T plot is in Region 2 RCS Superheated and trending toward Region 3 D. RCS Pressure is less than 150 psig and Channel 1 Tsat meter is blank

Davis Besse 1LOT21 NRC Written Exam Rev. 1

18. DB-SS-04150, MAIN TURBINE STOP VALVE TEST and DB-SS-04159, ONLINE ELECTRICAL TRIP DEVICE TEST are BOTH scheduled to be performed during the shift.

Complete the following statements.

During the performance of ___(1)___ the ARTS TURBINE-GEN Bypass Switch is required to be placed in the BYPASS position.

When the ARTS TURBINE-GEN Bypass Switch is placed in the BYPASS position, LCO 3.3.16 Anticipatory Reactor Trip System (ARTS) Instrumentation Condition A

___(2)___ met.

A. (1) DB-SS-04159, ONLINE ELECTRICAL TRIP DEVICE TEST (2) is B. (1) DB-SS-04159, ONLINE ELECTRICAL TRIP DEVICE TEST (2) is not C. (1) DB-SS-04150, MAIN TURBINE STOP VALVE TEST (2) is D. (1) DB-SS-04150, MAIN TURBINE STOP VALVE TEST (2) is not

Davis Besse 1LOT21 NRC Written Exam Rev. 1

19. The plant is at 100% power at minimum staffing levels.

During turnover the oncoming At The Controls (ATC) Reactor Operator notifies the Unit Supervisor that he has started taking a new blood pressure medication.

Upon further investigation it has been determined that the Medical Review Officer has not reviewed the change in medication.

As the Unit Supervisor, what action is required to ensure minimum staffing levels are met?

A. Allow the oncoming ATC RO to take the watch with additional oversight and notify the MRO of the change of medication.

B. Hold over the current ATC RO within requirements of fatigue rule and immediately call out for a replacement.

C. Allow the current ATC RO to leave when a relief has been contacted since the relief will arrive within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. The Medical Review Officer has 30 days to complete a review of the change in medical status, no further action is required.

Davis Besse 1LOT21 NRC Written Exam Rev. 1

20. The crew has entered DB-OP-02531 Steam Generator Tube Leak for a Steam Generator Tube Leak on #1 Steam Generator. During the plant shutdown, the ATC Crew Updates Steam Generator Tube leak rate is now greater than 50 gpm.

With these current conditions per NOP-OP-1002, Conduct of Operations, complete the following statements.

1. The ____________ shall clearly announce to the crew when transitioning to DB-OP-02000 SFAS, RPS, SFRCS Trip or SG Tube Rupture.
2. A _____________ should be used for this announcement.

A. 1. Shift Manager

2. Crew Update B. 1. Shift Manager
2. Transient Crew Brief C. 1. Command SRO
2. Crew Update D. 1. Command SRO
2. Transient Crew Brief

Davis Besse 1LOT21 NRC Written Exam Rev. 1

21. The Plant is operating at 100%
  • Surveillance DB-PF-03811, MISCELLANEOUS VALVES TEST, is performed to meet this requirement While performing DB-PF-03811, MISCELLANEOUS VALVES TEST
  • RC11, PRESSURIZER RELIEF POWER ISO VALVE CLOSES but WILL NOT OPEN.
  • Maintenance finds a bad power supply breaker to the MOV and replaces the entire breaker assembly at the MCC.
  • ALL of their required work package instructions have been completed.
  • The tagout has been lifted, RC11 is ENERGIZED and CLOSED.
  • RC11 is ready for operations post-maintenance testing.

For these conditions:

What MINIMUM post-maintenance testing will be REQUIRED to verify compliance with Technical Specification LCO 3.4.11?

(For each of the below actions, assume all valve stroke times and indications are within acceptable limits)

A. Open RC11, PRESSURIZER RELIEF POWER ISO VALVE, no other actions required.

B. Open RC11, PRESSURIZER RELIEF POWER ISO VALVE; then Close; then re-open.

C. Cycle RC2A, PRESSURIZER POWER OPERATED RELIEF VALVE through one complete cycle, then open RC11.

D. Cycle RC2A, PRESSURIZER POWER OPERATED RELIEF VALVE through one complete cycle, then open RC11; then Close; then re-open.

