ML21137A353

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Written Exam
ML21137A353
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 02/18/2021
From: Gregory Roach
NRC/RGN-III/DRS/OLB
To:
Roach G
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Davis Besse 1LOT21 NRC Written Exam Rev. AG1

1. Initial Plant Conditions:
  • DB-OP-02000, Section 3.0 IMMEDIATE ACTIONS have been completed
  • The crew has routed to DB-OP-02000, Section 4.0 SUPPLEMENTAL ACTIONS 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%
  • Uncompensated Pressurizer Level is 75 inches and slowly lowering Based on Current Plant Conditions, which of the following actions is required to be completed NEXT AND which procedure will be referred to for guidance?

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 Answer: C Explanation/Justification: This question meets the KA by asking which action/procedure will be taken based on given indications that will be checked during Vital System Status Verification.

SRO level based on ES 401 Attachment 2 E: knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures A. Incorrect: Emergency boration is not required for one stuck rod. Plausible since this is the first check (Step 4.2) after Symptoms Check. Plausible since Emergency boration is required for two or more stuck rods and CRD Malfunctions is referred to for a stuck rod.

B. Incorrect: In this scenario the TBVs would be placed in hand and throttled open to lower SG Pressure to seat the MSSV not the AVV.

Plausible because a MSSV could be the cause of #1 SG Pressure lowering. Attachment 2 is directed to be performed prior to checking for NNI Power available.

C. Correct: The given conditions do not meet the criteria for Specific Rule or Symptom mitigation. The candidate should realize the next required action listed is to Initiate AND Isolate SFRCS using MANUAL ACTUATION switches IAW DB-OP-02000 step 4.5 RNO, Loss of NNI X AC and refer to DB-OP-02532, Loss of NNI/ICS Power, for guidance.

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 D. Incorrect: The STBY MUP is not required to be started in this scenario. This abnormal procedure would not be addressed until step 4.19 of the supplemental actions. Also, if a MUP would need to be started (PZR Level < 40 inches), the MU Pump guidance that will be used is located in DB-OP-02000 SR3, ATT 1, or ATT 8. Plausible since, Annunciator 6-6-C SEAL INJ TOTAL FLOW is in alarm indicating a potential issue with the Makeup System and MU32 fails closed due to the loss of NNI X AC, which is also addressed in DB-OP-02512 Makeup and Purification System Malfunctions Sys # System Category KA Statement BW E02 Reactor Trip, EA2 - Ability to determine and interpret the following as they apply to Facility conditions and selection of

&E10 Stabilization, the (Vital System Status Verification): appropriate procedures during Recovery abnormal and emergency operations.

K/A# EA2.1 K/A Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

-DB-OP-02000 Step 4.5

-BASES AND DEVIATION DOCUMENT FOR DB-OP-02000

-DB-OP-01003 Step 6.5.3.a Question Source: New Level Of Difficulty: (1-5) 2 Question Cognitive Level: High 10 CFR Part 55 Content: 43.5 / 45.13 Objective: OPS-GOP-303-02K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 indications:
  • 6-5-C SEAL INJ LO
  • 1-3-D BUS C1 L/O
  • 4-2-E PZR LVL LO
  • Pressurizer Level is at 195 inches and lowering 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 Answer: B Explanation/Justification: This question meets the KA by asking which action/procedure will be taken next (ability to prioritize) based on given annunciators (Loss of a Makeup Pump due to a bus lockout).

SRO level based on ES 401 Attachment 2 E: assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed A. Incorrect: On a bus C1 lockout the running MUP and CCW pump will trip. The STBY CCW pump will auto start. This answer is plausible since RCP Cooling and Seal Injection are affected by this transient and this step will be performed later in the abnormal procedure.

B. Correct: On a bus C1 lockout, EDG 1 will auto start on UV. The Immediate Action is to emergency shutdown EDG 1 due to no cooling water. The next two steps in DB-OP-02512 are to isolate letdown which will be performed by the ATC and to perform ATT 1 which will be performed by the BOP.

C. Incorrect: Trip the Reactor, GO TO DB-OP-02000, RPS, SFAS, SFRCS Trip, or SG Tube Rupture is an IF AT ANY TIME step that will be performed if PZR level is <160 in. when Tave is at 582 degrees. Plausible since 4-2-E (<200 in.) is in alarm.

D. Incorrect: This step is a RNO step if both seal injection and CCW are lost. This is also the first step in DB-OP-02515 if the STBY CCW failed to auto start. This action should not be performed since it is stated in the stem that all systems responded as expected.

Sys # System Category KA Statement APE 22 Loss of Generic 2.4 Emergency Procedures / Plan Ability to prioritize and interpret the Reactor significance of each annunciator or alarm.

Coolant Makeup K/A# G2.4.45 K/A Importance 4.3 Exam Level SRO References provided to Candidate Technical

References:

DB-OP-02512, Makeup and Purification System Malfunctions pg 8, 10 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 41.10 / 43.5 / 45.3 /

45.12

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Answer: A Explanation/Justification: This question meets the KA by asking status of Absolute Position Indication for Groups 1 through 7 following restoration of power to busses E2 and F2 which are de-energized to remove power to the CRDMs when an ATWS occurs SRO level based on ES 401 Attachment 2 SRO only - unique to SRO position - SRO Task ID 334-005 0300 ACT AS EMERGENCY DIRECTOR IN THE CONTROL ROOM DURING AN EMERGENCY. Non-delegable Emergency Director Tasks such as making emergency classification shall NOT be delegated A. Correct (1) An UE is classified when an automatic or manual trip action does not shutdown the reactor and a subsequent action (deenergizing E2 and F2) is successful in shutting down the reactor.

(2) The rod indicating panel momentarily loses power while E2 and F2 are deenergized. Rod position indication returns when power is restored to E2 and F2.

