ML22143B002

2019 Deac Ile Administered Written Exam

ML22143B002
Person / Time
Site:	Duane Arnold
Issue date:	04/08/2019
From:	Randy Baker NRC/RGN-III/DRS/OLB
To:	NextEra Energy Duane Arnold
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ML17214A863	List: ML22143A999 ML22143B000 ML22143B001 ML22143B002 ML22143B003
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PDA 19-1 LOIT NRC Written Examination Student Reference Handouts RO References Power/Flow Map Core Flow vs. Core plate D/P EHC Logic Control System Drawing SRO References LCO 3.2.2 Minimum Critical Power Ratio (MCPR)

LCO 3.4.10 Reactor Steam Dome Pressure LCO 3.5.1 ECCS - Operating TRM 3.5.1 Drywell Spray System LCO 3.7.3 Emergency Service Water System LCO 3.7.7 The Main Turbine Bypass System LCO 3.8.1 AC Sources - Operating LCO 3.8.4 DC Sources - Operating LCO 3.8.7 Distribution Systems - Operating EOP Graph 7 Drywell Spray Initiation Limit EOP Graph 4 Heat Capacity Limit EOP 3 Table 6 Secondary Containment Limits RHR NPSH Limits RHR Vortex Limits EAL 01 Hot Matrix

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295001 AK1.02 Importance Rating 3.3 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Power/flow distribution.

Proposed Question: RO Question 1 The plant was operating at 100% power when the A recirculation pump tripped due to a fault.

Plant conditions are now as follows:

• Reactor power is 63%

• FR-4528 & PDR-4528 Total Core Flow / Core Plate Dp Recorder reads:

o Total Core flow: 26 Mlbm/hr o Core Plate D/P: 3.6 psid What action, if any, is required to be taken by the Operator?

A. insert control rods to lower power to approximately 58%

B. reduce recirculation speed to lower core flow to 25 Mlbm/hr C. place the mode switch in shutdown D. no further action is required Proposed Answer: A Explanation: AOP 255.2 For low core flow AND single loop operation conditions (i.e. <27 Mlbm/hr and/or < 7 psid), obtain Core Plate dp from PDR/FR-4528 and use Core Flow vs Core Plate dp graph under Attachment 2 to determine core flow in Mlbm/hr. Using 3.6 psid provided and graph core flow would be 22 Mlbm/hr. IPOI-3 plot on power to flow map using these values of core flow and power places the plant above the MELLA Limit.

AOP-264 Maintain the following administrative limits during single loop operation

• Core power shall be less than or equal to 60%

• Core Flow shall be less than or equal to 53%

• Active loop jet pump flow shall be less than or equal to 32.0 Mlb/hr.

A. Correct AOP 255.2 Rev 48. Follow up action 6.b determine core flow from Attachment 2 (22 Mlb/Hr. Plotting on IPOI-3 Power to flow map this places the unit above MELLA.

Follow up ACTION 8 If inadvertent entry into area area above power to flow map (ie exceeding the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average load line of 100.64%) exit this area by inserting control Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 1 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET rods. If unable to reduce power below the MELLA limit within one hour, manually scram the reactor.

B. Incorrect AOP 264 Rev 16 Follow up action 3, is required for single loop operation restrictions, however the first action would be to reduce core power to below the MELLA limit. 58%

with control rods will establish this.

Based on plate dp / core flow plot plant is below 25 Mlbm/hr C. Incorrect Manual scram is not required unless the unit is not restored to below MELLA within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Incorrect, further action is required.

AOP 264, Rev. 16 Technical Reference(s): (Attach if not previously provided)

AOP 255.2 Rev 48 IPOI 3, Rev. 160 Core Plate DP vs.

Proposed References to be provided to applicants during examination: Core Flow and IPOI 3 P/F map 12.01.01.01 Identify the appropriate procedures that govern the Recirculation System operation, Learning Objective: include the operator responsibilities (As available) during all modes of operation, and any actions required by personnel outside of the control room Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 2 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295003 G2.4.3 Importance Rating 3.7 Partial or Complete Loss of AC Power: Ability to identify post-accident instrumentation.

Proposed Question: RO Question 2 Which of the following post-accident instruments would be unavailable if the RCIC Inverter failed?

A. A Wide Range Level Indicator LI 4539 B. B Wide Range Level Indicator LI 4540 C. A Fuel Zone Level Indicator LI 4565C D. B Fuel Zone Level Indicator LI 4565B Proposed Answer: A Explanation:

A. Correct AOP 302.1 Rev 59 Los of 1D13 probable indication 1C05 A Wide Range LI-4539 fails low 1D13 ckt 03 is RCIC Inverter B. Incorrect: AOP 302.1 Loss of 125 VDC Power 1D23 probable indication 1C05 LI-4540 Wide Range fails low OI-152A1Rev HPCI Electrical Lineup. 1D2302 is HPCI Inverter OI152A1 HPCI Inverter is LT4540 C. Incorrect LI4565C Fuel Zone is 1Y11power D. Incorrect LI4565B Fuel Zone is 1Y21 power Technical Reference(s): AOP 301, Rev. (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 3 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET 5.01.01.02 Given a HPCI system operating mode and various plant conditions, predict how the HPCI Learning Objective: (As available) system will be impacted by the following supported system failures:

a. 125 VDC System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 (6) Design, components, and function of reactivity control mechanisms and instrumentation.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 4 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295004 AA2.03 Importance Rating 2.8 295004 (APE 4) Partial or Total Loss of DC Power / 6: AA2.03 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

Battery voltage. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: RO Question 3 Given the following conditions:

• 1D43 is in service

• 1D12 is in service A Loss of offsite power and 1A3 lockout has occurred. Several hours later the following plant conditions exist:

• 1A3 is damaged and will require several days to repair

• All non-essential DC loads have been shed

• It has been noted that Div 1 125 VDC is 110 volts and lowering In accordance with AOP 301, Loss of Essential Power, what action would be required to maintain 125 VDC Division 1 for long term operation?

Align 1D120 to 1D10 powered from _________________.

A. 1B32, CB 480VAC Essential Motor Control Center B. 1B42, CB 480VAC Essential Motor Control Center C. the FLEX Diesel D. the TSC Diesel Proposed Answer: B Explanation:

A. Incorrect: 1B32 is a Division 1 power source and the stem states that 1A3 is locked out.

The student may choose this because they may incorrectly believe 1D120 should be powered from the respective 480V source it is aligned to. This is not true.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 5 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET B. Correct: 1D120 can be supplied from 1B42 or 1B32. OI302 aligns the 1D120 battery charger to Division 2 480VAC. This is directed by AOP 302.1 for Loss of 1A3 power which was given in the stem. Bus voltage would lower over time as the battery power is being consumed.

C. Incorrect: The FLEX Diesel. Plausible since this is actually possible to be done via flex connections however since 1A4 Division 2 4160 power is available this option is unnecessary and not directed by the procedure. This option is only available if all essential 4160VAC power is lost (station blackout). Those conditions are not present.

1A4 still has power.

D. Incorrect: The TSC Diesel. Plausible since this is actually possible to be done via flex connections however since 1A4 Division 2 4160 power is available this option is unnecessary and not directed by the procedure. This option is only available if all essential 4160VAC power is lost (station blackout). Those conditions are not present.

1A4 still has power.

SD 375 Rev 9 Plant DC Power System Technical Reference(s): (Attach if not previously provided)

AOP 301 Rev 75 Loss of Essential Electrical Power Proposed References to be provided to applicants during examination: N 94.01.05.06 Relate how each step and its performance meets the mitigation Learning Objective: (As available) strategies of AOP 301.1 Station Blackout Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 6 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295005 AA1.01 Importance Rating 3.1 Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: Recirculation system: Plant-Specific.

Proposed Question: RO Question 4 The plant is operating at 50% when the Main Turbine TRIPS. The Operator carrying out the IPOI 5, SCRAM actions observes the following:

• Both Reactor Recirc Pumps are running at 20% speed

• Reactor Water Level is in the Green Band

• Reactor Pressure is 940 psig and stable being maintained with EHC What, if any, action(s) should have happened with regards to Reactor Recirc?

A. ONLY the B Reactor Recirc Pump TRIPS B. Both Reactor Recirc Pump speeds reduce to minimum C. Both Reactor Recirc Pumps trip D. NO additional actions Proposed Answer: C Explanation:

A. Incorrect.

Automatic action failed to occur, the operator should take the action to trip both Recirc pumps. This action is plausible if the student misses the RPT TRIP and assumes that LPCI loop select logic should have caused the only B recirc pump to TRIP. With Reactor water level and pressure normal and no containment conditions provided in the stem, LPCI loop select has no signal to select the loop and trip the pump.

B. Incorrect The recirc are already at minimum speed. Plausible, the operator may incorrectly believe that the recirc pump speed should be lowered to 0 on the controllers. Lowering the setpoint below 20 has no effect on pump speed since it is locked at 20%. At >26%

power with a Turbine Trip RPT breakers will trip (Recirc pumps 1P201A &B both trip).

C. Correct Recirc pumps should have auto tripped via RPT on the Turbine Trip at >26% power.

D. Incorrect.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 7 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Automatic action failed to occur, trip both Recirc pumps. The operator may believe that since plant conditions are now stable no action should be taken however, conduct of operations states if an automatic action fails to occur, the action should be taken.

ARP 1C05B (A-5) Rev 108 RPT Technical Reference(s): (Attach if not previously provided)

System A or B Trip IPOI-5 Rev 62 Appendix 2 Case 2 Scram caused by turbine trip Proposed References to be provided to applicants during examination: N 12.01.01.12 Describe the effect on Learning Objective: RPV water level, that is caused by (As available)

Recirculation Pump operations Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 8 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295006 AK3.01 Importance Rating 3.8 295006 (APE 6) Scram / 1: AK3.01 - Knowledge of the reasons for the following responses as they apply to SCRAM: Reactor water level response. (CFR: 41.5 / 45.6)

Proposed Question: RO Question 5 Which of the following is the expected initial response of reactor water level when a manual reactor scram is inserted from 100% reactor power and which system operation impacts this response?

Reactor water level will INITIALLY ______________.

A. rise due to the delayed response of the feed regulating valves B. rise due to the rise in CRD cooling water flow C. lower due to main turbine bypass valves opening D. lower due to control rod insertion Proposed Answer: D Explanation: IPOI-5 Rev 62 Appendix 2 Automatic Actions Case 1 Reactor Scram NOT initiated by turbine trip or main steam isolation. Step 3 The Condensate and reactor feed pump recirculation valves open and the feedwater control system attempts to maintain level. Reactor water level will initially shrink due void collapse.

A. Incorrect RPV/L will initially lower due to control rod insertion and void collapse. The student may choose this reason if they believe the rapid power reduction could result in rising indicated water level from the loss of 2 phase flow resistance.

B. Incorrect RPV/L will initially lower due to control rod insertion and void collapse. The student may choose this if they believe the rapid increase of CRD flow from the SCRAM Valves opening could cause RPV water level to rise initially. This only becomes true several minutes after the scram when losses from decay heat are overcome.

C. Incorrect RPV/L will initially lower due to control rod insertion and void collapse.

Turbine BPV opening will lower RPV pressure and will cause level rise on lowering pressure and swell D. Correct Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 9 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET RPV/L will initially lower due to control rod insertion and void collapse. This affect is instantaneous and pronounced often lowering RPV level below the SCRAM setpoint and coming close to the LO-LO level setpoint.

IPOI-5 Rev 62 Appendix 2 Technical Reference(s): (Attach if not previously provided)

Automatic Actions (page 13 of 15)

Proposed References to be provided to applicants during examination: N 93.22.01.02 Differentiate beween the plant response to a scram from a Learning Objective: (As available)

Turbine Trip, MSIV isolation, and any non-turbine related scram signal Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 10 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295016 AK2.02 Importance Rating 4.0 295016 (APE 16) Control Room Abandonment / 7: AK2.02 - Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Local control stations: Plant-Specific. (CFR: 41.7 / 45.8)

Proposed Question: RO Question 6 After control room abandonment an operator stationed at 1C208 is operating RCIC.

What indication or control is available to the Operator at 1C208 associated with RCIC operation?

A. MO-2404 RCIC TURBINE STEAM SUPPLY VALVE Control Switch B. RCIC System Flow Indication C. MO-2512 RCIC INJECT VALVE Control Switch D. RCIC Turbine Speed Control Proposed Answer: D Explanation: At 1C208 when operating RCIC, the only indications of RCIC system operation are from the Turbine speed controller.

A. Incorrect: RCIC Turbine Steam Supply Valve position is not available at 1C208. This indication is available at 1C390 Remote Shutdown panel B. Plausible as this indication is available remotely however not at 1C208 when this procedure is implemented.

B. Incorrect: RCIC System flowrate is not available when control room abandonment procedures are implemented. Plausible in that the student may incorrectly assume that the controller at 1C208 controls flow and not speed of the turbine.

C. Incorrect: RCIC Inject Valve position is not available at 1C208. This indication is available only in the control room or locally. Plausible as this indication is available when operating RCIC from the control room and the student may incorrectly assume it is available at the local station as well.

D. Correct: At 1C208 when operating RCIC, the only indications of RCIC system operation are from the Turbine speed controller.

SD-925 Rev 8 Remote Shutdown Panel Technical Reference(s): (Attach if not previously provided)

SD-150 Rev 9 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 11 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 3.08.01.04 Describe the RCIC System interlocks, including purpose, Learning Objective: (As available) setpoints, logic, and when/how they are bypassed Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.45 8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 12 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295018 AK1.01 Importance Rating 3.5 295018 (APE 18) Partial or Complete Loss of CCW / 8: AK1.01 - Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on component/system operations. (CFR: 41.8 to 41.10)

Proposed Question: RO Question 7 With the plant operating at 100% reactor power what action, if any, would an Operator take for a complete loss of RBCCW to the drywell that lasts for 15 minutes?

A. insert a manual reactor scram B. raise CRD mini-purge flow rate to 5 GPM C. lower recirculation pump speed to minimum D. no Operator action is required for this time period Proposed Answer: A Explanation: ARP 1C-06B (D-3) Operator Action Step 3.7 If continued cooling to the Reactor Recirc Pumps is not possible, manually scram the reactor per IPOI-5 (Reactor Scram) and secure the Reactor Recirc Pumps within 10 minutes or if seal cavity temperatures reach 250°F TR-4600 on 1C21.

OI-264 P&L 7, Do not Operate recirc pumps more than 10 minutes without RBCCW cooling water to the pump seals.

A. Correct ARP 1C06B (D-3) RBCCW Pump Discharge Low Pressure If continued cooling of the Reactor Recirc Pumps is not possible, manually scram the reactor per IPOI-5 Reactor Scram and secure Reactor Recirc Pumps within 10 minutes or if seal cavity temperatures reach 250°F TR4600 on 1C21.

B. Incorrect Since RBCCW provides cooling water to the Recirc Pump seal coolers, the student may think raising minipurge flow will be an action to be taken to provide cooling in the absence of RBCCW since it does provide a small amount of cooling to the seals.