Davis Besse 1LOT21 NRC Written Exam Rev. 1

22. An overhead Annunciator Alarm in the Control Room is not operating properly.

To avoid nuisance alarms, the Operations Manager has determined that the Annunciator will be disabled by removing the affected Annunciator Point Card.

Which of the following documents must be completed to remove this point card to disable the affected annunciator alarm?

1. Annunciator System Operating Procedure
2. Work Order for point card removal
3. 10CFR50.59 RAD, Screen, and/or Evaluation
4. Engineering Change Package
5. Maintenance Information Tag
6. Clearance and Red Danger Tags A. 1 and 3 B. 2, 4, and 5 C. 1 and 6 D. 2, 3 and 6

Davis Besse 1LOT21 NRC Written Exam Rev. 1

23. The Miscellaneous Waste Monitor Tank (MWMT) has been prepared for batch discharge.

The following radiation monitors and flow elements are out of service and Non-Functional

  • Miscellaneous RE 1878A
  • Miscellaneous RE 1878B
  • Clean RE 1770B
  • FE 4687 Storm Sewer Flow All other instrumentation is OPERABLE.

Based on these conditions, what Offsite Dose Calculation Manual (ODCM) actions will be required in order to discharge this tank?

REFERENCE PROVIDED A. The system/process flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the actual release.

B. At least two independent samples of the tank's content are analyzed and at least two independent verifications of the release rate calculations and discharge valve lineups are performed AND the system/process flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the actual release.

C. Grab samples are collected, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and analyzed, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for gross radioactivity (beta or gamma) at a lower limit of detection no greater than 1.0-07 µCi/ml or a gamma isotopic analysis meeting the LLD Requirement of Table 2-3.

D. At least two independent samples of the tank's content are analyzed and at least two independent verifications of the release rate calculations and discharge valve lineups are performed.

Davis Besse 1LOT21 NRC Written Exam Rev. 1

24. Plant Conditions:
  • The Plant is in a refueling outage.
  • Refueling Canal water level is lowering unexpectedly.

Sequence of Events:

  • Time= 1110: An unplanned rise in area radiation levels is indicated by RE8426 SFP Area.
  • Time= 1115: The Shift manager declares the event as an UNUSUAL EVENT.
  • Time= 1119: Prior to state and local notifications being made for the UNUSUAL EVENT, conditions change such that the Shift Manager upgrades the classification to an ALERT.
  • The Shift Manager informs the Communicator to make notifications of the ALERT, ONLY.

Given the above information, which one of the following identifies:

(1) The latest time that the State and County notifications must be made, and (2) What action is required to stop the clock?

A. (1) 1130 (2) State and County notifications initiated B. (1) 1130 (2) FENOC Nuclear Power Plant Initial Notification Form (DBEP-010) telefaxed C. (1) 1134 (2) State and County notifications initiated D. (1) 1134 (2) FENOC Nuclear Power Plant Initial Notification Form (DBEP-010) telefaxed

Davis Besse 1LOT21 NRC Written Exam Rev. 1

25. A Reactor Trip due to a Large Break LOCA occurred coincident with some fuel damage.
  • The Shift Manager/Emergency Director declared a General Emergency on FG1 10 minutes after the Reactor Trip
  • The initial notification has NOT been made at this time
  • Containment Radiation monitors RE4596A and B indicate 1.20E+4 R/hr
  • An unisolable gaseous release is in progress, from a failed containment penetration
  • The expected duration of the leakage is ~ 40 minutes
  • Wind direction is from 18°
  • Dose projections at 5 miles are 0.5 Rem TEDE and 1.5 Rem Child Thyroid CDE
  • There is no Hostile Action or Known Impediment to Evacuation Based on these conditions, what Protective Action Recommendation (PAR) is required?