B. Incorrect (1) Correct (2) Incorrect: Plausible since this would be correct if E2 and F2 remained deenergized.

C. Incorrect (1) Incorrect: Plausible since an Alert would be classified if deenergizing E2 and F2 did not shutdown the reactor (2) Correct D. Incorrect (1) Incorrect: Plausible since an Alert would be classified if deenergizing E2 and F2 did not shutdown the reactor (2) Incorrect: Plausible since this would be correct if E2 and F2 remained deenergized Sys # System Category KA Statement EPE 29 Anticipated EA2 - Ability to determine or interpret the following as they apply to Rod bank step counters and Rod Position Transient an ATWS: Indication Without Scram K/A# EA2.08 K/A Importance 3.5 Exam Level SRO References provided to Candidate Wallboard Technical

References:

CRD System Description SD-049 (Fig. 2.1-3)

DBRM-EMER-1500A SU6.1 (pg. 155-157)

DB-OP-06402 Attachment 4 (pg. 1)

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: Low 10 CFR Part 55 Content: 43.5 / 45.13 Objective: OPS-GOP-302 OPS-GOP-602-04K OPS-GOP-116-05K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 /> Answer: B Explanation/Justification: This question meets the KA based on ability to determine operability and/or availability of safety related equipment that mitigates Steam Line Rupture- Excessive Heat Transfer SRO level based on ES-401 Attachment 2, B SRO based on requires knowledge of TS Bases to determine requirements for Operability.

A. Incorrect - SFRCS Actuation Logic LCO 3.3.13 does not apply because the actuation channel terminates at the output relays. See B 3.3.13. Plausible because Main Steam Valve Control function is degraded. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is correct time for 3.3.13 Condition A.

B. Correct - EFW LCO 3.7.5 Condition A applies because each AFW Pump requires operable redundant steam supplies from each SG.

See B 3.7.5. Completion time is 7 days.

C. Incorrect - SFRCS Actuation Logic LCO 3.3.13 does not apply because the actuation channel terminates at the output relays. See B 3.3.13. 7 days is correct Completion Time for the correct LCO.

D. Incorrect - EFW LCO 3.7.5 Condition A Completion Time is 7 days. EFW is the correct LCO. Plausible because 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is the Completion Time for an AFW Train inoperable for a different reason.

Sys # System Category KA Statement BW E05 Steam Line Rupture Generic 2.2 Equipment Control Ability to determine operability and/or Excessive Heat Transfer availability of safety related equipment.

K/A# G2.2.37 K/A 4.6 Exam Level SRO Importance References provided to Candidate None Technical

References:

LCO 3.7.5, Bases 3.7.5 Question Source: Bank: DB1LOT15 Q79 Level Of Difficulty: (1-5)

Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: 41.7 / 43.5 / 45.12 Objective: OPS-GOP-437-02A

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

5. Initial conditions
  • Plant shutdown in progress
  • Reactor Power is 12%
  • Loss of Off-site 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, located 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, located 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, located 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, located on panel C5706, CTRM Center Console (AFW Control)

Panel

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 Answer: B Explanation/Justification: KA match based on candidate knowing the location of control room switches.

ES-401 Attachment 2 E: SRO only based on assessment of facility conditions and selection of appropriate procedure.

A. Incorrect - (1) is correct and (2) is plausible since other SFRCS reset switches are located on panel C5792N B. Correct - (1) The use of the Emergency Feedwater Pump is limited to beyond design bases events (e.g., loss of both Auxiliary Feedwater trains). Shift Managers permission under the provisions of 10CFR50.54(x) and (y) is required for any use of the Emergency Feedwater Pump that is not for beyond design bases response. (2) Reset switches HIS SP7AB and HIS SP7CB Are located on panel C5712, CTRM Right Console (MFW Control) Panel C. Incorrect - (1) is plausible since per attachment 5 the AFW header is preferred when feeding a dry steam generator (2) is correct D. Incorrect - (1) see answer B (2) is plausible since the AFW mimic and other AFW controls are located on panel C5706 Sys # System Category KA Statement APE 54 Loss of Main Generic 2.1 Conduct of Operations Ability to locate control room switches, Feedwater controls, and indications, and to determine that they correctly reflect the desired plant lineup.

K/A# G2.1.31 K/A Importance 4.3 Exam Level SRO References provided to Candidate Technical

References:

DB-OP-02000, Attachment 5 pg 1, 6 DBOP2000 bases and deviation pg 470 &

471 Question Source: Modified: Part 1 modified from DB 2020 NRC Q81 Level Of Difficulty: (1-5)

Part 2 New Question Cognitive Level: High 10 CFR Part 55 Content: 41.10 / 45.12 Objective: OPS-GOP-302-02K OPS-GOP-501-04K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 a Steam Generator overfill 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 a Steam Generator overfill event.

Answer: C Explanation/Justification: KA match based on candidate knowing availability of AFW CTRM controls (DC Powered) during a Station Blackout.

SRO level based on ES-401 Attachment 2 E: Assessment of Facility Conditions and Selection of Appropriate Procedures during normal, Abnormal, and Emergency Situations A. Incorrect: DB-OP-02521, Loss of AC Bus Power Sources is plausible since this is the procedure that used to direct Battery Load Shedding.

Part 2 of answer is correct.

B. Incorrect: DB-OP-02521, Loss of AC Bus Power Sources is plausible since this is the procedure that used to direct Battery Load Shedding.

Part 2 is also incorrect. Closing AF599 is plausible since this valve is closed to stop SG overfill events for other scenarios. It cannot be used in this case since it will be de-energized. AF6451, AFW 2 LEVEL CONTROL VALVE will fail open when depowered during performance of Attachment 2, Severe Battery Load Shed - In Plant Actions.

C. Correct: When C1 AND D1 remain deenergized, an Extended Loss of AC Power (ELAP) is declared. The CSRO will GO TO DB-OP-02700, Station Blackout. DB-OP-02700 will Initiate DB-OP-02704, Extended Loss of AC Power DC Load Management which will perform Selective Battery Load Shedding, for Loss of Power to the Battery Chargers Supplying BOTH DC MCC1 AND DC MCC2.

AF6451, AFW 2 LEVEL CONTROL VALVE will fail open when depowered during performance of Severe Battery Load Shed. Level will be controlled by adjusting pump speed using HIS521A, AFPT 2 GOVERNOR.