C. Incorrect Loss of RBCCW cooling to requires scram and securing Recirc pumps within 10 minutes. Although this action would reduce some of the heat addition to the pump seals, this is inadequate to remove the heat developed. The pumps must be stopped.

D. Incorrect Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 13 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Loss of RBCCW cooling to requires scram and securing Recirc pumps within 10 minutes. It is plausible to consider continued operation of the Recirc pumps if the student does not know that RBCCW is the cooling water supply to the Recirc pumps.

ARP 1C06B (D-3)

Technical Reference(s): (Attach if not previously provided)

OI-264 Rev 143 P&L 7 (page 4 of 65)

Proposed References to be provided to applicants during examination: N 29.01.01.02 Identify the appropriate procedures that govern the RBCCW system operation, include operator Learning Objective: (As available) responsibilities during all modes of operation, and any actions required by personnel outside the control room Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8-10 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 14 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295019 G2.1.7 Importance Rating 4.4 (APE 19) Partial or Complete Loss of Instrument Air: G2.1.7 - Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

Proposed Question: RO Question 8 The plant is operating at 50% reactor power when a complete loss of plant air occurs.

The following indications are received in the Control Room:

• Rod Drift Annunciator

• Scram Air Header HI/Low Pressure Alarm What Operator action should be taken to address this condition?

A. Scram the reactor immediately B. Continue to monitor for any additional drifting control rod C. Place the Emergency Rod In switch to EMERGENCY IN momentarily D. Run an ACUMEN Report to determine if thermal limits remain acceptable Proposed Answer: A Explanation: AOP-518 Rev 42 immediate actions require manually scramming the reactor if instrument air header pressure is rapidly decreasing or cannot be restored or any rod has drifted 1C-05B (F-1) Scram Air Header HI/LO Pressure, Operator Action, If any control rod starts drifting into the core, manually scram the reactpr per IPOI-5 (Reactor Scram)

A. Correct AOP-518 Rev 42 immediate actions require manually scramming the reactor if instrument air header pressure is rapidly decreasing or cannot be restored or any rod has drifted B. Incorrect, The condition in the stem requires a manual scram. Plausible in that monitoring for additional rod drift conditions is a follow up action for a rod drift annunciator with no scram air header pressure low alarm present.

C. Incorrect, Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 15 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET The condition requires a manual scram. This is plausible since it is an action that is taken to abort the sequence timer and is provided in the ARP response for ROD DRIFT.

In addition Scram Air header annunciator is a combined annunciator for both air header pressure high or low. The student may assume it is a high pressure condition and take the actions for the ROD Drifting only. In addition, the scram air header pressure alarm does not necessarily indicate that control rods are moving.

D. Incorrect, The condition in the stem requires a manual scram. This is plausible in that if the operator assumes that this is a single rod motion (single rod scram or drift), then an ACUMEN case would be run to verify all rod positions and thermal limits Per AOP 255.1 AOP-518 Rev 42 1C-05B (F-1) Rev 108 (page 103 Technical Reference(s): (Attach if not previously provided) of 113)

Proposed References to be provided to applicants during examination: N 36.00.00.05 Evaluate plant conditions and control room indication to determine if the instrument and service air system is operating as Learning Objective: (As available) expected, and identify any actions that may be necessary to place the instrument and service air system in the correct lineup Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43/45 5/12-13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 16 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295021 AA2.01 Importance Rating 3.5 295021 (APE 21) Loss of Shutdown Cooling / 4: AA2.01 - Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor water heatup/cooldown rate. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: RO Question 9 The plant has been shutdown for 18 days and is in MODE 5 with Shutdown cooling in service on the A side with the following conditions present:

• Reactor Coolant Temperature is 160 °F

• Cooldown rate is 5°F per hour

• Fuel pool gates are removed

• B Reactor Recirc Pump is in service Then the following occurs:

• Essential 4160V Bus 1A4 Lockout (1) What is the status of Shutdown Cooling AND (2) What is the effect on Reactor Vessel Water Heatup/Cooldown Rate?

A. (1) SDC remains in service (2) Cooldown Rate will be greater than 5°F per hour B. (1) SDC remains in service (2) Cooldown Rate will be less than 5°F per hour C. (1) SDC has been lost (2) Heatup Rate will be approximately 20°F per hour D. (1) SDC has been lost (2) Heatup Rate will be approximately 3.7°F per hour Proposed Answer: D Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 17 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Explanation: Loss of 1A4 results in loss of RPS B. This initiates a Group 4 isolation that isolates SDC by closing MO 1909 outboard SDC isolation. The RHR pumps that were running will TRIP as a result of the loss of suction. This results in a loss of SDC. The approximate heatup rate for the given condition is determined by utilizing the AOP for Loss of Shutdown cooling and selecting the correct heatup rate graph. The conditions given specify fuel pool gates removed implying a flood up condition. The student should use the flood up graphs Appendix 1 to determine the approximate heatup rate.

A. Incorrect The loss of 1A4 results in a Group 4 B logic isolation. This causes a loss of SDC due to MO-1909 SDC suction and MO 1905 LPCI inject valve interrupting SDC flow. The closure of MO 1909 results in A side RHR Pumps to Trip on no suction path.

Plausible if the student believes a lockout on 1A4 (div 2) would not affect 1A3 powered equipment. In addition if the student assumes the B Reactor Recirc Pump TRIPS when 1A4 is de-energized, and that pump heat addition is lost, the cooldown rate could rise.

SDC is lost so the cooldown rate will not remain the same. There would be a heatup rate to calculate.

B. Incorrect The loss of 1A4 results in a Group 4 B logic isolation. This causes a loss of SDC due to MO-1909 SDC suction and MO 1905 LPCI inject valve interrupting SDC flow. The closure of MO 1909 results in A side RHR Pumps to Trip on no suction path.

Plausible if the student believes SDC remains in service however a lockout on 1A4 (div

2) would result in the loss of RHRSW flow to the RHR Heat Exchangers resulting in a loss of cooldown rate until that flow is restored.

C. Incorrect The correct heatup rate from the vessel flooded graph is 3.7F/hr. 20F/hr is plausible if the student does not recognize the vessel flooded condition and utilizes the wrong appendix of the procedure.

D. Correct The loss of 1A4 results in a Group 4 B logic isolation. This causes a loss of SDC due to MO-1909 SDC suction and MO 1905 LPCI inject valve interrupting SDC flow. The closure of MO 1909 results in A side RHR Pumps to Trip on no suction path.

AOP-149 Rev 147 Loss of Shutdown Cooling AOP 301 Rev 74, Loss of Essential Electrical Power Technical Reference(s): (Attach if not previously provided)

AOP-358 Rev 32 (page 2 Auro Actions)

ARP 1C05B (D-4) Rev 108 (page 80 of 113)

AOP 149 App. 1 Proposed References to be provided to applicants during examination:

AOP 149 App. 2 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 18 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET 5.01.01.01 Evaluate plant conditions and control room indications and Learning Objective: (As available) determine the actions directed by AOP 149 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 19 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295023 AA1.06 Importance Rating 3.3 295023 (APE 23) Refueling Accidents / 8: AA1.06 - Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: Neutron monitoring. (CFR: 41.7 / 45.6)

Proposed Question: RO Question 10 The MODE switch is in REFUEL with the vessel head removed.

The following SRM indications are noted:

Period CPS SRM A 1E3 SRM B 1E3 SRM C +10 (steady) 2E4 SRM D +100 (steady) 5E5 What signals are automatically generated by the SRMs, if any?

A. Rod Block only B. Half Scram and Rod Block C. Full Scram and Rod Block D. No automatic signals are generated Proposed Answer: A A

SRM channel upscale neutron flux level trip was only used during initial fuel loading and low power physics testing to provide reactor protection until overlap between the SRM and IRM s was demonstrated. Any single SRM, IRM, or APRM upscale or IRM or APRM Inop trip would have produced a trip signal in both channels A3 and B3 of RPS A. Correct: 1E5 cps is the SRM Upscale setpoint. Only a Rod Block would result from this condition B. Incorrect: 1E5 cps is the SRM Upscale setpoint.

Only a Rod Block. With RPS shorting links installed the RPS Scram function is disabled.

The coincidence circuitry is not divisionalized however it is bypassed by shorting links.

This choice is plausible if the student thinks the circuitry is divisionalized.

C. Incorrect: 1E5 cps is the SRM Upscale setpoint.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 20 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Only a Rod Block. With RPS shorting links installed the RPS Scram function is disabled.

Plausible if the student thinks the shorting links are not installed.

D. Incorrect: 1E5 cps is the SRM Upscale setpoint.

Shorting Links installed preventing RPS trip. Plausible if the student believes the shoring links installed disables both the SCRAM and Rod Block functions. 5E5 CPS is also the original setpoint for the Scram function of the instrumentation.

SD358 Rev 9 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 78.02.01.06 Describe the SRM system interlocks (include alarms), including Learning Objective: (As available) purpose, setpoints, logic, and when/how they are bypassed Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 21 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295024 EK3.07 Importance Rating 3.5 295024 High Drywell Pressure / 5: EK3.07 - Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Drywell venting. (CFR: 41.5 / 45.6)

Proposed Question: RO Question 11 Primary Containment Pressure is rising following a large RCS break inside the primary containment. Primary Containment parameters are as follows:

• Drywell Pressure indication is 49 psig and slowly rising

• Torus pressure indication is 45 psig and slowly rising

• Wide Range Torus level indication is 12.5 feet and stable IAW EOP 2, Primary Containment Control, (1) What is the correct vent path?

AND (2) Why?

A. (1) Torus vent through CV 4301 OUTBD TORUS VENT ISOL, CV 4309 INBD TORUS VENT BYPASS ISOL and CV 4300 INBD TORUS VENT ISOL as required (2) The filtering of the SBGT system will reduce the offsite release rate B. (1) Drywell vent through CV 4303 OUTBD DRYWELL VENT ISOL, CV 4310 INBD DW VENT BYPASS ISOL and CV 4302 INBD DRYWELL VENT ISOL as required (2) This is the ONLY vent path available due to level in the TORUS C. (1) Drywell vent through CV 4303 OUTBD DRYWELL VENT ISOL, CV 4310 INBD DW VENT BYPASS ISOL and CV 4302 INBD DRYWELL VENT ISOL as required (2) This vent path will release to the Offgas Stack for an elevated release dilution D. (1) Torus vent through the hardened vent path via CV-4360, TORUS HARDPIPE VENT INBOARD ISOLATION and CV 4361 TORUS HARD PIPE VENT OUTBD ISOLATION as required (2) This path eliminates the potential for duct work or SBGT failure during venting which would significantly increase radioactivity levels in the reactor building.

Proposed Answer: A Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 22 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Explanation:

A. Correct:

Torus venting is normally preferred due to the scrubbing provided by the torus and this path also utilizes SBGT.

B. Torus venting is normally preferred due to the scrubbing provided by the torus and this path also utilizes SBGT Torus level is below 16 ft. The level at which the Drywell is vented C. Torus venting is normally preferred due to the scrubbing provided by the torus and this path also utilizes SBGT SEP 301.3 Torus Vent via Hardpipe vent is performed when the normal Torus vent path is unavailable. In this case normal Torus vent is available.

Torus level is below 16 ft. The level at which the Drywell is vented D. Torus level is below 16 ft. The level at which the Drywell is vented Torus venting is normally preferred due to the scrubbing provided by the torus and this path also utilizes SBGT. In the stem there is no mention of loss of RPS which would prevent alignment of the normal Torus vent path. This 2 vent path remains the preferred vent path and should be utilized. Plausible if the student believes the Hardened Vent path is preferred.

SEP 301.1 Torus Vent via SBGT Rev 13 Technical Reference(s): Sep 301.3 Torus Vent via (Attach if not previously provided)

Hardpipe Vent Rev 15 EOP-2 Rev 18, Step PC/P-10 Proposed References to be provided to applicants during examination: N 95.64.16.01 Contrast the effects of Learning Objective: venting the drywell and venting the (As available) torus Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 23 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295025 EK2.04 Importance Rating 3.9 295025 (EPE 2) High Reactor Pressure / 3: EK2.04 - Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: ARI/RPT/ATWS. (CFR: 41.7 / 45.8)

Proposed Question: RO Question 12 The plant is experiencing ATWS conditions with a GROUP 1 isolation. The following conditions present:

• Reactor Pressure is 1140 psig and rising

• RPV Level is 119.5 inches and lowering Based upon the conditions stated above (1) When will the ARI valves OPEN?

(2) When will the RPT breakers TRIP?

A. (1) After a 9 second time delay (2) IMMEDIATELY B. (1) IMMEDIATELY (2) IMMEDIATELY C. (1) IMMEDIATELY (2) After a 9 second time delay D. (1) After a 9 second time delay (2) After a 9 second time delay Proposed Answer: B Explanation:

On a High Reactor Pressure condition, the ARI Valves OPEN and RPT Breakers TRIP immediately. SD 358 Reactor Protection System Figure 10, Rev 9 A. Incorrect, ARI Valves OPEN immediately on a High Pressure Condition or Low Level condition.

Plausible if the student confuses the Low Level time delay RPT TRIP with the Valve Opening signal from ARI B. Correct ,

On a RPV pressure signal of 1140 psig both the RPT and ARI actuate immediately Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 24 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Incorrect, At 1140 psig both ARI and RPT actuate immediately. Plausible if student believes the Low-Low Level 9 second time delay trip for RPT on high reactor pressure.is applicable.

D. Incorrect At a high RPV pressure at 1140 psig, both RPT and ARI actuate immediately, neither have a 9 second time delay on high reactor pressure. Plausible if student believes that 9 second time delay is applicable to the 1140 psig setpoint. Low-Low Level will immediately actuate the ARI valves. Low-Low Level enforces a 9 second time delay fpr recirc pump trip only to allow LPCI Loop Select time to properly select the intact recirc loop for injection.

SD-358 RPS System Desctription Technical Reference(s): (Attach if not previously provided)

Rev 9 Proposed References to be provided to applicants during examination: N 52.01.01.02 Given an EHC system operating mode and various plant conditions, predict how the EHC Learning Objective: (As available) system will be impacted by failures in the following support systems:

a. Logic Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 25 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295026 EK1.02 Importance Rating 3.5 295026 (EPE 3) Suppression Pool High Water Temperature / 5: EK1.02 - Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Steam condensation. (CFR: 41.8 to 41.10)

Proposed Question: RO Question 13 The plant is operating at 100% reactor power.