REFERENCE PROVIDED A. Shelter is N/A Evacuate 2 mile radius & 5 mile downwind subarea 1, 2, 6, 12 B. Shelter is N/A Evacuate 2 mile radius & 5 mile downwind subareas 1, 2, 12 C. Shelter is N/A Evacuate 2 mile radius & 10 mile downwind subareas 1, 2, 4, 5, &12 D. Shelter 2 mile radius & 5 mile downwind subareas 1, 2 Evacuate 2 mile radius & 5 mile downwind subarea 12

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 lof 3 Davis-Besse EAL Wallboards APPROVED BY:

Emergency Response Manager Date Effective date: 6/29/17

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 2 of 3 The purpose of this reference material is to provide the Davis-Besse Emergency Action Level (EAL) wallboards. EAL wallboards are a display of the emergency action levels as described in RA-EP-01500, Emergency Classification procedure.

The intent of the wallboards is to provide an aid to be used by emergency response personnel who are evaluating plant conditions to determine if the event is classifiable.

1. The wallboards are comprised of two separate boards:
a. Hot Mode EALs - which provides as a single display those EALs that are applicable in operation Modes 1, 2, 3, and 4
b. Cold Mode EALs - which provides as a single display those EALs that are applicable in Modes 5 and 6
2. EALs that are associated with Hazards and Other Conditions Affecting Plant Safety, Dry Fuel Storage Facility (DFSF), and Abnormal Rad Levels / Rad Effluent are common to both the Hot Mode EALs and the Cold Mode EALs wallboards.
3. EAL wallboards are primarily found in the Control Room, Control Room Simulator, Technical Support Center and the Emergency Operations Facility.

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 3 of 3 TO OPEN HOT EAL OR COLD EAL PRESS CTRL AND CLICK ON DESIRED EAL.

(HOT) (COLD)

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 lof 3 Davis-Besse EAL Wallboards APPROVED BY:

Emergency Response Manager Date Effective date: 6/29/17

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 2 of 3 The purpose of this reference material is to provide the Davis-Besse Emergency Action Level (EAL) wallboards. EAL wallboards are a display of the emergency action levels as described in RA-EP-01500, Emergency Classification procedure.

The intent of the wallboards is to provide an aid to be used by emergency response personnel who are evaluating plant conditions to determine if the event is classifiable.

1. The wallboards are comprised of two separate boards:
a. Hot Mode EALs - which provides as a single display those EALs that are applicable in operation Modes 1, 2, 3, and 4
b. Cold Mode EALs - which provides as a single display those EALs that are applicable in Modes 5 and 6
2. EALs that are associated with Hazards and Other Conditions Affecting Plant Safety, Dry Fuel Storage Facility (DFSF), and Abnormal Rad Levels / Rad Effluent are common to both the Hot Mode EALs and the Cold Mode EALs wallboards.
3. EAL wallboards are primarily found in the Control Room, Control Room Simulator, Technical Support Center and the Emergency Operations Facility.

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 3 of 3 TO OPEN HOT EAL OR COLD EAL PRESS CTRL AND CLICK ON DESIRED EAL.

(HOT) (COLD)

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 lof 3 Davis-Besse EAL Wallboards APPROVED BY:

Emergency Response Manager Date Effective date: 6/29/17

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 2 of 3 The purpose of this reference material is to provide the Davis-Besse Emergency Action Level (EAL) wallboards. EAL wallboards are a display of the emergency action levels as described in RA-EP-01500, Emergency Classification procedure.

The intent of the wallboards is to provide an aid to be used by emergency response personnel who are evaluating plant conditions to determine if the event is classifiable.

1. The wallboards are comprised of two separate boards:
a. Hot Mode EALs - which provides as a single display those EALs that are applicable in operation Modes 1, 2, 3, and 4
b. Cold Mode EALs - which provides as a single display those EALs that are applicable in Modes 5 and 6
2. EALs that are associated with Hazards and Other Conditions Affecting Plant Safety, Dry Fuel Storage Facility (DFSF), and Abnormal Rad Levels / Rad Effluent are common to both the Hot Mode EALs and the Cold Mode EALs wallboards.
3. EAL wallboards are primarily found in the Control Room, Control Room Simulator, Technical Support Center and the Emergency Operations Facility.