D. Incorrect: Part 1 is correct.

Part 2 is incorrect. Closing AF599 is plausible since this valve is closed to stop SG overfill events for other scenarios. It cannot be used in this case since it will be de-energized. AF6451, AFW 2 LEVEL CONTROL VALVE will fail open when depowered during performance of Attachment 2, Severe Battery Load Shed - In Plant Actions.

Sys # System Category KA Statement EPE 55 Station EA2 - Ability to determine or interpret the following as they apply to a Instruments and controls operable with only Blackout Station Blackout: dc battery power available K/A# EA2.04 K/A Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02000 Attachment 28 DB-OP-02700 pg 6 DB-OP-02704 Attacment 1 pg 3 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 43.5 / 45.13 Objective: OPS-GOP-S144-04K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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-02516 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%

Answer: D Explanation/Justification: KA match based on candidate knowing indications of a dropped rod using ex-core instrumentation and loop temperature.

SRO level based on ES-401 Attachment 2 E: Assessment of Facility Conditions and Selection of Appropriate Procedures A. Incorrect: Part 1 is plausible since a stuck rod during a power maneuver may cause NIs to slightly skew and bring in the same annunciators. A stuck rod in group 7 at 95% power will not cause Tave to lower. Part 2 is plausible since 60% is the max power level for a dropped rod to comply with T.S. 3.1.4.

B. Incorrect: Part 1 is incorrect. Part 1 is plausible since a stuck rod during a power maneuver may cause NIs to slightly skew and bring in the same annunciators. A stuck rod in group 7 at 95% power will not cause Tave to lower.

C. Incorrect: Part 2 is incorrect. Part 2 is plausible since 60% is the max power level for a dropped rod to comply with T.S. 3.1.4.

D. Correct: These are indications for a dropped rod. IAW DB-OP-02516 power is initially lowered to 50%. Reducing power to 50% RTP will allow margin to recover the dropped rod without exceeding TS 3.1.4 power limits.

Sys # System Category KA Statement 003 Dropped AA2 - Ability to determine and interpret the following as they apply to Dropped rod, using in-core/ex-core Control Rod the Dropped Control Rod: instrumentation, in-core or loop temperature measurements K/A# AA2.03 K/A Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02516 pg. 5, 12

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 43.5 / 45.13 Objective: OPS-GOP-116-01K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Answer: A Explanation/Justification: SRO ONLY Tech Spec application of required actions > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> KA Match: Requires knowledge of LCO associated with a loss of Intermediate Range NI3.

A. Correct. Part 1- Candidate must know the IR and PR contact alignment for the given power level, then determine the impact of the blown fuse on the source range detectors.

Part 2 Candidate must apply the correct TS Action B. Incorrect. Right impact Wrong TS action. Plausible if candidate applies TS actions for both SR being inoperable (energized above its top range).

C. Incorrect. Wrong impact. Plausible if candidate does not know the contact alignment or setpoints for re-energizing the source ranges.

Right TS action.

D. Incorrect. Wrong impact. Plausible if candidate does not know the contact alignment or setpoints for re-energizing the source ranges.

Wrong TS action. Plausible if candidate applies TS actions for both SR being inoperable (both de-energized but a TS note allows this).

Sys # System Category KA Statement APE 33 Loss of Generic 2.2 Equipment Control Knowledge of limiting conditions for Intermediate operations and safety limits.

Range K/A# G2.2.22 K/A Importance 4.7 Exam Level SRO References provided to Candidate Tech Specs 3.3.9 and Technical

References:

DB-OP-02505 Rev. 06 Att. 1 pages 34 & 35; 3.3.10 TS Bases page B 3.3.10-1 Rev. 1 Question Source: Bank: DB NRC 2013 Q84 Level Of Difficulty: (1-5)

Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 41.5 / 43.2 / 45.2 Objective: OPS-GOP-105-06K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

9. Given the following conditions:
  • The crew has entered DB-OP-02518 High Condenser Pressure
  • Reactor power has been lowered to 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.

Answer: B Explanation/Justification: SRO Only: The question requires the applicant to assess plant conditions and to know the content of procedures in order to select a required course of action.

KA Match: Requires ability to determine when to trip the turbine during a loss of condenser vacuum.

A. Incorrect: Plausible since this is the action that is taken when Cond. Pressure reaches 5.0 inches HgA AND Turbine load is >280 MWe B. Correct: IF AT ANY TIME Condenser Pressure is greater than 5.0 inches HgA, AND Turbine load is less than 280 MWe THEN IF Reactor Power is less than 40% power (ARTS), THEN trip the Turbine.

C. Incorrect: : Plausible since, IF AT ANY TIME Condenser Pressure is greater than 5.0 inches HgA, AND Turbine load is less than 280 MWe THEN IF Reactor Power is greater than or equal to 40% power (ARTS) THEN trip the Reactor AND GO TO DB-OP-02000 D. Incorrect: Plausible since these are the actions if condenser pressure reaches 10 inches HgA.

Sys # System Category KA Statement APE 51 Loss of AA2 - Ability to determine and interpret the following as they apply to Conditions requiring reactor and/or turbine Condenser the Loss of Condenser Vacuum: trip Vacuum K/A# AA2.02 K/A Importance 3.9 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02518 pg. 8 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: Low 10 CFR Part 55 Content: 43.5 / 45.13 Objective: OPS-GOP-118-09K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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.

Answer: D Explanation/Justification: SRO Only: Facility Operating Limitations in the Technical Specifications and Their Bases (application of required actions)

KA Match: Requires ability to apply TRM action for a loss of a SCM Meter.

A. Incorrect: IAW TNC 8.3.7 Table 8.3.7-1 only one channel is required. This answer is plausible if the candidate misinterpreted the table.

B. Incorrect: SCM Meters are PAM Instrumentation, but not listed on TS 3.3.17 Table 3.3.17-1.

C. Incorrect: SCM Meters are PAM Instrumentation, but not listed on TS 3.3.17 Table 3.3.17-1. IAW TNC 8.3.7 Table 8.3.7-1 only one channel is required. This answer is plausible if the candidate misinterpreted these tables.

D. Correct: IAW TNC 8.3.7 Table 8.3.7-1 only one channel is required and SCM Meters are not listed on TS 3.3.17 Table 3.3.17-1.