• Suppression pool water temperature is 90°F and rising The first operational limit is reached based upon the Suppression Pool having capability to provide A. adequate Net Positive Suction Head to the ECCS pumps B. adequate heat capacity in the Torus to prevent exceeding HCTL during an ATWS C. complete steam condensation following a Loss of Coolant Accident D. complete steam condensation following inadvertent Safety Relief Valve actuation Proposed Answer: C Explanation: The first operational limit is 95°F which is the assumed initial temperature of the LOCA analysis. The suppression pool is designed to absorb the decay heat and sensible energy released during a reactor blowdown from safety/relief valve discharges or from Design Basis Accidents (DBAs) The suppression must quench all the steam released through the downcomer lines during a Loss of Coolant Accident (LOCA) TS Bases 3.6.2.1 The technical concerns that lead to the development of the suppression pool average temperature limits are as follows: a) complete steam condensation, the original limit for the end of a LOCA blowdown was 170 °F based on the Bodega Bay and Humboldt Bay Tests. b) Primary containment peak pressure and temp design pressure is 56 psig and design temp is 281°F A. Incorrect NPSH is not adversely affected at 90°F Torus water temperature B. Incorrect, heat capacity limit is not adversely affected by 90°F Torus water temperature.

Typically 160°F is where capacity limit is approached or exceeded C. Correct The suppression pool is designed to absorb the decay heat and sensible energy released during a reactor blowdown from safety/relief valve discharges or from Design Basis Accidents (DBAs) The suppression must quench all the steam released through the downcomer lines during a Loss of Coolant Accident (LOCA) TS Bases 3.6.2.1 The technical concerns that lead to the development of the suppression pool average Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 26 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET temperature limits are as follows: a) complete steam condensation, the original limit for the end of a LOCA blowdown was 170 °F based on the Bodega Bay and Humboldt Bay Tests. b) Primary containment peak pressure and temp design pressure is 56 psig and design temp is 281°F D. Incorrect: This is not the basis for the stuck open SRV. The basis for the 95°F torus Water Temperature is the assumed temperature at the onset of the LOCA and the ability of the suppression pool to accept the heat addition from a design basis accident.

Plausible if the student incorrectly attributes the Torus Water Temperature limit to a stuck open SRV.

Tech Bases 3.6.2.1 Amendment Technical Reference(s): (Attach if not previously provided) 223 Proposed References to be provided to applicants during examination: N 95.00.00.15 Explain the Bases of each Learning Objective: (As available) of the EOP Curves and Limits Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8, 10 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 27 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295028 G2.4.46 Importance Rating 4.2 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5: Generic K/A 2.4.46 -

Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3

/ 45.12)

Proposed Question: RO Question 14 The plant is operating at 100% reactor power.

• All well water cooling to the drywell is lost

• Drywell pressure is 1.1 psig and rising slowly Which one of the following would alarm FIRST in the control room to alert the operators?

A. 1C35A C-4 DW/TORUS 1C-219A GAS/PART/IODINE HI RAD OR DNSCL B. 1C05B A-1 DW HIGH PRESSURE TRIP C. 1C05B A-7 PCIS CHANNEL STEAM TUNNEL HI TEMP D. 1C07A C-11 DRYWELL COOLING PANEL 1C-25 TROUBLE Proposed Answer: D Explanation:

A. Incorrect: A loss of well water cooling to the drywell would not result in this alarm. This alarm is indicative of a reactor coolant system leak in the drywell. Plausible if the student interprets the rising drywell pressure and temperature as a leak.

B. Incorrect: This alarm would come in after Drywell Pressure exceeds 2.0 psig in the drywell. Plausible if the student does not understand that there are several other alarms that come in prior to exceeding 2.0 psig in the drywell such as, 1C05B B-1 CONTAINMENT HI/LO PRESS which is received when DW pressure exceeds 1.5 psig.

C. Incorrect: Well water does not provide cooling to the Steam Tunnel HVAC units.

Plausible if the student thought well water supplied cooling to these units.

D. Correct: Well Water is the only cooling water to the drywell cooling HVAC units thus if flow is lost, temperatures in the area will rise and individual cooler high inlet air temperatures will annunciate.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 28 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET ARP 1C25A/B A-4 DRYWELL COOLING LOOP A/B OVERTEMP AOP-408 rev 34 Technical Reference(s): (Attach if not previously provided)

SD 760 PRIMARY CONTAINMENT VENTILATION SYSTEM Proposed References to be provided to applicants during examination: N 26.01.01.13 Given a well water system operating mode and various conditions, predict how each supported system will be impacted by Learning Objective: (As available) the following well water system failures

a. One or more Well Water Pump(s) trip Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/5,12 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 29 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295030 EA2.01 Importance Rating 4.1 295030 (EPE 7) Low Suppression Pool Water Level / 5: EA2.01 - Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL:

Suppression pool level. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: RO Question 15 Which of the following describes the level and reason for performing an Emergency Depressurization on Torus Water Level low?

Torus water level cannot be maintained above ________.

A. 5.8 Feet, to ensure steam discharged from the drywell to the Torus following a primary system break will be adequately condensed B. 5.8 Feet, to ensure opening an SRV will not result in exceeding the code allowable stresses in the SRV Tailpipe C. 7.1 Feet, to ensure steam discharged from the drywell to the Torus following a primary system break will be adequately condensed D. 7.1 Feet, to ensure opening an SRV will not result in exceeding the code allowable stresses in the SRV Tailpipe Proposed Answer: C Explanation: Torus level of 7.1 Ft corresponds to the bottom of the drywell to torus downcomers. Torus level below 7.1 ft could result in loss of the pressure suppression function of the primary containment A. Incorrect: 5.8 FT Corresponds to the HPCI turbine exhaust elevation. Operation of HPCI with its exhaust unsubmerged will tend to directly pressurize the torus. Plausible if the student mistakes the HPCI Discharge level limit with the Downcomer coverage limit.

The reasons are similar in that both conditions can challenge overpressurizing the containment bypassing the pressure suppression function.

B. Incorrect: 5.8 FT Corresponds to the HPCI turbine exhaust elevation. Operation of HPCI with its exhaust unsubmerged will tend to directly pressurize the torus. Plausible if the student confuses this with the High Torus level limit of 13.8 Feet which also equates to potential containment damage.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 30 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Correct: A Torus level of 7.1 Ft corresponds to the bottom of the drywell to torus downcomers. Torus level below 7.1 ft could result in loss of the pressure suppression function of the primary containment.

D. Incorrect: A Torus level of 7.1 Ft corresponds to the bottom of the drywell to torus downcomers. Torus level below 7.1 ft could result in loss of the pressure suppression function of the primary containment. . Plausible if the student confuses this with the High Torus level limit of 13.8 Feet which also equates to potential containment damage.

Technical Reference(s): EOP-2 Rev 16 Bases page 10 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 95.00.00.15 Explain the Bases of each Learning Objective: (As available) of the EOP Curves and Limits Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43/45 5/13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 31 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295031 EA1.05 Importance Rating 4.3 295031 (EPE 8) Reactor Low Water Level / 2: EA1.05 - Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL: Reactor core isolation system.

(CFR: 41.7 / 45.6)

Proposed Question: RO Question 16 A plant transient has occurred which required the crew to insert a manual reactor scram.

• Reactor water level is 110 inches and lowering

• Drywell pressure is 1.5 psig and rising

• 1C05A (C-1), Reactor LO Level Trip is alarming

• 1C05A (B-1), Reactor LO-LO Level Trip is alarming

• 1C05B (B-1), Primary Containment HI/LO Pressure is alarming The Balance of Plant (BOP) Operator observes the following at 1C03:

• Reactor Core Isolation Cooling (RCIC) is in a standby readiness condition (1) Is this expected for the given plant conditions?

(2) What action is required?

A. (1) Yes, RCIC should receive an auto start signal when 1C05A (A-1), Reactor LO-LO-LO Level Trip, annunciator is received (2) No further action is required B. (1) No, RCIC should have received an auto start signal when 1C05A (C-1), Reactor LO Level Trip, annunciator was received (2) The BOP should manually start RCIC C. (1) No, RCIC should have received an auto start signal when 1C05A (B-1), Reactor LO-LO Level Trip, annunciator was received (2) The BOP should manually start RCIC D. (1) Yes, RCIC should receive an auto start signal when1C05B (A-1), Primary Containment HI Pressure Trip, annunciator is received (2) No further action is required Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 32 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed Answer: C Explanation: RCIC should have automatically started when reactor water level lowered to the Reactor Lo-Lo Level Trip of 119.5. EOP-1 RC-4 Initiate any of the following which should have initiated but did not: (Isolations, ECCS Initiations, SBDG initiations)

A. Incorrect: RCIC starts at RPV LO-LO level. Plausible if the student believes RCIC starts at RPV LO-LO-LO level B. Incorrect: RCIC starts at RPV LO-LO level. Plausible if the student believes RCIC starts at RPV LO level. This is the Reactor SCRAM signal. Plausible if the student believes RCIC starts at this level.

C. Correct: RCIC should have automatically started when reactor water level lowered to Reactor LO-LO Level Trip setpoint of 119.5 inches. The Operator is required to manually start RCIC.

D. Incorrect: The Primary Containment High Pressure signal of 2 psig is a HPCI automatic start. RCIC does not have a Primary Containment High Pressure start. HPCI does also have the LO-LO Reactor Level 119.5 auto start signal. Plausible if the student believes that RCIC starts on Hi Drywell Pressure of 2 psig as does HPCI. The systems are similar with the exception of starting on drywell pressure.

1C05A (B-1) Rev 90 Technical Reference(s): (Attach if not previously provided)

Reactor Water LO-LO Level Trip 1C05B (A-1) Rev. 108 Primary Containment Hi Pressure Trip Proposed References to be provided to applicants during examination: N 3.08.01.04 Describe the RCIC System interlocks, including purpose, Learning Objective: (As available) setpoints, logic, and when/how they are bypassed Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 33 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295037 EK3.02 Importance Rating 4.3 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1: EK3.02 - Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: SBLC injection. (CFR: 41.5 / 45.6)

Proposed Question: RO Question 17 The plant has experienced an ATWS. The following conditions are present:

• Reactor Power is 25%

• Reactor Water Level is 158 inches and stable

• Reactor Pressure is 1000 psig and stable

• ARI was unsuccessful shutting down the reactor

• Torus Water Temperature is 97°F and rising

• MSIVs are closed The RO reports The ATWS QRC is complete with the exception of Boron Injection (1) Should Boron be injected?

AND (2) Why?

A. No, Boron injection should wait until the Boron Initiation Injection Temperature is exceeded to prevent excessive cleanup efforts following the event.

B. No, Boron injection should await intentionally lowering RPV water level to promote homogenous mixing of the Boron solution minimizing thermal hydraulic instabilities.

C. Yes, Boron injection should occur to prevent uneven distribution of boron in the core resulting in inadequate shutdown margin and localized criticality.

D. Yes, Boron injection should occur prior to exceeding the Boron Initiation Injection Temperature to prevent exceeding the Heat Capacity Limit of the Torus.

Proposed Answer: D Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 34 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Explanation:

A. Incorrect, reactor is at 25% power and Torus temperature is rising.

B. Incorrect: Maintaining RPV water level low in the expanded band may reduce boron mixing efficiency and delay reactor shutdown/ (ATWS Bases Rev 19, page 41)

C. Incorrect, uneven distribution of boron in the core is not a concern at this level. At 25%

and greater than 87 in RPV power level control will be entered and lowered intentionally to choke reactor power.

D. Correct: The boron injection requirement is established in the Steps Q-6 and Q-7 (before torus water temperature reaches the BIIT) If boron injection is required, heat is being added to the containment and emergency depressurization may be required. The second condition, found in Step / Q-6, is based on inserting boron before exceeding the Heat Capacity Limit due to high torus water temperature. (page 91 of 96).

A SCRAM failure coupled with an MSIV isolation, however results in rapid heatup of the torus due to steam discharged from the RPV via SRVs. The challenge to the primary containment thus becomes the limiting factor which defines the requirement for boron injection.

Technical Reference(s): ATWS bases Rev 19 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 95.00.00.15 Explain the Bases of each Learning Objective: (As available) of the EOP Curves and Limits Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 35 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295038 EK2.07 Importance Rating 3.5 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9: EK2.07 - Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Control room ventilation. (CFR: 41.7 / 45.8)

Proposed Question: RO Question 18 The Control Building Standby Filter Units have actuated on a valid signal. The BOP investigates and finds that A and B Battery room exhaust fans are operating.

What action(s) should the Operator take?

A. Trip only the A Battery Room Exhaust fan B. Trip only the B Battery Room Exhaust fan C. Trip all running Battery Room Exhaust fans D. Start all non-running Battery Room Exhaust fans Proposed Answer: A Explanation:

A. Correct: OI-730 after an isolation only 1V-EF-30B or C shall be running. If flow from the battery rooms is not limited to 100 cfm following an isolation, the control room SFU may not be able to maintain the control room at a positive pressure. Therefore to maintain positive pressure during a Control Building Isolation, only one Battery Exhaust Fan 1V-EF-30B or 1V-EF-30C shall be running 1V-EF-30A capacity is 700 CFM 1V-EF-30B and C capacity is 100 CFM In a normal configuration 1V-EF-30A is normally running along with the B or C fan.

B. Incorrect C. Incorrect D. Incorrect Proposed References to be provided to applicants during examination: N Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 36 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET 65.01.01.01 Relate the precautions and limitation, operating cautions, or Learning Objective: procedural notes of OI 730 to any (As available) component or Control Building HVAC system operating status STP 3.0.0-02 Rev 59 (page 22 or Question Source: Bank # 31)Control Room Panel Checks SD 730 Rev 12 Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.45 8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 37 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 600000 AK1.02 Importance Rating 2.9 Plant Fire On Site - Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Fighting.

Proposed Question: RO Question 19 An Operator observes the following alarms.

• There are NO additional annunciators at 1C040 In accordance with AOP 913, Fire, the Fire Brigade ___(1)___ required to muster and the pre-action system ____(2)____ activated?

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 38 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET (1) (2)

A. is has B. is has not C. is not has D. is not has not Proposed Answer: B Explanation:

A. Incorrect: From the information provided in the STEM of the question, No additional annunciators are observed to be in alarm at 1C040 or on the RAN 16, indicates that the preaction system has not actuated. Activation would have been followed by the start of the electric or diesel fire pump.

B. Correct: In accordance with AOP 913, FIRE, annunciator A DIESEL GEN. RM. DET.

ZONE 21 (EAST) requires immediate fire brigade activation. From the information provided in the STEM of the question, No additional annunciators are observed to be in alarm at 1C040 or on the RAN 16, indicates that the preaction system has not actuated. Activation would have been followed by the start of the electric or diesel fire pump.

C. Incorrect: This would be true if an Operator is required to verify a presence of a fire. The building in-plant Operator would respond to investigate to determine if a fire is present and to size up the fire.

D. Incorrect: This would be true if an Operator is required to verify a presence of a fire. The building in-plant Operator would respond to investigate to determine if a fire is present and to size up the fire.

Technical Reference(s): AOP 913 Rev 83 (Attach if not previously provided)

ARP 1C40, Rev. 77 Proposed References to be provided to applicants during examination: N Learning Objective: 94.25.01.03 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: PDA 17-1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 39 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 700000 G2.4.31 Importance Rating 4.2 Generator Voltage and Electric Grid Disturbances - Knowledge of annunciator alarms, indications, or response procedures.