DAVIS-BESSE REFERENCE MATERIAL Number:

DBRM-EMER-1500B

Title:

Revision: Page Davis-Besse EAL Wallboards 02 3 of 3 TO OPEN HOT EAL OR COLD EAL PRESS CTRL AND CLICK ON DESIRED EAL.

(HOT) (COLD)

Page 1 of 14 Davis-Besse Nuclear Power Station EMERGENCY PLAN IMPLEMENTING PROCEDURE RA-EP-02245 Protective Action Guidelines REVISION 09 Prepared by: G. S. Van Wey Procedure Owner: Emergency Response Manager Effective Date: 11/14/19 LEVEL OF USE:

IN-FIELD REFERENCE

2 RA-EP-02245 Revision 09 TABLE OF CONTENTS Page 1.0 PURPOSE ................................................................................................................................ 3

2.0 REFERENCES

......................................................................................................................... 3 3.0 DEFINITIONS ......................................................................................................................... 4 4.0 RESPONSIBILITIES ............................................................................................................... 6 5.0 INITIATING CONDITIONS .................................................................................................. 6 6.0 PROCEDURE ......................................................................................................................... 7 6.1 Onsite Protective Actions ............................................................................................... 7 6.2 Offsite Protective Actions .............................................................................................. 8 7.0 FINAL CONDITIONS............................................................................................................. 9 8.0 RECORDS ............................................................................................................................... 9 ATTACHMENT 1 Flowchart for Offsite Protective Action Recommendation Determination ................................................................................................ 11 ATTACHMENT 2 Comparison of Offsite Sectors and Subareas ................................................ 13 COMMITMENTS .......................................................................................................................... 14

3 RA-EP-02245 Revision 09 1.0 PURPOSE 1.1 This procedure defines specific guidelines for determining protective action recommendations for emergencies involving abnormal releases of radioactivity at the Davis-Besse Nuclear Power Station (DBNPS).

2.0 REFERENCES

2.1 Developmental 2.1.1 Davis-Besse Nuclear Power Station Emergency Plan.

2.1.2 NUREG-0654 FEMA REP-1, Rev. 1, Supplement 3, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, November 2011 2.1.3 NEI 12-10 (Rev. 1), Guideline for Developing a Licensee Protective Action Recommendation Procedure Using NUREG-0654, Supplement 3, August 2014 2.1.4 EPA-400-R-92-001, May 1992, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents 2.1.5 KLD Associates, Inc., Davis-Besse Nuclear Power Station Development of Evacuation Time Estimates, KLD TR-482, Rev. 2, October 2012 2.1.6 U.S. Food and Drug Administration, Federal Register, Vol. 47, No. 205, Oct. 22, 1982 2.1.7 Anno, George, Dore, and Michael: The Effectiveness of Sheltering as a Protective Action Against Nuclear Accidents Involving Gaseous Releases, EPA 520/I 001A, April 1978.

2.1.8 SAND 77-1725, Public Protection Strategies for Potential Nuclear Reactor Accidents - Sheltering Concepts with Existing Public and Private Structures.

2.1.9 NRC IE Information Notice No. 83-28: Criteria for Protective Action Recommendations for General Emergencies, dated May 4, 1983.

2.1.10 Regulatory Information Summary (RIS) 2002-16, Current Incident Response Issues 2.1.11 Regulatory Information Summary (RIS) 2003-12, NRC Regulatory Issue Summary 2003-12: Clarification of NRC Guidance for Modifying Protective Actions 2.1.12 Regulatory Information Summary (RIS) 2004-13, Consideration of Sheltering in Licensee's Range of Protective Action Recommendations 2.1.13 NEI Position Paper, Range of Protective Actions for Nuclear Power Plant Incidents, July, 2004.

4 RA-EP-02245 Revision 09 2.1.14 NRC Regulatory Issue Summary 2005-08, Endorsement of Nuclear Energy Institute (NEI) Guidance Range of Protective Actions for Nuclear Power Plant Incidents, June 6, 2005 2.2 Implementation 2.2.1 RA-EP-02110, Emergency Notification 2.2.2 NOP-LP-5022, Davis-Besse Unified RASCAL Interface (URI) Dose Assessment Software.