Sys # System Category KA Statement BW E03 Inadequate Generic 2.2 Equipment Control Ability to apply Technical Specifications for a Subcooling system.

Margin K/A# G2.2.40 K/A Importance 4.7 Exam Level SRO References provided to Candidate TNC 8.3.7 and TS 3.3.17 Technical

References:

TNC 8.3.7 and TS 3.3.17 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: Low 10 CFR Part 55 Content: 41.10 / 43.2 / 43.5 /

45.3 Objective: OPS-GOP-400

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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.

Answer: A Explanation/Justification: SRO Only: ES401 Att. 2 item II. E Page 7 first bullet. Similar to question example from page 12.

A. Correct - Flow Balancing HPI is required during single train operation to protect against an HPI Line Break to ensure at least one HPI injection line flow is reaching the core. SRO ONLY since it requires the candidate to select the appropriate procedure attachment to mitigate the event.

B. Incorrect - Long term Boron dilutions is required when RCS temperatures are less than 333 ºF. At higher temperatures, the boron in the RCS will not precipitate out of solution. As a result, Long Term Boron Dilution is not required for these plant conditions.

C. Incorrect - HPI Alternate Minimum Recirc is required when BWST level is being reduce at less than 2 foot per hour. At higher flow rates, the RCS will not repressurize above the shutoff head of the HPI Pump. As a result, HPI Alternate Minimum Recirc Flow is not required for these plant conditions.

D. Incorrect - LPI Pump Discharge is required when a single LPI train is not available. Although no LPI flow exists in this scenario, LPI flows are consistent with the current Plant conditions. As a result, cross connecting LPI discharge is not required.

Sys # System Category KA Statement 006 Emergency A2 Ability to (a) predict the impacts of the following Improper discharge pressure Core Cooling malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR:

K/A# A2.04 K/A Importance 3.8 Exam Level SRO References provided to Candidate Steam Tables Technical

References:

DB-OP-02000 Attachment 11, page 1 Question Source: Bank DB 2020 NRC Q87 Level Of Difficulty: (1-5) 3 Question Cognitive Level: High 10 CFR Part 55 Content: 41.5 / 45.5 Objective: OPS-GOP-300

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Power Operated Relief Valve (PORV) failure D. Rod withdrawal transient from subcritical condition Answer: D Explanation/Justification: SRO Only: Facility Operating Limitations in the Technical Specifications and Their Bases A. Incorrect. No credit for RCS Pressure Safety Limit for Make-up flow events. Plausible because the Makeup and Purification System Malfunction is credited for other accident analyses and Make-up Flow operation is described in the Tech Spec Bases for RCS Pressure Safety Limits.

B. Incorrect. No credit for RCS Pressure Safety Limit for Pressurizer Spray Valve failure. Plausible because a purpose of the Pressurizer Spray Valve is to mitigate Reactor Coolant System pressure transients and the Pressurizer Spray Valve operation is described in the Tech Spec Bases for RCS Pressure Safety Limits.

C. Incorrect. No credit for RCS Pressure Safety Limit for Pressurizer PORV failure. Plausible because a purpose of the PORV is to mitigate Reactor Coolant System pressure transients and the PORV operation is described in the Tech Spec Bases for RCS Pressure Safety Limits.

D. Correct. The RCS Pressure Safety Limit credits the Pressurizer Code Safeties and RPS during a rod withdrawal transient from a subcritical condition. The analysis assumes full RCS flow but no heat transfer out of primary system to maximize system conditions.

Sys # System Category KA Statement 012 Reactor Generic 2.2 Equipment Control Knowledge of the bases in Technical Protection Specifications for limiting conditions for operations and safety limits.

K/A# G2.2.25 K/A Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

TS Bases B2.1.2 Reactor Coolant System (RCS) Pressure SL Question Source: New Level Of Difficulty: (1-5) 4 Question Cognitive Level: Low 10 CFR Part 55 Content: 41.5 / 41.7 / 43.2 Objective: OPS-GOP-433-04K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 what action will the CSRO direct FIRST?

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 Answer: C Explanation/Justification: SRO Only: The question requires the applicant to assess plant conditions and to know the content of procedures in order to select a required course of action.

KA Match: SP6A, SG 2 Main Feedwater Control valve is an AOV.

A. Incorrect: Plausible since both Main Feed Water Control valves should reposition when the RCP is tripped. Second part is correct B. Incorrect: Plausible since both Main Feed Water Control valves should reposition when the RCP is tripped. Second part is plausible since the SFRCS will initiate due to SG high and low levels.

C. Correct: IAW DB-OP-02515 Attachment 1 (CSRO directed) Verify proper Feedwater flow ratios of 2.4 to 1. (Feedwater flow should be approximately 5.74 MPPH to the SG with 2 RCPs vs. 2.38 MPPH to the SG with one RCP at 72% power - Approximately 70% to 30%

ratio for other power levels). SP6A will need to open to feed SG2 since it is providing more steam flow when a RCP in loop 1 is tripped.

D. Incorrect: First part correct. Second part same as answer B.

Sys # System Category KA Statement 059 Main A2 - Ability to (a) predict the impacts of the following malfunctions or Failure of feedwater regulating valves Feedwater operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

K/A# A2.12 K/A Importance 3.4 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02515 Attachment 1 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 41.5 / 43.5 / 45.3 /

45.13 Objective: OPS-GOP-115-06K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Off-site power concurrent with a loss of EDG 2 D. A loss of Off-site power concurrent with a loss of the SBODG Answer: A Explanation/Justification: SRO only - unique to SRO position - Emergency Director Tasks such as making emergency classification and making offsite protective action recommendations shall NOT be delegated.

A. Correct - Since Bus C1 is Tagged out and D1 Locks out, Both C1 and D1 are de-energized. With the Plant Defueled, this is an Alert per CA2.1.

B. Incorrect - plausible if the operator forgets the Off-site power is included in the power sources that can be credited. These plant conditions would be an Unusual Event per CU2.1.

C. Incorrect - plausible if the operator uses the Hot Mode EALs. These plant conditions would be an Unusual Event per CU2.1 since power from SBODG to D1 would be available.