Proposed Question: RO Question 20 The plant is operating at 100% reactor power with voltage regulation in AUTO. A grid transient occurs. Prior to the transient, The GENERATOR REGULATOR VOLTS were nulled.

After the transient the following indications are observed:

To null the meter without changing the main generator terminal voltage, the Operator will place the Generator ____(1)____ Voltage Adjust Control Switch in the ___(2)___ direction?

A. (1) Manual (2) LOWER Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 40 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET B. (1) Manual (2) RAISE C. (1) Automatic (2) LOWER D. (1) Automatic (2) RAISE Proposed Answer: A Explanation: At 100% Auto Voltage Regulator is normally in service. The Manual Regulator should be periodically adjusted so that the Generator Regulator Volts meter indicates zero Typically the adjustment is towards the direction the meter is deflected from zero. In this case towards lower A. Correct: Section 4.2 Periodic Checks step 4 Voltage regulation should remain in automatic. The Manual Regulator should be periodically adjusted so that the Generator Regulator Volts meter indicates zero.

B. Incorrect: Raising the MANUAL regulator setting would exacerbate the condition and make the mismatch greater. Plausible if the student does not understand the relationship between the MANUAL Regulator and AUTOMATIC Regulator when configured according to the stem.

C. Incorrect: Adjusting the AUTOMATIC Voltage regulator will directly affect Terminal Voltage of the Generator. The action was to null the two instruments therefore this should not be achieved by actually changing VARS and Voltage of the machine that is in service. In section 4.2 of OI 698 periodic checks it details the method of nulling the two instruments.

D. Incorrect: Adjusting the AUTOMATIC Voltage regulator will directly affect Terminal Voltage of the Generator. The action was to null the two instruments therefore this should not be achieved by actually changing VARS and Voltage of the machine that is in service. In section 4.2 of OI 698 periodic checks it details the method of nulling the two instruments.

Technical Reference(s): OI 698, Rev. 103 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 57.00.00.02 Evaluate plant conditions and control room indications to Learning Objective: determine if the main generator (As available) system is operating as expected, and identify any actions that may be Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 41 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET necessary to place the main generator system in the correct lineup Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 42 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295008 AA2.05 Importance Rating 2.9 High Reactor Water Level - Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL: Swell. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: RO Question 21 The plant is operating at 100% reactor power.

• B Reactor Recirc MG set spuriously runs back to 45% speed Which one of the following describes the RPV level response to this malfunction?

Indicated RPV Water Level _________.

A. lowers and the Reactor scrams on low level B. lowers and the Reactor does NOT scram on low level C. rises and the Main Turbine TRIPs on high level D. rises and the Main Turbine does NOT TRIP on high level Proposed Answer: D Explanation: 45% runback will cause power to lower and less steam from the reactor resulting in Downcomer level to SWELL resulting in an indicated Higher Reactor Water level. The feedwater control system will sense the rising water level and automatically adjust to close the feedwater reg valves and control reactor level.

A. Incorrect, on runback level will rise and may come close to 211 reactor feed pump and turbine trip setpoints. RPV level will not lower to the low scram setpoint. Plausible if the student does not understand the expected water level trend for a loss of reactor Recirc flow and how the plant responds.

B. Incorrect level dose not lower it will rise. Plausible if the student does not understand the expected water level trend for a loss of reactor Recirc flow and how the plant responds.

C. Incorrect level will rise and feedwater control system will sense the high level and error signal will close the feedwater reg valves. Plausible if the student does not understand plant response to a runback signal and the response time of the FWLCS to the lowering Reactor power level.

D. Correct Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 43 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET SD-644 Rev 16 page 42 Figure 5 Technical Reference(s): (Attach if not previously provided)

Feedwater Control System Proposed References to be provided to applicants during examination: N 45.02.01.04 Given a Feed and Condesate System operating mode and various plant conditions, predict Learning Objective: how each supported system will be (As available) impacted by railures in the feed and condensate system:

F. Recirculation System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 44 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295012 AA1.01 Importance Rating 3.5 High Drywell Temperature - Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell ventilation system.

Proposed Question: RO Question 22 The plant was operating at 100% power with the Drywell Cooling system in its normal full power configuration when a small steam leak in the Drywell occurred.

The following is the sequence of events:

• EOP 2 is entered when Drywell temperature exceeds 150°F

• Drywell pressure exceeds 2 psig

• Drywell spray was initiated to maintain Drywell temperature < 280°F Assuming no other operator action has been taken, which of the following is correct regarding the automatic response of the drywell cooling fans to the above sequence?

All running fans __________.

A. tripped when Drywell pressure exceeded 2 psig and no other automatic action occurred B. shifted to slow speed when Drywell pressure exceeded 2 psig and no other automatic action occurred C. shifted to slow speed when Drywell pressure exceeded 2 psig and all fans tripped when Drywell spray was initiated D. remained at their original speed when Drywell pressure exceeded 2 psig and all fans tripped when Drywell spray was initiated Proposed Answer: C Explanation:

A. Incorrect: Fans shift to slow speed when drywell pressure exceeds 2 psig. Additionally the fans tripped when drywell spray was initiated sprays are initiated.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 45 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET B. Incorrect: Drywell fans all tripped once drywell spray was initiated.

C. Correct: All fans shift to slow speed when drywell pressure exceeds 2 psig. All fans trip when drywell sprays are initiated.

D. Incorrect: All drywell fans are normally running in fast speed. Fans shift to slow speed when drywell pressure exceeds 2 psig. Additionally all drywell fans trip when drywell SD760 Rev 8 page 10, 14 (Note 2)

Technical Reference(s): Figure 2 page 18 logic trips fan (Attach if not previously provided) with Drywell Spray valves open Proposed References to be provided to applicants during examination: N 2.01.01.07 Given an RHR system operating mode and various plant conditions, predict how each Learning Objective: supported system will be impacted by (As available) the following RHR system operations/failures:

c. Containment Spray Initiation Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 46 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295014 AK3.02 Importance Rating 3.7 Inadvertent Reactivity Addition - Knowledge of the reasons for the following responses as they apply to INADVERTENT REACTIVITY ADDITION: Control rod blocks.

Proposed Question: RO Question 23 Which one of the following describes the reason for the Rod Block Monitor System to enforce a Control Rod Block?

The Rod Block Monitor will enforce a Control Rod Block to prevent ________________.

A. inadvertent criticality during refueling B. creation of a high worth control rod notch C. inserting a control rod out of its step sequence D. violating the MCPR limit for a single control rod withdrawal error Proposed Answer: D Explanation:

A. Incorrect: This is a function of the Rod Out Block function of the Reactor Manual Control System. Since this function generates a Control Rod Block, the student may associate it with the Rod Block Monitor system.

B. Incorrect: This is minimized by development of the control rod sequence design and the Rod Worth Minimizer system. Plausible if the student believes the RBM is designed to prevent a core design error or out of sequence error.

C. Incorrect: This is a function of the Rod Worth Minimizer to enforce the control rod sequence. Plausible if the student confuses the purpose of the RBM system with the RWM system.

D. Correct: RBM uses LPRM outputs to determine the thermal power production in a localized area around any selected rod. This information is utilized to block control rod withdraw that could result in a violation of Safety Limit Minimum Critical Power Ratio (SLMCPR) from a single rod withdraw.

SD878.5 RBM Rev 10(Purpose Technical Reference(s): (Attach if not previously provided) page 4)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 47 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 82.01.01.04 State when the Rod Block monitoring system is required to be Learning Objective: operable by technical specifications (As available) and describe the bases of the Rod Block monitoring system LCOs Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 48 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295020 AK2.11 Importance Rating 3.2 Inadvertent Containment Isolation - Knowledge of the interrelations between INADVERTENT CONTAINMENT ISOLATION and the following: Standby gas treatment system/FRVS: Plant-Specific.

Proposed Question: RO Question 24 The Fuel Pool Exhaust RIS-4131B Rad Monitor has failed with an INOPERABLE signal. The following is observed from panel 1C24:

• Both SBGT Trains are in a standby readiness lineup.

Is this an expected response for this failure?

A. No, the A Train of SBGT should have started B. No, the B Train of SBGT should have started C. No, both trains of SBGT should have started D. Yes, an INOPERABLE signal does not start SBGT Proposed Answer: B Explanation:

A. Incorrect: An INOPERABLE signal created by the Radiation Monitor will cause start signal to the B Standby Gas Treatment System (SBGT). Plausible if the student believes any GP.3 Rad monitor would cause both trains to start and isolate this could be true.

B. Correct: An INOPERABLE signal created by the Radiation Monitor will cause start signal to the B Standby Gas Treatment System (SBGT). From the indications provided in the question, the SBGT train remained in Standby.

C. Incorrect: The A SBGT train would receive a low flow autostart signal if it had received a start signal and was manually placed in Standby by an Operator. Plausible if the student believes any GP.3 Rad monitor would cause both trains to start and isolate this could be true.

D. Incorrect: An INOPERABLE signal created by the Radiation Monitor will cause start signal to the B Standby Gas Treatment System (SBGT). Plausible if the student does not correctly identify that a Fuel Pool Rad Monitors are an input to the GP. 3 Isolation Logic.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 49 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET 1C03A (C-1) Fuel Pool Exhaust RIS-4131A/B Rad Monitor Dnscl/Inop Rev 63 1C05B (C-8) PCIS Group 3 Technical Reference(s): (Attach if not previously provided)

Isolation Initiated Rev 108 OI-170 Section 4.1 Automatic Initiation of SBGT Rev 66 (page 8

&9)

Proposed References to be provided to applicants during examination: N 7.00.00.02 Given a SBGT system operating mode and various plant conditions, predict how the SBGT Learning Objective: (As available) system will be impacted by failures in the following support systems:

a. PCIS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 50 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295029 EK1.01 Importance Rating 3.4 295029 (EPE 6) High Suppression Pool Water Level / 5: EK1.01 - Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Containment integrity. (CFR: 41.8 to 41.10)

Proposed Question: RO Question 25 The basis for maintaining Torus level below 13.5 feet is to prevent ________.

A. a loss of Torus level indication B. covering the DW-to-Torus Vacuum Breakers C. HPCI from tripping due to high exhaust pressure D. SRV tailpipe damage with an SRV open Proposed Answer: B Explanation:

A. Incorrect: 13.5 feet in the Torus does not challenge the Torus level indication which becomes impacted at 16 feet. Plausible if the student confuses Torus level indication tap at 16 feet with the 13.5 feet given in the stem.

B. Correct: Initially Torus level is maintained below the elevation of the bottom of the torus to drywell vacuum breakers to preserve the operability of these valves and thereby permit operation of drywell sprays. These vacuum breakers will not function as designed if any portion of the valve is covered with water. Keeping torus level below 13.5 ft assures that no portion of the drywell side of the valve is submerged C. Incorrect: 13.5 feet in the Torus will not challenge HPCI exhaust pressure (150 psig).

The Torus maintained below these limits even during emergency operation keeping HPCI available for injection. Plausible if the student believes rising Torus level will cause elevated pressure in the Torus.

D. Incorrect: 13.8 feet is the SRV Tailpipe Level Limit which is the bottom of the Torus ring header at which point if an SRV were to be opened there is a possibility of damaging SRV tailpipe supports and containment structures. Plausible if the student applies this limit to the 13.5 foot limit in the Torus.

EOP-2 Bases Rev 16 page 16 Technical Reference(s): (Attach if not previously provided)

EOP Breakpoints Rev 14 page 9 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 51 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 95.00.00.15 Explain the Bases of each Learning Objective: (As available) of the EOP Curves and Limits Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: DAEC 2015 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8-10 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 52 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295034 G2.4.47 Importance Rating 4.2 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9: Generic K/A 2.4.47 -

Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 / 43.5 / 45.12)

Proposed Question: RO Question 26 The plant has experienced an event resulting in the following conditions:

• Reactor water level is 145 inches and rising slowly

• Reactor Building Vent Shaft radiation levels are 30 mrem/hr and rising slowly

• Reactor Water Cleanup Room temperatures are 185°F and rising slowly The Control Room Supervisor orders DEFEAT 9, Group 3 High DW Press & RX Low Level Isolation Defeat, installation to address the conditions above.

Is this order appropriate for plant conditions and why?

A. No, DEFEAT 9 cannot be installed with elevated radiation levels in the reactor building vent shaft B. No, DEFEAT 9 cannot be installed until RPV Level is raised above RX Lo-Lo Level Setpoint C. Yes, DEFEAT 9 will reset the SBGT lockout relays and allow restoration of Reactor Building Ventilation D. Yes, DEFEAT 9 will reset the SBGT lockout relays and allow SBGT to ventilate all areas of the Reactor Building Proposed Answer: A Explanation: EOP 3 Continuous Recheck Statement requires All the following conditions apply:

Reactor Building HVAC Isolated, Fuel Pool Exhaust RIS4131A(B) Radiation Level is below 8 mR/hr RB Vent Shaft RIM-7606A(B) Radiation Level is below 8 mR/hr Offgas Vent Pipe RM-4116A(B) is belwo HI-HI Rad Trip setpoint RB HVAC is the system normally used to maintain secondary containment temperature and Dp within operational limits. If isolated, it is appropriate to restart this system and use it to restore Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 53 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET and maintain control of secondary containment temperature and pressure once it is confirmed that restart will not result in excessive release of radioactivity to the environment.

A. Correct: Defeat 9 does reset the L/R-5830A & B but it only bypasses the High DW (2 psig) and low RPV level (170 inch) signals. EOP-3 CRS requires fuel pool exhaust, reactor building vent shaft and offgas vent pipe radiation levels to be below their group 3 setpoints prior to installing the defeat.

B. Incorrect Defeat 9 bypasses the Low RPV/L 170 inch signal and high drywell pressure it is not authorized to bypass high radiation. Plausible if the student believes that Defeat 9 will defeat High Radiation Level inputs to the GP. 3 Isolation logic. It will only Defeat the Hi DW pressure and LO RPV Level signals.

C. Incorrect DEFEAT 9 cannot be installed with elevated radiation levels in the reactor building vent shaft. Plausible if the student believes that Defeat 9 will defeat High Radiation Level inputs to the GP. 3 Isolation logic. It will only Defeat the Hi DW pressure and LO RPV Level signals.

D. Incorrect DEFEAT 9 cannot be installed with elevated radiation levels in the reactor building vent shaft. Plausible if the student believes the GP.3 Isolation has to be reset to allow SBGT to RUN. It is already running.

EOP Defeat 9 Rev 4 EOP-3 Rev 22 Continuous Technical Reference(s): (Attach if not previously provided)

Recheck Statement EOP-3 Bases Rev 13 (page 9)

Proposed References to be provided to applicants during examination: Defeat 9, Rev 4 95.00.00.20 Relate how each step and its performance meets the mitigation Learning Objective: (As available) strategies of the EOP support procedures Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/12 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 54 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295036 EA2.03 Importance Rating 3.4 Secondary Containment High Sump/Area Water Level - Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:

Cause of the high water level Proposed Question: RO Question 27 A reactor scram occurs as a result of a seismic event. The following conditions exist:

• A break in the cooling water supply lines to the HPCI and RCIC room coolers has occurred

• RPV pressure is 900 psig and stable

• HPCI room water level is above its Max Safe Operating Level

• RCIC room water level has just risen above its Max Safe Operating Level Which of the following describes the source of the leak and the action required to mitigate the condition?