2.2.3 RA-EP-02520, Assembly and Accountability 2.2.4 RA-EP-02530, Evacuation 2.2.5 RA-EP-02620, Emergency Dose Control and Potassium Iodide Distribution 2.2.6 RA-EP-02720, Recovery Organization 3.0 DEFINITIONS 3.1 ALARA - As Low As Reasonably Achievable, means making every reasonable effort to maintain exposures to radiation as far below the dose limits in 10CFR20 as is practical and consistent with the purpose for which the licensed activity is undertaken.

3.2 COMMITTED DOSE EQUIVALENT (CDE) - The dose equivalent to organs or tissues that will be received from an intake of radioactive material by an individual during the 50-year period following the intake.

3.3 EMERGENCY PLANNING ZONE (EPZ) - The two zones that are established around a nuclear power station in which predetermined protective actions plans are needed.

3.3.1 The first zone has an approximate radius of 10 miles for the plume exposure pathway.

3.3.2 The second zone has an approximate radius of 50 miles for the ingestion exposure pathway.

3.4 EVACUATION DOSE - The dose that a potential evacuee would receive if he or she were openly exposed during the evacuation.

3.5 EVACUATION EXPOSURE PERIOD - The period during which those people being evacuated are exposed to the radioactive plume.

3.6 EXPOSURE TIME - That period of time during which the offsite population will be exposed to radiation as a result of an airborne radioactive release.

3.7 HOSTILE ACTION - An act toward DBNPS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that

5 RA-EP-02245 Revision 09 satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on DBNPS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA.

3.8 KI FOR THE GENERAL PUBLIC - Recommending potassium iodide (KI) for the general public is the responsibility of the State of Ohio Department of Health. The station will recommend to the State administrating KI in accordance with State procedures upon declaration of a General Emergency.

3.9 KNOWN IMPEDIMENTS TO EVACUATION - Those conditions that may affect the ability to evacuate a large portion of the population within the Emergency Planning Zone (such as inclement weather, roadway/bridge damage following a natural event). State or counties must have provided prior knowledge of these impediments to the utility prior to the Protective Action decision-making.

3.10 LAKE BREEZE - A meteorological condition that may occur on clear, sunny days. During a lake breeze, a radioactive release can travel inland, rise, reverse course in an overhead return flow, and then return to land in a convoluted path.

3.11 MINIMUM RADIOACTIVE RELEASE PROTECTIVE ACTION RECOMMENDATIONS (PAR) - At a minimum a PAR will be issued for Subarea 1, Subarea 12, and affected downwind subareas within five miles AND advise the general public to take KI in accordance with the Ohio Plan for Response to Radiation Emergencies at Licensed Nuclear Facilities.

3.12 MONITOR AND PREPARE - Guidance provided to the population within the Emergency Planning Zone, typically for those Sheltering in Place. This advices the public to maintain an awareness of the event through media broadcasts and subsequent Emergency Alert System messages. The public should make preparations for evacuation (or other protective actions) as directed.

3.13 OFFSITE - Any area outside the Owner Controlled Area surrounding Davis-Besse Nuclear Power Station.

3.14 RAPIDLY ESCALATING SEVERE ACCIDENT - This describes an accident in which there has been significant degradation of the reactor core, and because of the accident, the integrity of Containment is challenged. It is postulated that this accident will result in a significant release of radioactive material and subsequent dose to the public.

3.15 RELEASE - A release is defined as a radiological release attributable to the emergency event. Two levels of radiological release exist:

a. A minor unplanned release below levels that require offsite Protective Action Recommendations.
b. A release that requires offsite Protective Action Recommendations.

3.16 SAFETY PARAMETER DISPLAY SYSTEM (SPDS) - The SPDS is a group of graphic displays developed to assist with monitoring plant operations.