D. Incorrect - plausible if the operator uses the Hot Mode EALs. These plant conditions would be an Unusual Event per CU2.1 since power from EDG 2 to D1 would be available.

Sys # System Category KA Statement 062 AC Electrical Generic 2.4 Emergency Procedures / Plan Knowledge of the emergency action level Distribution thresholds and classifications.

K/A# G2.4.41 K/A Importance 4.6 Exam Level SRO References provided to Candidate DB-EMER-1500B, EAL Technical

References:

DB-EMER-1500B, EAL Wallboard Wallboard DBRM-EMER-1500A CA2.1 Question Source: Bank: 97009 SRO Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 41.10 / 43.5 / 45.11 Objective: OPS-GOP-602-04K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

15. Initial conditions:
  • Plant is at 100% Power
  • Loss of Off-site Power (LOOP)
  • EDG 2 trips Which of the following correctly states the immediate action of the Control Room Operator to the LOOP and which procedure which will be entered NEXT?

A. Verify EDG 1 output breaker, AC 101, is open GO TO DB-OP-02521 Loss of AC Bus Power Sources B. Verify bus C1 feeder breaker, AC 110, is open GO TO DB-OP-02521 Loss of AC Bus Power Sources C. Verify EDG 1 output breaker, AC 101, is open GO TO DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture D. Verify bus C1 feeder breaker, AC 110, is open GO TO DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture Answer: D Explanation/Justification: NUREG-1021, Revision 11, Section ES-401 Attachment 2, step II.E. SRO based on procedural priority determination by knowing Loss of all AC is not one of the abnormal procedures that would take precedence over DB-OP-02000. K/A match based on requiring knowledge of the effect of a LOOP while EDG is loaded during testing (causes a highly loaded condition) and the correct procedure to enter to mitigate.

A. Incorrect - Plausible since if an SFAS Level 2 with no LOOP, the EDG output breaker, AC 101 will open and DB-OP-02521 will be the procedure to direct recovery of a 4160V essential bus in the event the SBODG cannot. 13.8 bus is de-energized on LOOP B. Incorrect - Plausible since DB-OP-02521 will be the procedure to direct recovery of a 4160V essential bus in the event the SBODG cannot. DB-OP-02000 Specific Rule 6 directs entry into this AB if the initial attempts to start the EDG(s) have failed. At this point, DB-OP-02521 is performed in parallel with the remainder of the actions in DB-OP-02000. There are some abnormal procedures that will take precedence over DB-OP-02000 C. Incorrect - Plausible since if an SFAS Level 2 with no LOOP, the EDG output breaker, AC 101 will open. 13.8 bus is de-energized on LOOP D. Correct - Per DB-OP-6316 Step 2.2.16, If a LOOP occurs, with no SFAS, the EDG may overload reaching 3300 to 3500 KW. If a LOOP occurs, the local and CTRM operators are directed to immediately open AC110. The plant will trip due to the LOOP, requiring entry to DB-OP-02000.

Sys # System Category KA Statement 064 Emergency A2 - Ability to (a) predict the impacts of the following Operating unloaded, lightly loaded, and Diesel malfunctions or operations on the ED/G system; and highly loaded time limit Generator (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

K/A# A2.06 K/A Importance 3.3 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-06316 steps 2.2.16, 3.3.5 DB-OP-01003 step 6.5.2 Question Source: Modified DB1LOT16 NRC Q90 Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 41.5 / 43.5 / 45.3 /

45.13 Objective: SYS-406, GOP-500

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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

Instrumentation is not met. Condition A has been entered for RPS CH 1 being inoperable and required action is complete.

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

Which of the following is required IAW Limiting Condition for Operation (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 CH 1 and CH 4, place CH 4 in Bypass within one hour.

C. Enter Condition B for RPS CH 1 and CH 4, verify CH 4 is tripped within one hour.

D. Enter Condition A for RPS CH 4 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 />.

Answer: C Explanation/Justification:

A. Incorrect: Plausible since separate condition entry is allowed for other instrumentation tech specs. (Ex. T.S. 3.3.5)

B. Incorrect: Plausible since placing a channel in Bypass or Trip is the action when one channel is inoperable.

C. Correct: Since the reactor did not trip when RPS CH 4 tripped, CH 1 must have been in Bypass. Condition B must be entered for two channels inoperable.

D. Incorrect: Plausible since this is the action for Condition D Sys # System Category KA Statement 015 Nuclear Generic 2.2 Equipment Control Ability to analyze the effect of maintenance Instrumentation activities, such as degraded power sources, on the status of limiting conditions for operations.

K/A# G2.2.36 K/A Importance 4.2 Exam Level SRO References provided to Candidate Technical

References:

LCO 3.3.1 Question Source: Mod DB1LOT13 Q92 Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 41.10 / 43.2 / 45.13 Objective: OPS-GOP-410

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Answer: C Explanation/Justification: KA match based on candidate knowing and predicting operations of the ITM system during Core Damage scenarios. If the RCS P-T point is superheated and in Region 2, then ICC conditions exist, but they are not serious enough to cause immediate Core damage.

SRO level based on ES-401 Attachment 2 E: Assessment of Facility Conditions and Selection of Appropriate Procedures during normal, Abnormal, and Emergency Situations A. Incorrect. Incore thermocouples have an instrument error of +/-24.2°F and it is possible to be saturated and be slightly to the right of the saturation curve. Plausible because 20°F SCM is required to establish subcooled conditions.

B. Incorrect. If only one channel is available, then a total of three or more working incore detectors displaying a NEG MARGIN confirms that superheated conditions exist C. Correct. Due to the nature of the actions that will be taken when in Section 9, Inadequate Core Cooling, DO NOT route to Section 9 unless superheated conditions actually exist (Incore Thermocouple trending away (increasing) from the saturation line at a value greater than the maximum instrument error).

D. Incorrect. If both RCS pressure inputs are <150 psig, then the SCM Meter display is blank. Plausible because a blank indicator occurs when RCS pressure conditions result in errors in the SCM meters accurately determining actual status of RCS conditions for subcooled, saturated, superheated conditions.