The __(1)__ System piping is the source of the leak and an Emergency Depressurization

__(2)__ required?

A. (1) ESW (2) is B. (1) ESW (2) is NOT C. (1) GSW (2) is D. (1) GSW (2) is NOT Proposed Answer: B Explanation: Primary systems comprise the pipes, valves, and other equipment which connect directly to the RPV, a reduction in RPV pressure will effect a decrease in the flow of steam or water being discharged through an unisolated break in the system.

GSW does not supply the coolers. EOP-3 Bases point of emphasis. If a MSOL was reached in one area due to a leak and MSOL reached in another area because of a fire The BWROG Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 55 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET guidelines were reviewed and it was concluded that ED should not be performed based on the decisions that were previously made in step 4 (will RPV pressure reduction decrease leakage into secondary containment). An ESW leak is not a primary system leak. ED is not warranted.

A. Incorrect: ESW is the cooling water supply to the HPCI and RCIC essential room coolers.

GSW does not supply the coolers. EOP-3 Bases point of emphasis. If a MSOL was reached in one area due to a leak and MSOL reached in another area because of a fire.

The BWROG guidelines were reviewed and it was concluded that ED should not be performed based on the decisions that were previously made in step 4 (will RPV pressure reduction decrease leakage into secondary containment). An ESW leak is not a primary system leak. ED is not warranted.

B. Correct GSW does not supply the coolers. EOP-3 Bases point of emphasis. If a MSOL was reached in one area due to a leak and MSOL reached in another area because of a fire The BWROG guidelines were reviewed and it was concluded that ED should not be performed based on the decisions that were previously made in step 4 (will RPV pressure reduction decrease leakage into secondary containment). An ESW leak is not a primary system leak. ED is not warranted.

C. Incorrect GSW does not supply the coolers. EOP-3 Bases point of emphasis. If a MSOL was reached in one area due to a leak and MSOL reached in another area because of a fire The BWROG guidelines were reviewed and it was concluded that ED should not be performed based on the decisions that were previously made in step 4 (will RPV pressure reduction decrease leakage into secondary containment). An ESW leak is not a primary system leak. ED is not warranted.

D. Incorrect GSW does not supply the coolers. EOP-3 Bases point of emphasis. If a MSOL was reached in one area due to a leak and MSOL reached in another area because of a fire The BWROG guidelines were reviewed and it was concluded that ED should not be performed based on the decisions that were previously made in step 4 (will RPV pressure reduction decrease leakage into secondary containment). An ESW leak is not a primary system leak. ED is not warranted.

EOP-3 Bases Rev 13 (page 18 &

Technical Reference(s): (Attach if not previously provided) 19)

Proposed References to be provided to applicants during examination: N 33.03.01.02 Describe the flowpath of Learning Objective: (As available) the ESW system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 56 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 57 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 203000 A1.01 Importance Rating 4.2 203000 (SF2, SF4 RHR/LPCI) RHR/LPCI: Injection Mode: A1.01 - Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) controls including: Reactor water level. (CFR: 41.5 / 45.5)

Proposed Question: RO Question 28 A Loss of Offsite Power has occurred and RPV Emergency Depressurization is in progress.

• A and B Core Spray pumps failed to automatically START and will not run

• All remaining Low Pressure Emergency Core Cooling Systems (ECCS) pumps are running on minimum flow

• Reactor pressure is 600 psig and lowering In accordance with EOP 1, RPV Control, below which reactor pressure will the Balance of Plant Operator first observe reactor water level being to rise?

A. 450 psig B. 330 psig C. 260 psig D. 135 psig Proposed Answer: C Explanation: EOP-1 Rev 20 Table 1A Preferred Injection Systems Condensate/ Feedwater Shutoff Head is 650 psig (Loss of Offsite Power 1A1 and 1A2 will not transfer to the Startup Transformer on Turbine Trip and have no power)

Core Spray Shutoff Head is 330 psig Given in stem that both Core Spray systems have failed to operate.

RHR Shutoff Head is 260 psig Next available system is RHR with a shutoff head of 260 psig A. Incorrect: This is the pressure in the reactor vessel which the low pressure Emergency Core Cooling Systems (ECCS) inject valves will receive an OPEN signal. At this pressure, each low pressure ECCS pumps will not have sufficient pressure to cause indicated flow at 1C03. Additionally with a LOOP 1A1 and 1A2 will be unavailable to power the condensate pumps 1P8A/B.

B. Incorrect: EOP 1 identify that the nominal shutoff head for Core Spray is 330 psig. From the question stem, both core spray pumps are not available.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 58 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Correct: At this pressure LPCI will be injecting at designed flow rates, however this is not the pressure which the Balance of Plant Operator will first observe Core Spray flow indication. This is the pressure identified for Low Pressure Coolant Injection (LPCI)

System will demonstrate system flow into the vessel and reactor water level will rise.

D. Incorrect: At this pressure LPCI will be injecting at designed flow rates, however this is not the pressure which the Balance of Plant Operator will first observe Core Spray flow indication. This is the pressure which the Shutdown Cooling Interlocks will clear.

EOP-1 Table 1A Preferred Technical Reference(s): (Attach if not previously provided)

Injection Systems. Rev 18 EOP-1 Bases Rev 18 (page 22 of 72)

Proposed References to be provided to applicants during examination: N 95.00.00.15 Explain the Bases of each Learning Objective: (As available) of the EOP Curves and Limits Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.45 5 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 59 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 203000 A2.06 Importance Rating 3.8 203000 (SF2, SF4 RHR/LPCI) RHR/LPCI: Injection Mode: A2.06 - Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Emergency generator failure. (CFR: 41.5 / 45.6)

Proposed Question: RO Question 29 The plant has experienced a Loss of Coolant Accident with a Loss of Offsite Power.

• Both SBDGs started and powered their respective Essential Busses

• RPV Level recovers to 120 inches and is rising slowly with LPCI injection The B SBDG catastrophically fails. The following indications are observed:

• 1C03B A-2 RHR PUMP 1P-229 A TRIP OR MOTOR OVERLOAD is in alarm

• 1C03B A-8 RHR PUMP 1P-229 B TRIP OR MOTOR OVERLOAD is in alarm

• 1C03B A-3 RHR PUMP 1P-229 C TRIP OR MOTOR OVERLOAD is in alarm

• 1C03B A-9 RHR PUMP 1P-229 D TRIP OR MOTOR OVERLOAD is in alarm (1) What is the current status of LPCI following these events AND (2) What, if any, action should be taken?

(1) A and C RHR Pumps will continue to run with no change in pump flowrate A.

(2) Place both RHRSW pumps in service per OI 149 QRC 2, Initiate Torus Cooling (1) A and C RHR Pumps will continue to run with elevated pump flowrate B.

(2) Lower LPCI flowrate to reduce pump amps per ARP 1C03B A-2, A-3 (1) B and D RHR Pumps will continue to run with no change in pump flowrate C. (2) Place both RHRSW pumps into service per OI 149 QRC 2, Initiate Torus Cooling (1) B and D RHR Pumps will continue to run with elevated pump flowrate D.

(2) Lower LPCI flowrate to reduce pump amps per ARP 1C03B A-8, A-9 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 60 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed Answer: B Explanation: LPCI will have an initiation signal and be injecting through the select RHR Loop (B is the default injection loop). 3 RHR Pumps are required for full LPCI injection flow 14400 GPM (4800 GPM per pump). With the A & B SBDG supplying power to their respective essential buses 1A3 and 1A4 A/C will be powered by 1G31 (A SBDG), B/D will be powered by 1G21 (B SBDG). When the B SBDG fails B/D RHR pumps will trip on loss of power. A/C will remain powered but still be injecting through the full open LPCI inject valves, RHR pump flow will rise for the operating pumps. With pump amps provided in the stem to be in the RED region, ARP actions for the running LPCI pumps should be taken to lower pump flowrate to lower pump amps prior to the pump breaker TRIP.

A. Incorrect: When the B SBDG fails B/D RHR pumps will trip on loss of power. A/C will remain powered but still be injecting through the full open LPCI inject valves, RHR pump flow will rise for the operating pumps. The student may incorrectly conclude that flowrate remains unchanged for these conditions. RHRSW being placed in service would be the next logical sequence as a mitigating action however the annunciator condition should be addressed first. In addition, you should not place both RHRSW Pumps and both RHR Pumps in service when the essential bus is being powered by the SBDG B. Correct: When the B SBDG fails B/D RHR pumps will trip on loss of power. A/C will remain powered but still be injecting through the full open LPCI inject valves, RHR pump flow will rise for the operating pumps. With pump motor overload alarm provided in the stem. ARP actions for the running LPCI pumps should be taken to lower pump flowrate to lower pump amps prior to the pump breaker TRIP.

C. Incorrect: When the B SBDG fails B/D RHR pumps will trip on loss of power. Plausible in that the annunciator for a pump breaker TRIP condition and Motor Overload condition is the same. RHRSW being placed in service would be the next logical sequence as a mitigating action however the annunciator condition should be addressed first. In addition, you should not place both RHRSW Pumps and both RHR Pumps in service when the essential bus is being powered by the SBDG D. Incorrect: When the B SBDG fails B/D RHR pumps will trip on loss of power. Plausible in that the annunciator for a pump breaker TRIP condition and Motor Overload condition is the same.

SD-149 RHR, Rev 14 (page 9)

AOP-301 rev 75 (page 12 of 58 )

Technical Reference(s): Auto Actions Load shedding of (Attach if not previously provided) loads ARP 1C03B A-2, A-3 Proposed References to be provided to applicants during examination: N 2.01.01.06 Given an RHR system Learning Objective: operating mode and various plant (As available) conditions predict how the RHR Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 61 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET system will be impacted by operation, or failure of the following support systems Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 62 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 205000 K6.04 Importance Rating 3.6 205000 (SF4 SCS) Shutdown Cooling: K6.04 - Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): Reactor water level. (CFR: 41.7 / 45.7)

Proposed Question: RO Question 30 The plant was in shutdown cooling when a leak developed on the running recirculating pump discharge piping.

• RPV water level is lowering At what level will shutdown cooling FIRST become unavailable?

A. 186 inches B. 170 inches C. 119 inches D. 64 inches Proposed Answer: B Explanation: Group 4 is the RHR Shutdown Cooling Isolation. Group 4 isolation signals are RPV level 170 inches, Drywell pressure 2 psig, and RPV pressure < 135 psig.

MO-1908 and 1909 SDC inboard and outboard suction valves will close on the 170 inch RPV level signal.

A. Incorrect 186 is the low RPV level alarm signal. It does not result in a Group 4 isolation signal.

Plausible if the student believes that a low level alarm would result in a Group 4 isolation signal. In addition this level is utilized in the single feed pump reactor recirc runback signal.

B. Correct 170 is the SDC Group 4 signal C. Incorrect 119.5 is a Group 5 RWCU isolation signal. Plausible if the student believes that the LO-LO level is the initiator of the Group 4 isolation signal.

D. 64 is a Group 1 and 7 isolation signal. Plausible if the student believes that the LO-LO-LO level is the initiator of the Group 4 isolation signal.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 63 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET SD 959.1 Rev 13 Primary Containment Isolation System Page 10, Table 1 (PCIS Isolation Signals)

Technical Reference(s): ARP 1C05A Rev 90, (Attach if not previously provided)

A-1 Reactor Lo-Lo-Lo Level Trip B-1 Reactor Lo-Lo Level Trip C-1 Reactor Lo Level Trip Proposed References to be provided to applicants during examination: N 2.03.01.04 Describe the RHR system interlocks, including purpose, Learning Objective: (As available) setpoints, logic, and when/how they are bypassed, overridden, or reset Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 64 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 206000 K4.02 Importance Rating 3.9 Knowledge of HIGH PRESSURE COOLANT INJECTION SYSTEM (HPCIS) design feature(s) and/or interlocks which provide for the following: System Isolation: Plant-Specific.

Proposed Question: RO Question 31 The plant is operating at 100% reactor power when a steam leak developed in the Torus Area.

• Ambient temperature has risen to 155°F and has stabilized due to Operator actions Which of the following is correct for the current plant status?

A. HPCI will isolate in 15 minutes B. RCIC will isolate in 15 minutes C. HPCI and RCIC should have immediately isolated at 150°F D. HPCI and RCIC will isolate when ambient temperature reaches 165°F Proposed Answer: A Explanation: At 150°F in the Torus Area a 15 minute delay starts and isolate HPCI after 15 minutes. RCIC TD is 30 minutes A. Correct HPCI isolates 15 minutes after > 150°F in Torus area B. Incorrect RCIC isolation time delay is 30 minutes. Plausible if the student believes that RCIC isolates in 15 minutes.

C. Incorrect Each HPCI and RCIC will isolate after their respective time delays of 15 and 30 minutes. Plausible if candidate believes that both HPCI and RCIC will isolate IMMEDIATELY upon reaching the temperature isolation setpoint.

D. Incorrect: 165°F is the EOP-3 Max Safe Operating Limit for the Torus area but not a direct SLD input. . Plausible if the students believe that the isolation setpoint is 165°F for both systems.

ARP 1C04B (B-4) Rev 85 Technical Reference(s): SD-858 Steam Leak Detection (Attach if not previously provided)

Rev 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 65 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET SD-959.1 Primary Containment Isolation System Rev 13 (Page Page 10, Table 1 (PCIS Isolation Signals)

Proposed References to be provided to applicants during examination: N 5.06.01.07 Describe how the HPCI Learning Objective: (As available) system responds to an isolation signal Question Source: Bank #

X (2015 NRC #25)

Modified Bank # Question attached (Note changes or attach parent) below New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 66 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 67 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 68 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 209001 K5.01 Importance Rating 2.6 209001 (SF2, SF4 LPCS) Low Pressure Core Spray: K5.01 - Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM: Indications of pump cavitation. (CFR: 41.5 / 45.3)

Proposed Question: RO Question 32 The plant is experiencing a LOCA. Core Spray is injecting with the following plant conditions:

• RPV Level is +90 inches and slowly rising

• A Core Spray is injecting at 3300 GPM

• B Core Spray is injecting at 3250 GPM The Balance of Plant Operator notices that both Core Spray Discharge pressures are oscillating.

What actions, if any, should the Operator take for Core Spray?

A. None, the Core Spray pumps are injecting within their design capabilities B. The Core Spray pumps flow should be lowered to prevent pump runout conditions C. The Core Spray pumps flow should be raised to meet injection requirements D. The Core Spray pumps flow should be lowered to prevent fuel damage Proposed Answer: B Explanation: Section 4.0 Automatic Startup / Initiation of the Core Spray System. Step 4 verifies system parameters (flow < 3100 gpm)

Step 5 As RPV pressure lowers, throttle INBD Inject MO-2117 (MO-2137) valve using HS-2117 (HS-2137) on 1C03 to maintain <3100 gpm.