6 RA-EP-02245 Revision 09 3.17 SECTOR - One of the 16 areas bounded by radii 221/2 degrees apart into which the 10-mile EPZ is divided. Sectors are designated by the Letters A through P, excluding I and O.

Sector A is north, E is East, J is south, and N is west.

3.18 SHELTERING - The use of a structure for radiation protection from an airborne plume and from deposited radioactive material. A wood frame home without a basement is the assumed structure for sheltering in the Davis-Besse EPZ.

3.19 SHORT-TERM (PUFF) RELEASE - A controlled venting of radioactive gases (usually from Containment), sometimes referred to as a puff release. To be considered a short-term release, it must meet all of the following criteria:

  • Initiated by the operator, not the accident.
  • Performed in a controlled manner and able to be secured, if necessary
  • Monitored by effluent radiation monitors (e.g., Station Vent radiation monitors)
  • Short-term (< 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
  • Doses are less than EPA guidance 3.20 TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE) - The sum of the deep-dose equivalent (for external exposure) and the Committed Effective Dose Equivalent (for internal exposure).

4.0 RESPONSIBILITIES 4.1 The Emergency Director is responsible for directing protective actions for Station personnel and recommending protective actions to offsite officials for the Plume Exposure Pathway (10-mile EPZ).

4.2 The Dose Assessment Coordinator is responsible for collecting and analyzing offsite dose assessment data used to provide the basis for protective action recommendations.

5.0 INITIATING CONDITIONS Initiate this procedure when a declared emergency has the potential for an abnormal release of radioactivity.

7 RA-EP-02245 Revision 09 6.0 PROCEDURE Steps in this procedure may be performed simultaneously.

6.1 Onsite Protective Actions 6.1.1 The Emergency Director shall initiate the necessary actions to protect DBNPS personnel.

a. Evacuate personnel in accordance with RA-EP-02530, Evacuation.
b. Account for personnel in accordance with RA-EP-02520, Assembly and Accountability.
c. Distribute potassium iodide in accordance with RA-EP-02620, Emergency Dose Control and Potassium Iodide Distribution.
d. If Onsite Protective Actions are required due to a HOSTILE ACTION, refer to DB-OP-02544, Security Events Or Threats.

6.1.2 All supervisors shall ensure that appropriate safety and ALARA precautions are implemented.

8 RA-EP-02245 Revision 09 NOTE 6.2

  • ANY CONDITION THAT JUSTIFIES ISSUING AN OFFSITE PROTECTIVE ACTION REQUIRES A GENERAL EMERGENCY DECLARATION.
  • Offsite Protection Action Recommendations shall be made with initial notification of a General Emergency.
  • Davis-Besse will always recommend EVACUATION of Subarea 12 (Lake Erie) and when appropriate Subarea 10 (Wildlife area) due to lack of shelters in these areas.
  • A SHELTERING PAR will NOT be issued for any subarea in which an EVACUATION PAR has already been recommended.

6.2 Offsite Protective Actions 6.2.1 Refer to Attachment 1 to determine recommended Protective Action Recommendation.

CAUTION 6.2.2 Protective Action Recommendations once issued start in motion a sequence of events in the 10-mile emergency planning zone that, if modified, have the potential to cause confusion that may hamper the orderly implementation of protective actions for the general public.

6.2.2 Notify offsite agencies and the NRC of the PARs and the affected subareas using RA-EP-02110, Emergency Notification, and Initial Notification Form, DBEP-010.

a. IF these are revised PARs THEN DO NOT downgrade a previously issued PAR for a specific subarea until the conditions that caused the PAR to be issued are fully under control. The new PAR should include those subareas that were previously evacuated or sheltered and any new subareas.
b. IF a lake breeze is occurring, the wind direction is unknown, or the wind direction is from between 162° and 277°,

THEN inform the NRC that the release may enter Canadian territory.

c. IF TEDE doses are > 1,000 mrem or thyroid dose > 5,000 mrem are projected beyond 10 miles, THEN coordinate with state and county officials to determine appropriate PAR.
d. IF an Initial PAR was to Shelter a Subarea(s), and consultation with the State and County Emergency Management Agencies determine that it is safe to do so, THEN change the PAR from Shelter to Evacuate.