Sys # System Category KA Statement 017 In Core A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the ITM Core damage Temperature system; and (b) based on those predictions, use procedures to correct, control or mitigate Monitor the consequences of those malfunctions or operations:

K/A# A2.02 K/A Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

Bases and Deviation Document for DB-OP-02000 R22 Step 5.13, 9.13 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: Low 10 CFR Part 55 Content: 41.5 / 43.5 / 45.3 / 45.5 Objective: OPS-GOP-300-08K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Answer: C Explanation/Justification: KA Match: Requires Knowledge of Main Turbine surveillance procedures SRO Level: Facility Operating Limitations in the Technical Specifications A. Incorrect (1) False: DB-SS-04159 does not require the ARTS TURBINE-GEN Bypass Switch placed in the BYPASS position. Plausible since a failed ETD could cause EHC pressure to lower to the ARTS Trip setpoint.

(2) True: According to LCO 3.3.16 Table 3.3.16-1 only 3 channels are required to be operable when >45% power therefore LCO 3.3.16 is met.

B. Incorrect (1) False: DB-SS-04159 does not require the ARTS TURBINE-GEN Bypass Switch placed in the BYPASS position. Plausible since a failed ETD could cause EHC pressure to lower to the ARTS Trip setpoint.

(2) False: According to LCO 3.3.16 Table 3.3.16-1 only 3 channels are required to be operable when >45% power therefore LCO 3.3.16 is still being met. Plausible since 4 channels of RPS and SFAS are required to be operable in Mode 1.

C. Correct (1) True: During the performance of DB-SS-4150 the ARTS TURBINE-GEN Bypass Switch is required to be placed in the BYPASS position.

(2) True: According to LCO 3.3.16 Table 3.3.16-1 only 3 channels are required to be operable when >45% power therefore LCO 3.3.16 is met.

D. Incorrect (1) True: During the performance of DB-SS-4150 the ARTS TURBINE-GEN Bypass Switch is required to be placed in the BYPASS position.

(2) False: According to LCO 3.3.16 Table 3.3.16-1 only 3 channels are required to be operable when >45% power therefore LCO 3.3.16 is still being met. Plausible since 4 channels of RPS and SFAS are required to be operable in Mode 1.

Sys # System Category KA Statement 045 Main Turbine Generic 2.2 Equipment Control G2.2.12: Knowledge of surveillance Generator procedures.

K/A# G2.2.12 K/A Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

LCO 3.3.16, DB-SS-04150 pg. 6 (ex)

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 Question Source: New Level Of Difficulty: (1-5)

Question Cognitive Level: Low 10 CFR Part 55 Content: 41.10 / 45.13 Objective: OPS-GOP-434

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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.

Answer: B Explanation/Justification: SRO only ES401 Att. 2, Section 2 item A third bullet. The SRO is required to know the content of the administrative procedures related to shift staffing and the Technical Specification requirements. The actions to restore shift staffing are a SRO responsibility.

SRO Task ID 336-005-03-0300 CALL IN ADDITIONAL PERSONNEL/SITE RESOURCES AS NECESSARY KA Match: Requires knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements (Change in Medical Status)

A. Incorrect - Plausible since non-licensed individuals, that have declared unfitness due to consumption prescription medications, can be called in for work with extra oversite IAW NOP-LP-1002, Fitness For Duty Program.

B. Correct - Licensed operators are required to formally report these conditions prior to assuming licensed duties. It is the Licensed Operators responsibility to communicate with the Health Center to ensure compliance and to track his/her medical information to the NRC to maintain their Operators License. In the event the shift crew composition is less than the minimum required, Immediately callout the required personnel needed to return to a minimum functional shift complement.

C. Incorrect -Plausible Shift crew composition may be one less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Technical Requirements Manual (TRM)/Operations Requirements Manual (ORM)/Licensing Requirements Manual (LRM) Specifications for a period of time not to exceed-two hours in order to accommodate unexpected absence of on-duty shift crew members, provided immediate action is taken to restore the shift crew composition to within the minimum requirements. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crew person being late or absent.

D. Incorrect - Plausible since there are other 30 day requirements listed in the Fitness for Duty program and Medical Condition requirements.

Sys # System Category KA Statement N/A N/A Generic 2.1 Conduct of Operations Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 K/A# 2.1.4 K/A Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

NOP-OP-1002 4.1.13, 4.13.2.12 (pg. 22, 69)

DBBP-DBNA-0001 pg. 9, 10 Question Source: New Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 41.10 / 43.2)

Objective: OPS-GOP-485-06K OPS-GOP-501-05K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Answer: C Explanation/Justification: SRO and KA match based on knowledge of SRO responsibilities during EOP usage A. 1. Incorrect - plausible since the shift manager has overall license responsibility for the plant
2. Correct B. 1. Incorrect - plausible since the shift manager has overall license responsibility for the plant
2. Incorrect - plausible since Transient crew briefs are normally led by the Command SRO and Transient crew briefs are used to ensure the crew understands overall plant status, anticipates the direction to take, discusses contingencies, establishes priorities and owners, and solicits input or questions from crew members.

C. 1. Correct -Step 4.10.2.11 The Command SRO shall clearly announce to the crew when entering or exiting AOPs/ONIs/EOPs, and when transitioning to different sections/flow charts of AOPs/ONIs/EOPs. Crew Updates should be used for these announcements.

2. Correct - Step 4.10.5.7 Crew updates should be used for entering or exiting AOP/ONIs/EOPs, and when transitioning to different sections/flow charts of AOP/ONIs/EOPs.

D. 1. Correct

2. Incorrect - plausible since Transient crew briefs are normally led by the Command SRO and Transient crew briefs are used to ensure the crew understands overall plant status, anticipates the direction to take, discusses contingencies, establishes priorities and owners, and solicits input or questions from crew members Sys # System Category KA Statement N/A N/A Generic 2.1 Conduct of Operations Ability to manage the control room crew during plant transients.

K/A# 2.1.6 K/A Importance 4.8 Exam Level SRO References provided to Candidate None Technical

References:

NOP-OP-1002 steps 4.10.2.11 and 4.10.5.7 Question Source: New Level Of Difficulty: (1-5) 2 Question Cognitive Level: Low 10 CFR Part 55 Content: 41.10 / 43.5 / 45.12 / 45.13 Objective: OPS-GOP-501-04K OPS-GOP-300-06K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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.