Note in section also states maintaining flow <3100 gpm is operational guidance to ensure pump run out does not occur while RPV pressure lowers A. Incorrect: Core spray is injecting above 3100 gpm. OI 151 section 4 states As RPV pressure lowers, throttle INBD Inject MO-2117 (MO-2137) valve using HS-2117 (HS-2137) on 1C03 to maintain <3100 gpm. Plausible if the student believes throttling flow is inappropriate given the conditions provided.

B. Correct: Section 4.0 Automatic Startup / Initiation of the Core Spray System. Step 4 verifies system parameters (flow < 3100 gpm)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 69 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Step 5 As RPV pressure lowers, throttle INBD Inject MO-2117 (MO-2137) valve using HS-2117 (HS-2137) on 1C03 to maintain <3100 gpm.

Note in section also states maintaining flow <3100 gpm is operational guidance to ensure pump run out does not occur while RPV pressure lowers.

C. Incorrect: As RPV pressure lowers, flow will increase and should be lowered. This is incorrect since RPV level is at the point where Operations strategies require the operator to throttle injection as necessary to maintain the core submerged within operating bands. Plausible if the student believes flowrate is insufficient and should be raised.

D. Incorrect: This choice is plausible if the student believes the pumps will fail resulting in loss of injection capability and eventual core uncovery. In addition there are limitations for the use of Core Spray during ATWS conditions since Core Spray injects within the core shroud and directly above the fuel.

OI-151 Core Spray Rev 85 (page Technical Reference(s): (Attach if not previously provided) 7)

Proposed References to be provided to applicants during examination: N 95.00.00.14 Evaluation plant status and control room indications to Learning Objective: (As available) determine the applicability and effect of any EOP caution Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.45 3 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 70 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 209001 A1.08 Importance Rating 3.3 209001 (SF2, SF4 LPCS) Low Pressure Core Spray: A1.08 - Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including: System lineup. (CFR: 41.5 / 45.5)

Proposed Question: RO Question 33 The plant was operating at 100% reactor power when a Recirc line break occurred. The following conditions are now present:

• Reactor pressure is at 410 psig and stable

• A and B Core Spray pumps have automatically initiated

• CORE SPRAY LOOP A INBD INJECT MO 2117 is CLOSED

• CORE SPRAY LOOP B INBD INJECT MO 2137 is CLOSED

• Core Spray MIN FLOW BYPASS VALVES MO 2104 and MO 2124 are OPEN Based on the conditions above which of the following describes the response of the Core Spray System valves and if any operator actions are required?

The Core Spray Inboard Injection Valves __________(1)_______ and the Core Spray MIN Flow Bypass Valves will auto-close ________(2)________.

A. (1) should have opened and must be manually opened (2) when Core Spray system flow reaches 600 gpm B. (1) should have opened and must be manually opened (2) ONLY when the Injection Valves are fully open C. (1) will open once reactor pressure lowers to below the shut off head of the Core Spray pumps (2) when Core Spray system flow reaches 600 gpm D. (1) will open once reactor pressure lowers to below the shut off head of the Core Spray pumps (2) ONLY when the Injection Valves are fully open Proposed Answer: A Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 71 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Explanation: When reactor vessel pressure drops below the low pressure permissive setpoint of 450 psig, verify that the INBD INJECT MO 2117 [MO 2137] valves OPEN to inject to the reactor vessel. The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating to "verify" they open.

When system flow reaches 600 gpm, as indicated on (A[B] CORE SPRAY PUMP)

INJECT/TEST FLOW indicator FI 2110 [FI 2130] on Panel 1C03, verify MIN FLOW BYPASS MO 2104 [MO 2124] valve CLOSES.

Proposed Answer: A A. Correct - When reactor vessel pressure drops below the low pressure permissive setpoint of 450 psig, verify that the INBD INJECT MO 2117 [MO 2137] valves OPEN to inject to the reactor vessel. The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating to "verify" they open.

When system flow reaches 600 gpm, as indicated on (A[B] CORE SPRAY PUMP)

INJECT/TEST FLOW indicator FI 2110 [FI 2130] on Panel 1C03, verify MIN FLOW BYPASS MO 2104 [MO 2124] valve CLOSES.

B. Incorrect - The min flow bypass valve will automatically close when system flow reaches 600 gpm.

C. Incorrect - The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating to "verify" they open.

D. Incorrect - The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating to "verify" they open.The min flow bypass vlv will close when system flow reaches 600 gpm.

OI 151, pgs 6 & 7, steps 4.0 (2) &

Technical Reference(s): (3). (Attach if not previously provided)

SD-151 rev 8 page 13 Proposed References to be provided to applicants during examination: N 4.02.01.07 List the signals which cause a Core Spray system Auto Learning Objective: Initiation, including setpoints and logic. (As available)

Describe how they are bypassed and how they are reset Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 72 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.45 5 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 73 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 21100 K3.01 Importance Rating 4.3 211000 (SF1 SLCS) Standby Liquid Control: K3.01 - Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM will have on following: Ability to shut down the reactor in certain conditions. (CFR: 41.7 / 45.4)

Proposed Question: RO Question 34 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 74 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET The plant has experienced an ATWS, the Control Room Supervisor has directed the initiation of Standby Liquid Control System (SBLC).

• The SBLC system mode switch HS-2613 has been placed to PUMPS A and B RUN The following is observed:

What is the status of the SBLC system?

A. The squib valves still have continuity B. The system is operating as designed C. The squib valves fired but the squib valves failed to provide a flow path D. Sodium Pentaborate has precipitated out of solution blocking the pump suction line Proposed Answer: C Explanation:

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 75 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET A. Incorrect - The squib valves ready lights OFF and the squib continuity loss annunciator ON indicates that the valves have fired (lost continuity).

B. Incorrect The SLC tank level is steady and discharge pressure is above the relief valve setpoint.

C. Correct - The squib valves ready lights OFF and the squib continuity loss annunciator ON indicates that the valves have fired. However the relief valve is lifting and relieving back to the pump suction (1350 psig is the lift point) because there is no discharge path to the reactor. Of the choices given the only possible explanation is that squib valves fired but the squib valves failed to open D. Incorrect - With the pump suction line blocked there would NOT be any discharge pressure.

OI 153,Rev 45 pg 6 Manual Startup/Initiation of the SBLC Technical Reference(s): (Attach if not previously provided) system OI-153 QRC Rev 4 Proposed References to be provided to applicants during examination: N 6.00.00.05 Describe how the Standby Learning Objective: Liquid Control System responds to (As available) and initiation signal Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 4 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 76 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 212000 K2.01 Importance Rating 3.2 212000 (SF7 RPS) Reactor Protection: K2.01 - Knowledge of electrical power supplies to the following: RPS motor-generator sets. (CFR: 41.7)

Proposed Question: RO Question 35 Which of the following are the normal power supplies to the A RPS MG Set and B RPS MG Set?

A. 1B32, CB 480VAC Essential Motor Control Center 1B42, CB 480VAC Essential Motor Control Center B. 1B33, Turbine Building 480 VAC Motor Control Center 1B43, RB 757 Level 480 VAC Motor Control Center C. 1B34, RB 786 Level 480 VAC Motor Control Center 1B44, RB 757 Level 480 VAC Motor Control Center D. 1B35, RB 786 Level 480 VAC Motor Control Center 1B45, Turbine Building 480 VAC Motor Control Center Proposed Answer: A Explanation: 1B32 and 1B42 are the 480 VAC essential MCCs that power the RPS MG sets A. Correct: 480 VAC Essential MCCs are the RPS MG Set power Supplies B. Incorrect: These are 480 VAC essential MCCs. Plausible if the student believes these essential MCCs provide power to the RPS MG sets.

C. Incorrect: These are 480 VAC essential MCCs however they do not provide power to the RPS MG Sets.

D. Incorrect: These are 480 VAC essential MCCs however they do not provide power to the RPS MG Sets.

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 77 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 78 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 212000 K6.01 Importance Rating 3.6 212000 (SF7 RPS) Reactor Protection: K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM: A.C. electrical distribution. (CFR: 41.7 / 45.7)

Proposed Question: RO Question 36 A grid disturbance has occurred and essential bus voltage has lowered to 90% of rated voltage for 15 seconds.

What is the effect on the reactor protection system?

A. RPS busses remain energized due to the 1Y16/1Y26, Regulated Power AC Distribution Panel auto transfer function.

B. RPS bus voltage and frequency are maintained throughout the transient by the RPS MG Set Flywheel.

C. RPS busses de-energize and remain de-energized due to the trip of EPA breakers.

D. RPS busses de-energize and subsequently re-energize on the 1Y16/1Y26, Regulated Power AC Distribution Panel auto transfer function.

Proposed Answer: C Explanation:

A. Incorrect. 1Y16/1Y26, Regulated Power AC Distribution Panel does not have an auto transfer function. This could be confused with instrument AC power functionality.

B. Incorrect. The RPS MG Set Flywheel is designed for momentary fluctuations in MG set frequency and voltage (less than one second). This is not the design function of the MG set flywheel. Plausible if the candidate believes it will maintain MG set parameters for >

15 seconds.

C. Correct: Essential busses would de-energize on degraded bus voltage of 3798Volts for 8-8.5 seconds. This would result in the loss of power to 1A3 and 1A4 and subsequently RPS A and RPS B MG sets. Flywheels could not sustain voltage for extended loss of power.

D. Incorrect. 1Y16/1Y26, Regulated Power AC Distribution Panel does not have an auto transfer function. This could be confused with instrument AC power functionality.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 79 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Plausible if the candidate confuses an AUTO transfer similar to uninterruptible power transfer/

Technical Reference(s): SD-358 RPS Rev 9 (page 8) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 22.00.00.07 Describe the operation of the following principal Reactor Learning Objective: Protection system components: (As available)

a. Reactor Protection System Motor Generator Sets Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 80 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215003 K1.01 Importance Rating 3.9 215003 (SF7 IRM) Intermediate Range Monitor: K1.01 - Knowledge of the physical connections and/or cause effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: RPS. (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Proposed Question: RO Question 37 The Mode Switch is in STARTUP.

• The plant experiences a Loss of 24 VDC System 1 Which of the following describes the effect on RPS for this loss?

A. No impact to RPS B. Full Reactor SCRAM C. A RPS AUTO SCRAM only D. One RPS Backup Scram valve OPENS Proposed Answer: C Explanation: Loss of 24 VDC with the Reactor Mode Switch not in Run will generate a RPS A (B) Half Scram due to an INOP signal from the IRMS in division 1.

A. Incorrect: The loss of 24VDC system 1 results in IRM inop signal for A/C/E IRMs only.

This results in an INOP signal to the IRMs which will generate an RPS scram signal.

Plausible if the candidate does recognize the IRM INOP signal to RPS resulting from this loss.

B. Incorrect: The loss of 24VDC system 1 results in IRM inop signal for A/C/E IRMs only.

This results in an INOP signal to the IRMs which will generate an RPS scram signal.

Plausible if the student believes the circuitry is similar to SRM coincidence SCRAM circuitry when enabled would result in a FULL SCRAM.

C. Correct: Loss of 24 VDC Div 1(Div2) with the Reactor Mode Switch not in Run will generate a RPS A(B) Half Scram D. Incorrect: Backup SCRAM valves energize to OPEN and they are powered from 125VDC not 24VDC. Plausible in that scram air valves de-energize OPEN when RPS power is lost.

Technical Reference(s): AOP-375 Loss of 24 VDC Rev 27 (Attach if not previously provided)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 81 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET 1C05A (E-4) Rev 90 SD 358 Figure 3 RPS Trip System Simplified Rev 9 (page 42)

Proposed References to be provided to applicants during examination: N Given an IRM system operating mode and vatious plant conditions, p[redict how the IRM system will be impacted Learning Objective: (As available) by failures in the following support systems:

c. DC Electrical System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2-9 55.45 8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 82 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215004 G2.4.45 Importance Rating 4.1 215004 (SF7 SRMS) Source Range Monitor: Generic K/A 2.4.45 - Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.12)

Proposed Question: RO Question 38 The reactor is in MODE 2 with a reactor startup commencing. All control rods are inserted.

When attempting to withdraw the first control rod the following annunciators were observed:

• 1C05A (D-2), APRM Downscale

• 1C05A (D-3), IRM Downscale

• 1C05A (D-4), LPRM Downscale

• 1C05A (D-5), SRM Downscale

• 1C05B (A-6), Rod Out Block Based on annunciator significance, which of the following annunciators should have first priority to allow control rod withdrawal?

A. 1C05A (D-2), APRM Downscale B. 1C05A (D-3), IRM Downscale C. 1C05A (D-4), LPRM Downscale D. 1C05A (D-5), SRM Downscale Proposed Answer: D Explanation: With the Mode Switch in Startup and a SRM downscale activated a Rod Block is activated to prevent control rod withdrawal. This condition must be addressed first to allow control rod withdrawal to commence.

A. Incorrect The APRM downscale is not priority in the STARTUP mode. This would be plausible with the mode switch in RUN.

B. Incorrect The IRM downscale is not a priority due to not being in the intermediate range at the beginning of the startup. This would be plausible if the candidate does not remember the IRM downscale rod block is bypassed with the associated IRM range switch on 1.

C. Incorrect Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 83 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET The LPRM downscale is not a priority due to being expected during reactor startup initial conditions. This would be plausible because LPRMs could result in a RBM inop signal which during plant operation would result in a ROD BLOCK when >30% power.

D. Correct. With the Mode Switch in Startup and a SRM downscale activated a Rod Block is activated to prevent control rod withdrawal. This condition must be addressed first to allow control rod withdrawal to commence.

SD 878.1 Rev 7 Source Range Monitoring System Figure 10 SRM Technical Reference(s): (Attach if not previously provided)

Channel Trip Outputs to RMCS page 41 Proposed References to be provided to applicants during examination: N 78.02.01.06 Describe the SRM system interlocks (include alarms) including Learning Objective: (As available) the purpose, setpoints, logic, and when/how they are bypassed.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/12 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 84 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215005 K5.06 Importance Rating 2.5 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor: K5.06 -

Knowledge of the operational implications of the following concepts as they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: Assignment of LPRM's to specific APRM channels. (CFR: 41.5 / 45.3)

Proposed Question: RO Question 39 The plant is operating at 100% reactor power with the following plant conditions:

• B and C APRMs are bypassed

• LPRM 2A-08-33 is assigned to the APRM B cabinet

• LPRM 2A-08-33 slowly fails upscale Which of the following describes the operational implication of the LPRM failure?

A. Only A APRM will indicate higher than before the failure.

B. Only B APRM will indicate higher than before the failure.

C. APRMs A and B will indicate higher than before the failure.

D. APRMs A and B will indicate lower than before the failure.

Proposed Answer: C Explanation: LPRM 2A-08-33 is shared between A and B APRMs and with its output failing upscale, the associated APRMs will indicate a higher power.