6.2.3 As Radiation Monitoring Team (RMT) data becomes available, compare it to dose projections and verify that Protective Action Recommendations are adequate.

9 RA-EP-02245 Revision 09 6.3 IF the General Emergency condition still exists, THEN continue to monitor radiological and meteorological conditions, AND repeat Steps 6.1 and 6.2 to ensure current PARs are adequate.

7.0 FINAL CONDITIONS Terminate this procedure when the Emergency Director, and offsite agencies determine that dose assessment and protective actions are no longer necessary.

NOTE: 8.0 All logs, documents and forms generated during the implementation of this procedure are to be retained and processed in accordance with RA-EP-02720, Recovery Organization.

8.0 RECORDS 8.1 The following records that are completed by this procedure shall be processed and retained as part of the Event Package in accordance with RA-EP-02720, Recovery Organization.

8.1.1 None

10 RA-EP-02245 Revision 09 THIS PAGE INTENTIONALLY LEFT BLANK

11 RA-EP-02245 Revision 09 ATTACHMENT 1: FLOWCHART FOR OFFSITE PROTECTIVE ACTION RECOMMENDATION DETERMINATION Page 1 of 2

12 RA-EP-02245 Revision 09 ATTACHMENT 1: FLOWCHART FOR OFFSITE PROTECTIVE ACTION RECOMMENDATION DETERMINATION Page 2 of 2 A - Evacuate B - Evacuate C - Shelter/Evacuate 2-Mile Radius & 2-Mile Radius & 2-Mile Radius &

Wind Direction 5-Miles Downwind 10-Miles Downwind 5-Miles Downwind From Protective Action Recommendation Unknown or Shelter N/A N/A 1, 2, 6 Lake Breeze Evacuate 1, 2, 6, 10, 12 ALL Subareas 10, 12 141° to 278° Shelter N/A N/A 1 Evacuate 1, 12 1, 12 12 Shelter N/A N/A 1, 6 279° to 286° Evacuate 1, 6, 12 1, 6, 7, 9, 12 12 Shelter N/A N/A 1, 6 287° to 293° Evacuate 1, 6, 12 1, 6, 7, 8, 9, 12 12 Shelter N/A N/A 1, 2, 6 294° to 330° Evacuate 1, 2, 6, 12 1, 2, 6, 7, 8, 9, 12 12 Shelter N/A N/A 1, 2, 6 331° to 005° Evacuate 1, 2, 6, 12 1, 2, 5, 6, 7, 8, 12 12 Shelter N/A N/A 1, 2, 6 006° to 013° Evacuate 1, 2, 6, 12 1, 2, 4, 5, 6, 7, 8, 12 12 Shelter N/A N/A 1, 2 014° to 020° Evacuate 1, 2, 12 1, 2, 4, 5, 12 12 Shelter N/A N/A 1, 2 021° to 065° Evacuate 1, 2, 12 1, 2, 3, 4, 5, 12 12 Shelter N/A N/A 1, 2 066° to 072° Evacuate 1, 2, 12 1, 2, 3, 4, 12 12 Shelter N/A N/A 1, 2 073° to 078° Evacuate 1, 2, 10, 12 1, 2, 3, 10, 12 10, 12 Shelter N/A N/A 1, 2 079° to 117° Evacuate 1, 2, 10, 12 1, 2, 3, 10, 11, 12 10, 12 Shelter N/A N/A 1 118° to 122° Evacuate 1, 10, 12 1, 3, 10, 11, 12 10, 12 Shelter N/A N/A 1 123° to 140° Evacuate 1, 10, 12 1, 10, 11, 12 10, 12 Once the PAR and subareas are selected GO TO Step 6.2.2

13 RA-EP-02245 Revision 09 ATTACHMENT 2: COMPARISON OF OFFSITE SECTORS AND SUBAREAS Page 1 of 1 R

Q P

N M

G L

H K

J

14 RA-EP-02245 Revision 09 COMMITMENTS None

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