Answer: B Explanation/Justification:

A. Incorrect. The surveillance requirement is for a complete cycle. Opening the valve would ONLY meet half of a cycle. If the candidate believed that the other half was satisfactorily performed earlier, then the candidate would select this choice. Since maintenance was on the breaker the valve must be again cycled through a complete cycle (open and closed)

B. Correct. The surveillance requirement is for a complete cycle. Since maintenance was on the breaker the valve must be again cycled through a complete cycle (open and closed)

C. Incorrect. Since maintenance was on the breaker the valve must be again cycled through a complete cycle (open and closed). The LCO addresses both the PORV and the block valve, but maintenance was only performed on the block valve. NO requirement to perform any surveillance activities for the PORV.

D. Incorrect. Right actions for the block valve. Wrong actions for the PORV. The LCO addresses both the PORV and the block valve, but maintenance was only performed on the block valve. NO requirement to perform any surveillance activities for the PORV.

Sys # System Category KA Statement N/A N/A Generic 2.2 Equipment Control Knowledge of pre- and post-maintenance operability requirements.

K/A# 2.2.21 K/A Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

Technical Specification 3.4.11 SR 3.4.11.1; DB-PF-03811 R33 Acceptance Criteria

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 Question Source: Bank: BV2LOT6 Q96 Level Of Difficulty: (1-5)

Question Cognitive Level: High 10 CFR Part 55 Content: 41.10 / 43.2 Objective: OPS-GOP-501-12K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

22. An overhead Annunciator in the Control Room is determined to be a nuisance alarm due to equipment conditions. The Shift Manager has determined that the Annunciator Window should be disabled.

Which one of the following procedures will be utilized to disable the annunciator and what type of tag will be used when it is removed?

1. DB-OP-06411 Station Annunciator Operating Procedure
2. NOP-OP-1002 Conduct of Operations Procedure
3. Maintenance Information Tag
4. Red Danger Tag A. 1 and 3 B. 1 and 4 C. 2 and 3 D. 2 and 4 Answer: A Explanation/Justification: Meets the requirements of the SRO only ES401 Att. 2,Section II .C page 6 third bullet. The SRO is required to know the administrative requirements for disabling annunciators, which is considered a temporary configuration change.

A. Correct Disabling an Annunciator Window is directed using DB-OP-06411, Station Annunciator Procedure Section 4.5 A Maintenance Information Tag is directed to be placed on the alarm card.

B. Incorrect Part 1 is correct.

A Clearance is not necessary or directed to perform this activity. Red Danger Tag is plausible since they can be used to implement plant modifications when needed for safety requirements. Also, plausible to use OPS Only Clearance for equipment control per NOP-OP-1001 Clearance and Tagging Program Section 4.10. At one time, Danger Tags were required to ensure equipment remained OOS for all engineering changes until the new or modified equipment was ready to be turned over to operations.

C. Incorrect -

Disabling an Annunciator Window is directed using DB-OP-06411, Station Annunciator Procedure.

Plausible since Conduct of Operations provides guidance for nuisance alarms.

Part 2 is correct.

D. Incorrect -

Disabling an Annunciator Window is directed using DB-OP-06411, Station Annunciator Procedure.

Plausible since Conduct of Operations provides guidance for nuisance alarms.

A Clearance is not necessary or directed to perform this activity. Red Danger Tag is plausible since they can be used to implement plant modifications when needed for safety requirements. Also, plausible to use OPS Only Clearance for equipment control per NOP-OP-1001 Clearance and Tagging Program Section 4.10. At one time, Danger Tags were required to ensure equipment remained OOS for all engineering changes until the new or modified equipment was ready to be turned over to operations.

Sys # System Category KA Statement

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 N/A N/A Generic Knowledge of the process for controlling temporary design changes K/A# 2.2.11 K/A Importance 3.3 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-06411 Section 4.5 R28 Question Source: Modified from DB 2020 NRC Exam Q96 Level Of Difficulty: (1-5)

Question Cognitive Level: Low 10 CFR Part 55 Content: 41.10 / 43.3 / 45.13 Objective: GOP-504

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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.

Answer: D Explanation/Justification: K/A Match: The SRO is responsible for authorizing release permits and know the requirements if equipment becomes unavailable.

A. Incorrect. Plausible if the candidate believes the tank being discharged will pass thru the storm sewer FE and that having Clean RE 1770A operable meets the one RM channel operable requirement.

B. Incorrect. Storm sewer FE is not required for this discharge flowpath. Independent actions are correct.

C. Incorrect. These are the correct compensatory actions for the liquid waste flow indicator being out of service D. Correct. IAW ODCM Rev. 30 Table 2-1 pages 19 and 20.

Sys # System Category KA Statement N/A N/A Generic 2.3 Radiation Control Ability to approve release permits.

K/A# 2.3.6 K/A Importance 3.8 Exam Level SRO References provided to Candidate ODCM Rev. 37 Table 2-1 Technical

References:

ODCM Rev. 37 Table 2-1 pg 19, 20 and 2-2 pages 19 thru 22 Question Source: Bank: DB 2016 NRC Q99 Level Of Difficulty: (1-5) 3

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 Question Cognitive Level: High-Application 10 CFR Part 55 Content: 41.13 / 43.4 / 45.10 Objective: OPS-GOP-521-04K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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) Davis Besse Nuclear Power Plant Initial Notification Form (DBEP-010) telefaxed C. (1) 1134 (2) State and County notifications initiated D. (1) 1134 (2) Davis Besse Nuclear Power Plant Initial Notification Form (DBEP-010) telefaxed Answer: A Explanation/Justification: The KA is matched because the candidate must demonstrate knowledge of SRO responsibilities in emergency plan implementation.

The question is at the SRO level because the candidate must demonstrate reporting requirements. This is a job of the SRO only and therefore is not RO level knowledge.