A. Incorrect - The B APRM indication will also rise due to the 2A-08-33 failing upscale.

Plausible in that the A APRM is unbypassed and the student may assume that only A APRM would rise due to not being bypassed.

B. Incorrect - The A APRM indication will also rise due to the 2A-08-33 failing upscale.

Plausible in that the B APRM cabinet is where 2A-08-33 is assigned.

C. Correct - LPRM 2A-08-33 is shared between A and B APRMs and with its output failing upscale, the associated APRMs will indicate a higher power.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 85 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET D. Incorrect - The student may assume that there is an auto bypass feature when the LPRM output exceeds predicted output values. This is similar to PANACEA rejected for the core thermal monitoring program.

SD-878.3, Rev 12; Pages 37-38 Technical Reference(s): (Attach if not previously provided)

SD-900, Rev. 7, pgs. 6-8 Proposed References to be provided to applicants during examination: N 80.01.01.03 Describe the interrelationships between the LPRMs, Learning Objective: the APRMs, and the RBM systems, (As available) including the effects on one from an operation or failure of the other Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 3 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 86 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215005 A4.05 Importance Rating 3.4 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor: A4.05 -

Ability to manually operate and/or monitor in the control room: Trip bypasses. (CFR: 41.7 / 45.5 to 45.8)

Proposed Question: RO Question 40 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 87 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET The plant is at 100% reactor power with the following initial APRM E Detectors conditions: LPRM (Status) 2A-16-33 Bypassed

• APRM A and APRM B are bypassed to support I&C maintenance 5A-32-17 OK

• An edge rod is selected 3A-16-25 OK The I&C Technician inadvertently bypasses the following LPRMs 6A-32-09 Bypassed associated with APRM E.

3B-24-25 OK NOTE: The current status of the APRM E detectors is as shown on the chart to the right. 4B-08-09 Bypassed 4B-08-17 Bypassed Prior to any Operator action, what is the expected plant response to these actions? 5B-40-17 OK 3B-24-33 OK 3C-32-33 Bypassed 4C-16-17 OK 1C-16-41 Bypassed 4C-32-25 OK 5C-16-09 OK 1D-24-41 Bypassed 2D-08-25 Bypassed 3D-40-25 OK 4D-24-09 OK 2D-08-33 OK 4D-24-17 OK A. RBM Downscale B. Control Rod Block, ONLY C. Half SCRAM and Control Rod Block D. Full SCRAM and Control Rod Block Proposed Answer: C Explanation: APRM E requires 13 LPRM inputs to be Operable. In current configuration it only has 12 LPRM inputs. This would generate an INOP signal from E APRM Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 88 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET A. Incorrect, no re-null signal, provided in the stem, an edge rod is selected. RBM is bypassed so, no RBM downscale will be generated. This would be possible provided a non-edge rod was selected.

B. Incorrect, APRM E is not bypassed, so an A RPS INOP will also be generated C. Correct, too few LPRM inputs will generate the E APRM INOP signal causing a Rod Block and A Side Half Scram D. Incorrect, not a full scram signal too few LPRM inputs will generate the E APRM INOP signal causing a Rod Block and A Side Half Scram ARP 1C05A (B-2)

Technical Reference(s): SD 878.3 Rev 12 page 37, 38 (Attach if not previously provided)

OI-878.4 Rev 43, P&L 3 Proposed References to be provided to applicants during examination: N 81.01.01.06 Given an APRM system operating mode and various plant conditions, predict how the APRM Learning Objective: system will be impacted by the (As available) operation or failure of the following support system or components:

c. LPRMS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: 2013 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 5-8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 89 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 217000 A3.02 Importance Rating 3.6 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling: A3.02 - Ability to monitor automatic operations of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) including: Turbine startup. (CFR: 41.7 / 45.7)

Proposed Question: RO Question 41 The RCIC Turbine is in service injecting to the RPV following automatic initiation. RCIC is aligned to its normal suction path when the following annunciator is received:

• 1C04C D-5 RCIC LOW SUCTION PRESSURE Which one of the following describes the response of the RCIC system for this condition?

A. RCIC suction will swap from the CST to the TORUS and continue to operate B. RCIC suction will swap from the TORUS to the CST and continue to operate C. RCIC will AUTO ISOLATE D. RCIC Turbine will TRIP Proposed Answer: D Explanation: Low Suction Pressure is a RCIC Turbine TRIP.

A. Incorrect, suction swap from CST to Torus is on CST low level, not low suction pressure. Plausible if the student thinks that RCIC suction would realign on low suction pressure.

B. Incorrect, no low suction pressure suction swap Plausible if the student thinks the TORUS is the normal suction path and that it would realign on low suction pressure.

C. Incorrect, this is not an Auto Isolation signal. Plausible if the student perceives this as an isolation signal vice a TRIP signal.

D. Correct ARP 1C04C (D-5)

SD-150 Rev 9 Page 10-11 Technical Reference(s): OI-150 Rev 88 Page 50 (Appendix (Attach if not previously provided) 1 RCIC Turbine Trips and Isolations)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 90 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 3.08.01.04 Describe the RCIC System interlocks, including purpose, Learning Objective: (As available) setpoints, logic, and when/how they are bypassed Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 91 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 218000 A2.06 Importance Rating 4.2 218000 (SF3 ADS) Automatic Depressurization: A2.06 - Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS Initiation Signals present. (CFR: 41.5 / 45.6)

Proposed Question: RO Question 42 The plant was operating at 100% power with HPCI out of service for planned maintenance. A LOCA concurrent with a loss of offsite power occurs:

• RCIC has initiated and has been maximized

• CRD has been maximized

• SBLC is injecting

• RPV water level is 80 inches lowering 5 inches per minute

• RPV pressure is 600 psig and lowering 10 psig per minute

• Drywell pressure is 8 psig and rising 1 psig per minute

• Drywell air temperature is 210°F and rising 2°F per minute

• A RHR is in Torus Spray mode Which one of the following is the FIRST action that should be taken?

A. Close the MSIVs B. Anticipate Emergency Depressurization C. Lockout Automatic Depressurization System D. Install Defeat 4 DRYWELL COOLER ISOLATION DEFEAT Proposed Answer: C Explanation:

EOP-1 CRS if ADS Timer has initiated Lockout ADS: Undesirable for the following reasons:

ADS actuation can cause impose a severe thermal transient on RPV and may complicate efforts to control RPV level. The given parameters show that ADS actuation is impending and requires immediate attention to prevent an inadvertent actuation. If only steam driven systems are available for injection ADS actuation may directly lead to loss of adequate core cooling. In addition, the operating crew can draw on much more information than is available to ADS logic Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 92 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET and can better judge based on instructions contained in the EPGs/SAGs when and how to depressurize the RPV.

A. Incorrect: The MSIV closed on the loss of RPS power when offsite power was lost. This is plausible if the student does not recognize the impact of the loss of offsite power.

Closing the MSIVs would be an action to mitigate loss of inventory.

B. Incorrect: The MSIV closed on the loss of RPS power when offsite power was lost. This is plausible if the student does not recognize the impact of the loss of offsite power.

Lowering reactor pressure is plausible if the student assumes this action is appropriate to make low pressure injection systems available to address the lowering RPV level.

C. Correct: ADS is overridden per EOP directions. See above D. Incorrect: The LOOP LOCA load shed results in a loss of Drywell Cooling fans and well water pumps rendering defeat 4 ineffective. This is plausible if the student does not recognize the LOOP LOCA load shed impact to the essential busses.

SD-183 Rev 9, ALC leg of EOP-1 RC/L-2 locks Technical Reference(s): out ADS Rev 20 (Attach if not previously provided)

EOP-1 Continuous Recheck Statement Proposed References to be provided to applicants during examination: N 8.00.00.01 State the purpose of the Learning Objective: (As available)

ADS System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 93 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 223002 A1.02 Importance Rating 3.7 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff: A1.02 -

Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: Valve closures. (CFR: 41.5 / 45.5)

Proposed Question: RO Question 43 Which of the following conditions will DIRECTLY cause a Primary Containment Isolation Valve to change position?

A. RPV Level at 211 inches B. RPV Pressure at 1140 psig C. HPCI Steam Line Flow at 150%

D. RWCU HX Room ambient temperature at 135°F Proposed Answer: D Explanation:

A. Incorrect, RPV level at 211 is a RFP and Turbine trip but not a direct PCIS signal.

Plausible if the student believes RPV Level of 211 inches is a PCIS signal. It only results in a TRIP signal to the HPCI, RCIC, Main Turbine and Reactor Feed Pumps.

B. Incorrect, 1140 PSIG is RPT breaker trip signal, not a direct PCIS signal. Plausible if the student believes this is a PCIS signal. This pressure does result in an RPT and ARI TRIP signal however no isolation signal is generated as result of this signal.

C. Incorrect, HPCI PCIS isolation is 300% . Plausible if the student believes the isolation setpoint for HPCI steam flow is 150% rated. Similar to the Group 1 Hi steam flow isolation setpoint.

D. Correct: PCIS Group 5 signal is 130°F in the RWCU Hx Room SD-959.1 Rev 13 (page 10) Table Technical Reference(s): (Attach if not previously provided) 1 PCISW Isolation Signals Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 94 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.45 5 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 95 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 239002 K6.04 Importance Rating 3.0 239002 (SF3 SRV) Safety Relief Valves: K6.04 - Knowledge of the effect that a loss or malfunction of the following will have on the RELIEF/SAFETY VALVES: D.C. power: Plant-Specific. (CFR: 41.7 / 45.7)

Proposed Question: RO Question 44 The plant is operating at 100% reactor power when the following events occur:

• 1D13 Circuit 14 "AUTO BLOWDOWN RELAY PANEL 1C45" trips

• Annunciator 1C03A (C-6) ADS/LLS 125 VDC CONTROL POWER FAILURE alarms Regarding ADS operation, which ONE of the following describes the effect of the breaker trip?

A. "A" ADS logic shifts to its alternate power supply so there is ONLY a temporary loss of power to the ADS logic B. "A" ADS logic has lost power; however, all 4 ADS SRVs have control power, and ADS remains functional via the B ADS Logic C. "A" ADS logic has lost power; however, ONLY 2 ADS valves have control power and will open upon ADS initiation D. "A" ADS logic shifts to its alternate power supply; however, control power is lost to 2 ADS valves, therefore, these valves will NOT open during ADS initiation Proposed Answer: B Explanation: ADS Logic A does not have a backup 125 VDC supply A. Incorrect: A ADS Logic has no backup 125 VDC Power supply. Plausible if the operator believes that the A ADS logic has a backup power supply. It does not.

B. Correct ADS Relief Valves PSV4400,4402, 4405, 4406 have control power via 1D23-14 C. Incorrect: all ADS PSVs will function. 4401 and 4407 are LLS SRVs. Plausible if the operator believes that the ADS Valves do not have a backup power supply. All 4 ADS valves will function.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 96 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET D. Incorrect, A ADS Logic has no backup. All ADS SRVs will shift to backup power 1D23-

14. Plausible if the operator believes that the ADS Valves do not have a backup power supply. All 4 ADS valves will function.

SD-183 ADS and LLS DC Power Supplies Technical Reference(s): (Attach if not previously provided) page 22 (rev 9)

Proposed References to be provided to applicants during examination: N 8.01.01.02 Given an ADS System operating mode and various plant conditions predict how the ADS Learning Objective: (As available) system will be impacted by failures in the following support systems:

d. 125 VDC buses 1D1 and 1D2 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 97 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 259002 K5.03 Importance Rating 3.1 259002 (SF2 RWLCS) Reactor Water Level Control: K5.03 - Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM: Water level measurement. (CFR: 41.5 / 45.3)

Proposed Question: RO Question 45 The plant is operating at 100% reactor power with Feedwater Level Control aligned in a normal configuration.

The Operator at the Controls observes the following:

How will the Feedwater Control System respond with no operator action?

A. RPV Water Level will remain at its setpoint B. Recirc MG sets will runback and RPV Water Level will return to its setpoint C. A Low RPV Water Level SCRAM will occur D. A High RPV Water Level SCRAM will occur Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 98 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed Answer: A Explanation: OI-644 P&L 14 Due to A/C GEMACS sharing a common variable leg, the B level control instrument is preferred. RX WATER LEVEL CONTROL INPUT SELECT HSS-4560 should normally be selected to B Level.

AOP 302.1 Loss of 1D11 with B Level Selected and no operator actions results in Feedwater Control opens Feed Reg Valves RPV level goes high B RFP and Main Turbine Trip on high RPV level (A RFP cannot be tripped remotely)

Reactor Scram (turbine trip)

A. Correct With FWLC in a normal lineup, B is the selected level. The failure of the A narrow range GEMAC will not affect the level control system. Level will remain at its previous setpoint.

B. Incorrect: At full power, both reactor feed pumps are in service therefore the 45%

runback would not occur with a low level instrument failure. Plausible if the operator believes that the runback would occur with the low level alarm in.

C. Incorrect RPV level will remain at the setpoint. The input from the A instrument will not affect FWLC. Plausible if the operator believes A is normally selected.and level would lower to 170 inches and a SCRAM would occur.

D. Incorrect RPV Level will remain at its setpoint. In addition, there is no High RPV level SCRAM.

This is plausible if the student believes that A is normally selected and RPV level would rise to the Main Turbine Trip setpoint indirectly causing a SCRAM..

OI-644 Rev 182 P&L 14 page 7 AOP-644 Rev 23 (page 3 step 5)

SD 644 Rev 5, Figure 5 Feedwater Control System Technical Reference(s): (Attach if not previously provided)

AOP-302.1 Loss of 125 VDC Rev 59 1D11 Auto Actions page 2 (1D11 was not lost) no action required Proposed References to be provided to applicants during examination: N 45.05.01.05 Describe the operation of Learning Objective: the feedwater control circuitry include: (As available)

a. RPV level control circuitry Question Source: Bank #

Modified Bank # (Note changes or attach parent)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 99 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 3 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 100 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 261000 K4.05 Importance Rating 2.6 261000 (SF9 SGTS) Standby Gas Treatment: K4.05 - Knowledge of STANDBY GAS TREATMENT SYSTEM design feature(s) and/or interlocks which provide for the following:

Fission product gas removal. (CFR: 41.7)

Proposed Question: RO Question 46 What component of the Standby Gas Treatment System provides for the removal of Fission Product Gases?

A. Prefilter B. HEPA Filter C. Roughing Filter D. Carbon Bed Filter Proposed Answer: D Explanation:

A. Incorrect: removes particulates B. Incorrect: Remove particulates greater than 0.3 micron in size C. Incorrect: removes moisture D. Correct The activated carbon iodine filter is a high efficiency deep bed type with a 6 inch layer of charcoal. Each train contains approx. 1224 pounds of potassium iodide impregnated activated charcoal for trapping elemental iodine and radioiodine in the form if organic compounds.