A. Correct

1. Notifications must be initiated within 15 minutes of declaration of an emergency. If a higher classification is made prior to transmitting an event notification, then notification for the higher classification can supersede the previous event notification, provided that it can be performed within the 15-minute timeframe of the previous event. The UE was declared at time 1115, therefore notifications must be initiated by 1130.
2. The clock stops and is documented when the initial notification of the state and county representatives was initiated.

B. Incorrect

1. Correct As described for A. 1.
2. Wrong: Plausible since, Telefax the completed FENOC Nuclear Power Plant Initial Notification Form (DBEP-010) using the fax machine Group Tx key labeled INF FORM. Is the next step after the initial notification is made.

C. Incorrect

1. Wrong: Plausible since this would be the latest time to initiate the notification if the UE was not declared first.
2. Correct: As described for A. 2.

D. Incorrect

1. Wrong: As described for B. 2
2. Wrong: As described for C. 1.

Sys # System Category KA Statement

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 N/A N/A Generic 2.4 Emergency Procedures/Plan Knowledge of SRO responsibilities in emergency plan implementation. (CFR:)

K/A# 2.4.40 K/A Importance 4.5 Exam Level SRO References provided to Candidate None Technical

References:

RA-EP-01700, RA-EP-02110 Question Source: Bank TMI 2014 NRC Q100 Level Of Difficulty: (1-5)

Question Cognitive Level: High-Comprehension/Analysis 10 CFR Part 55 Content: 41.10 / 43.5 / 45.11 Objective: OPS-GOP-130-07K

Davis Besse 1LOT21 NRC Written Exam Rev. AG1

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 Answer: B Explanation/Justification: The KA is matched because the candidate must demonstrate knowledge of emergency plan protective action recommendations.

SRO only - unique to SRO position - SRO Task ID 334-005-05-0300 ACT AS EMERGENCY DIRECTOR IN THE CONTROL ROOM DURING AN EMERGENCY. Non-delegable Emergency Director Tasks such as making emergency classification and making offsite protective action recommendations shall NOT be delegated.

A. Incorrect. GE Declared-Yes, Initial Par-Yes, Loss of CTMT Fission Product Barrier and CT Rad Mon > Table F2 CT Pot Loss-No, Hostile Action OR Known Impediment Exists-NO, Release in Progress-Yes, Short Term-No, Dose Projections >Limits-No, Misreads Table for Wind Direction Column A (006°-013°)

B. CORRECT: GE Declared-Yes, Initial Par-Yes, Loss of CTMT Fission Product Barrier and CT Rad Mon > Table F2 CT Pot Loss-No, Hostile Action OR Known Impediment Exists-NO, Release in Progress-Yes, Short Term Release-No, Dose Projections >Limits-No, Column A (014°-

020°)

C. Incorrect. GE Declared-Yes, Initial Par-Yes, Loss of CTMT Fission Product Barrier and CT Rad Mon > Table F2 CT Pot Loss-Yes, Column B (014°-020°)

D. Incorrect. GE Declared-Yes, Initial Par-Yes, Loss of CTMT Fission Product Barrier and CT Rad Mon > Table F2 CT Pot Loss-No, Hostile Action OR Known Impediment Exists-NO, Release in Progress-Yes, Short Term Release-YES Column C (014°-020°)

Sys # System Category KA Statement N/A N/A Generic Knowledge of emergency plan protective action recommendations.

Davis Besse 1LOT21 NRC Written Exam Rev. AG1 K/A# 2.4.44 K/A Importance 4.4 Exam Level SRO References provided to Candidate RA-EP-02245 R9 and Technical

References:

RA-EP-02245 Rev. 9 Attachment 1 (2 pgs)

Wallboard RA-EP-01900 Rev Question Source: Modified DB 2013 NRC Q100 Level Of Difficulty: (1-5) 3 Question Cognitive Level: High Application 10 CFR Part 55 Content: 10 CFR: 55.43(b)(4)

Objective: OPS-GOP-604-05K

Davis Besse 2021 Initial License Examination Question References Question 3: Emergency Action Level Wall Boards Question 8: Technical Specifications 3.3.9 and 3.3.10 Question 10: Technical Specification 3.3.17 and Technical Requirements Manual TNC 8.3.7 Question 11: Steam Tables Question 14: Emergency Action Level Wall Boards Question 23: Offsite Dose Calculation Manual Tables 2-1 and 2-2 Question 25: Emergency Action Level Wall Boards and RA-EP-02245

DB1LOT21 NRC Written Exam ANSWER KEY NAME:_____________________________

SSN (Last 4) :________________________

1. A B D D 26. A B C D 51. A B C D 76. A B C D
2. A D C D 27. A B C D 52. A B C D 77. A B C D
3. D B C D 28. A B C D 53. A B C D 78. A B C D
4. A D C D 29. A B C D 54. A B C D 79. A B C D
5. A D C D 30. A B C D 55. A B C D 80. A B C D
6. A B D D 31. A B C D 56. A B C D 81. A B C D
7. A B C D 32. A B C D 57. A B C D 82. A B C D
8. D B C D 33. A B C D 58. A B C D 83. A B C D
9. A D C D 34. A B C D 59. A B C D 84. A B C D
10. A B C D 35. A B C D 60. A B C D 85. A B C D
11. D B C D 36. A B C D 61. A B C D 86. A B C D
12. A B C D 37. A B C D 62. A B C D 87. A B C D
13. A B D D 38. A B C D 63. A B C D 88. A B C D
14. D B C D 39. A B C D 64. A B C D 89. A B C D
15. A B C D 40. A B C D 65. A B C D 90. A B C D
16. A B D D 41. A B C D 66. A B C D 91. A B C D
17. A B D D 42. A B C D 67. A B C D 92. A B C D
18. A B D D 43. A B C D 68. A B C D 93. A B C D
19. A D C D 44. A B C D 69. A B C D 94. A B C D
20. A B D D 45. A B C D 70. A B C D 95. A B C D
21. A D C D 46. A B C D 71. A B C D 96. A B C D
22. D B C D 47. A B C D 72. A B C D 97. A B C D
23. A B C D 48. A B C D 73. A B C D 98. A B C D
24. D B C D 49. A B C D 74. A B C D 99. A B C D
25. A D C D 50. A B C D 75. A B C D 100. A B C D
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