SD 170 Rev 14 (page 7)

Technical Reference(s): (Attach if not previously provided)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 101 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 7.04.01.06 describe the operation of the following principle SBGT system Learning Objective: (As available) components:

d. Charcoal Adsorber Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 102 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262001 K3.03 Importance Rating 2.9 262001 (SF6 AC) AC Electrical Distribution: K3.03 - Knowledge of the effect that a loss or malfunction of the A.C. ELECTRICAL DISTRIBUTION will have on following: D.C. electrical distribution. (CFR: 41.7 / 45.4)

Proposed Question: RO Question 47 A lockout occurs on 4160 VAC Bus 1A3, what is the expected response to the 125 VDC distribution system?

A. The AC supply to 1D12, Division 1 Battery Charger will automatically align to 1A4 B. The 125 VDC Swing Charger, 1D120, will automatically align to supply bus 1D10 C. The Division 1, 125 VDC system will be powered by station battery 1D1 D. The Division 1, 125 VDC system will be de-energized Proposed Answer: C Explanation: The battery chargers normally supply system power with the batteries providing supplemental power during periods of high demand. The batteries provide at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of complete backup power during battery charger failure. The 125 VDC system is arranged into two redundant and separate subsystems each consisting of a battery, battery charger, distribution apparatus and the necessary instrumentation, controls, and protective devices required to satisfy the system functional and design objectives. A third or swing charger is provided as a permanently connected spare for either of the normal battery chargers, This third charger can also be fed from either essential AC bus via interlock.

The third or swing battery charger (1D120) is connected via a pair of interlocked breakers to either 125 VDC panels. AC input to 1D120 is fed from 1B32x42 which interlocks AC input from either 1B3202 or 1B4210A. Only one battery charger for each of the 125 VDC panels is in service , while the third charger (1D120) is a spare for either panel. A mechanical interlock on this charger prevents it from supplying both 125 VDC panels.

A. Incorrect: There is no auto power swap for the station battery chargers Plausible if the student believes there is an feature that auto aligns the charger to the bus.

B. Incorrect: 1D120 the swing charger does not automatically realign. Plausible if the student believes there is an feature that auto aligns the charger to the bus.

C. Correct see explanation above D. Incorrect: The station batteries are sized to carry loads for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Plausible if the student believes that if AC power is lost, this will result in the loss of the distribution system.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 103 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET SD-375 Rev 10 (page 4, 8, 13)

Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 13.00.00.05 Evaluate plant conditions and Control room indications to determine if the 125 VDC power Learning Objective: system is functioning as expected and (As available) identify any actions that may be necessary to place the system in the correct condition Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 4 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 104 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262002 K1.19 Importance Rating 2.9 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC): K1.19 - Knowledge of the physical connections and/or cause-effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following: Power range neutron monitoring system: Plant-Specific. (CFR:

41.2 to 41.9 / 45.7 to 45.8)

Proposed Question: RO Question 48 What is the impact to the APRM/LPRM system if a loss of 1Y23, 120V Uninterruptible Power Supply occurs?

The ____(1)____ IRM/APRM Recorders will immediately ____(2)____.

A. (1) A and C (2) de-energize B. (1) A and C (2) indicate downscale C. (1) B and D (2) de-energize D. (1) B and D (2) indicate downscale Proposed Answer: C Explanation: AOP 357 probable indications: NMR-09254B/D screen goes dark AOP-317 Loss of Inst AC Probable indications NMR-9254A/C oss or failure A. Incorrect: Inst AC power not uninterruptible NMR-9254B/D screens go blank B. Incorrect: Inst AC power not uninterruptible NMR-9254B/D screens go blank C. Correct: AOP-357 NMR-9254B/D screens go blank D. Incorrect NMR-9254B/D screens go blank Technical Reference(s): AOP-357(rev 50) (Attach if not previously provided)

AOP-317 Rev 109 (page 10 )

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 105 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 81.01.01.06 Given an APRM system operating mode and various plant conditions, predict how the APRM Learning Objective: system will be impacted by the (As available) operation or failure of the following support systems or components:

g. Instrument AC Power Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2-7 55.45 7-8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 106 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 263000 K.2.01 Importance Rating 3.1 263000 (SF6 DC) DC Electrical Distribution: K2.01 - Knowledge of electrical power supplies to the following: Major D.C. loads. (CFR: 41.7)

Proposed Question: RO Question 49 With the plant operating at 100% reactor power and all systems in their normal lineup, what would be the effect of a loss of 1D40, 250VDC Distribution Panel?

A. HPCI cannot operate as designed B. RCIC cannot operate as designed C. Reactor Recirc MG Sets will trip D. The Main Turbine will trip Proposed Answer: A Explanation: HPCI valves and pumps are supplied by 250 VDC power (exception MO2238 inboard steam supply). Without 250 VDC the Aux Oil pump will not start to provide the initial oil pressure required to initiate HPCI, MO2312 normally closed discharge valve has no power to open and allow HPCI to inject.

A. Correct B. RCIC is supplied by 125 VDC power. Plausible if the operator confuses the RCIC and HPCI system electrical supplies.

C. Incorrect. Plausible since the RRMG Set Emergency Lube Oil Pumps will lose power however this will not cause a loss of lube oil pressure and subsequent RRMG set trip D. Incorrect. Although the Main Turbine Emergency Bearing Oil Pump is powered by 250 VDC, this component backs up the normal oil supply pumps. The loss of the EBOP will not cause a Turbine Trip. Plausible if the student believes the EBOP loss will result in a loss of lube oil pressure and a resultant TT will occur.

AOP-388 Rev 21 page 7) 1D41 Technical Reference(s): (Attach if not previously provided) load list Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 107 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 5.01.01.02 Given a HPCI system operating mode and various plant conditions predict how the HPCI Learning Objective: (As available) system will be impacted bu the following support system failures:

c. 250 VDC Distribution Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 108 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 264000 A4.05 Importance Rating 3.6 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG: A4.05 - Ability to manually operate and / or monitor in the control room: Transfer of emergency generator (with load) to grid. (CFR: 41.7 / 45.5 to 45.8)

Proposed Question: RO Question 50 The Standby Diesel Generator (SBDG) is running and is the only power supply to 1A3, 4160V Essential Bus.

The CRS directs transferring 1A3, 4160V Essential Bus from the A SBDG to the Startup Transformer.

The following indications are observed:

Which of the following is correct to parallel the Startup Transformer with the SBDG?

Should the Operator close the Startup Transformer to 1A3 supply breaker?

A. Yes; No additional adjustments are required and the Startup Transformer is ready to be paralleled to the SBDG Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 109 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET B. No; The Startup Transformer Supply Breaker to 1A3 cannot be closed due to a failure of the synch check relays lights C. No; The operator will have to lower the running voltage to parallel the Startup Transformer to the SBDG D. No; The operator will have to raise the incoming voltage to parallel the Startup Transformer to the SBDG Proposed Answer: C Explanation: OI304.2 it is required to have incoming voltage slightly higher than running voltage.

A. Incorrect: per OI it is required to have incoming voltage slightly higher than running.

Plausible if the candidate believes the given conditions are in accordance with the procedure.

B. Incorrect: normal indication with scope needle at 12 position. Plausible if the candidate believes the lights are illuminated with the sync scope needle at the 12 oclock position.

C. Correct: per OI it is required to have incoming voltage slightly higher than running D. Incorrect: per OI it is required to have incoming voltage slightly higher than running.

Plausible if the candidate believes paralleling is similar to bringing on the SBDG with the Startup Transformer already carrying the load.

Technical Reference(s): OI-304.2 rev 101 page section 7.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 19.01.01.03 Describe the operation of the following SBDG components and Learning Objective: (As available) controls:

g. governor system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 110 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 5-8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 111 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 300000 K4.03 Importance Rating 2.8 300000 (SF8 IA) Instrument Air: K4.03 - Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following: Securing of IAS upon loss of cooling water. (CFR: 41.7)

Proposed Question: RO Question 51 The plant is operating at 100% power with the following plant conditions:

• 1K-1, Backup Instrument Air Compressor, power is aligned to 1B45

• 1K-1 is the only available Air Compressor operating to support plant air needs

• A and C Well Water Pumps are operating A Loss Of Offsite Power occurs and the following conditions are now present:

• Both SBDGs start and energize their respective busses

• The A SBDG fails after several minutes of operation With no operator action, 1K-1 will _____________.

A. trip on high first stage air outlet temperature B. continue to run with cooling supplied by GSW C. trip due to the loss of motor cooling D. continue to run with cooling supplied by Well Water Proposed Answer: A Explanation: 1B33 supplies power to 1P58A and C Well pumps which are lost on 1B3 loss GSW is the backup cooling to 1K1 and requires manual re-alignment Motor is air cooled A. Correct, loss of well water cooling will cause a high air outlet trip B. Incorrect: On loss of 1B3 MCC 1B33 will be lost thereby causing the loss of 1P-58A and C well water pumps. GSW is the backup cooling water, but requires manual re-alignment. Plausible if the student believes that GSW auto aligns as the cooling source to the compressor.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 112 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Incorrect: 1K1 motor is air cooled. Plausible if the student believes the compressor motor is water cooled.

D. Incorrect: On loss of 1B3 MCC 1B33 will be lost thereby causing the loss of 1P-58A and C well water pumps. GSW is the backup cooling water, but requires manual re-alignment. Plausible if the student believes GSW pumps will auto start when the SBDG is supplying the essential loads. Well water flow is lost due to loss of 1B33 when 1A3 is lost.

SD-518 Section 7.9 Shifting 1K-1 Cooling Water from Well Water to GSW (page 58) Rev 103 Technical Reference(s): (Attach if not previously provided)

AOP-301 rev 75 (page 47 1B33 loads 1P58A/C).

SD-518 rev 9, page 13 Proposed References to be provided to applicants during examination: N 36.00.00.05 evaluate plant conditions and control room indications to determine if the instrument and Learning Objective: service air system is operating as (As available) expected and identify any actions that may be necessary to place the system in the correct condition Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 113 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 300000 G2.1.28 Importance Rating 4.1 300000 (SF8 IA) Instrument Air: A3.02 - G2.1.28 - Knowledge of the purpose and function of major system components and controls. (CFR: 41.7)

Proposed Question: RO Question 52 Which of the following is a plant function directly supported by the CB/SBGT Instrument Air Compressors?

A. Breathing air for the control room B. Outboard Main Steam Line Isolation Valves C. Secondary Containment Isolation Dampers D. Primary Containment Isolation Damper/Valve T-Seals Proposed Answer: D Explanation: 1K-3/4 support primary containment vent and purge valve T-seals.

SD 959 Primary Containment Control valves are equipped with inflatable T-ring seals that provide a leak-tight seating surface for butterfly discs on the valves.

SD-573 Containment Purge and Vent Valves the 18 inch containment purge and vent valves are pneumatically operated, butterfly dampers with a fail close actuator. An isolation signal causes the air supply solenoid to de-energize which vents air from the actuator and allows actuator spring force to close the valve. As the valve closes a mechanical arm will trip a limit switch, which pressurizes T-ring seals. These T-ring seals provide a seal around the valve disc to prevent any leakage from the valve disc and seat area.

A. Incorrect: 1K-3/4 do not supply breathing air. This is via the main plant air system.

Plausible if the student believes that breathing air is supplied by these compressors.

B. Incorrect: N2 is the motive force for the outboard MSIVs. Plausible if the student believes Outboard MSIVs are supplied by air. Many similar design BWRs have this alignment.

C. Incorrect: Main plant air supplies secondary containment isolation dampers. Plausible if the student mistakenly identifies the CB/SBGT Instrument Air compressors as the motive air to these dampers. They actually provide air for the Secondary Containment Dampers.

D. Correct Technical Reference(s): SD-573 rev 15 page 9 (Attach if not previously provided)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 114 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET OI-730 rev 126 table 1 1K-3 and 1K-4 Technical Specification Related loads (page 65).

SD-959 rev 6 page 10 TS Bases 3.6.1.3 A.1 and A.2 page 3.6-19.

TS Bases 3.7.9 Background page B3.7-40 (TSCR-044A)

Proposed References to be provided to applicants during examination: N 7.00.00.02 given a SBGT system operating mode and various plant conditions predict how the SBGT Learning Objective: (As available) system will be impacted by failures in the following support systems:

b. Instrument and Service air Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 115 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 400000 A3.01 Importance Rating 3.0 400000 (SF8 CCS) Component Cooling Water: A3.01 - Ability to monitor automatic operations of the CCWS including: Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS (CFR: 41.7 / 45.7)

Proposed Question: RO Question 53 A tube to shell leak in the Reactor Building Equipment Drain Sump Heat Exchanger system occurs. How will the level in the RBCCW surge tank respond to this condition?

RBCCW surge tank level will ________________.

A. lower; automatic makeup will restore level in the surge tank B. raise; the surge tank will overflow to the floor drain system C. lower; manual operator action is required to refill the surge tank D. raise; an automatic vent will OPEN to vent the surge tank to atmosphere.

Proposed Answer: C 1T-78 has an automatic makeup system however it is isolated to allow quantification of leakage into or out of the tank. Manual Operator Action is required to OPEN an isolation valve and fill the surge tank. The RBCCW surge tank overflows to open radwaste in the case of high water level. Any leakage from components at reactor pressure will be into the RBCCW system. The Equipment drain sump is at a lower pressure than RBCCW system pressure and therefore leakage from that component will be out of the RBCCW system. Level will lower.

Explanation:

A. Incorrect: Level will lower however automatic makeup to the surge tank is normally isolated. Manual Operator action is required to makeup to the tank. Plausible if the candidate believes that the auto makeup is aligned to FILL the surge tank upon the low level signal. This is not true.

B. Incorrect: The Equipment drain sump is at a lower pressure than RBCCW system pressure and therefore leakage from that component will be out of the RBCCW system.

Level will lower. Plausible if the candidate believes the equipment sump HX is at a higher pressure than the RBCCW supply pressure (Like NRHX). In this case however, it is not. Surge tank level would lower not rise.

C. Correct: The Equipment drain sump is at a lower pressure than RBCCW system pressure and therefore leakage from that component will be out of the RBCCW system.

Level will lower. Manual Operator Action is required to OPEN an isolation valve and fill the surge tank.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 116 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET D. Incorrect, level will lower and the tank is vented to Open Rad Waste. There is no automatic vent for the surge tank. Plausible if the candidate believes the equipment sump HX is at a higher pressure than the RBCCW supply pressure (Like NRHX). In this case however, it is not. Surge tank level would lower not rise. In addition, there is no auto vent. The tank is vented to atmosphere.

SD 414 rev 9 Technical Reference(s): (Attach if not previously provided)

OI-414 rev 42, P&L 1 page 3 Proposed References to be provided to applicants during examination: N 29.03.01.02 for any given RBCCW system operation or failure, describe the impact of that operation or failure Learning Objective: (As available) on the following systems or components:

g. RWCU Non-regenerative HX Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 117 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 201001 K3.03 Importance Rating 3.1 201001 (SF1 CRDH) CRD Hydraulic: K3.03 - Knowledge of the effect that a loss or malfunction of the CONTROL ROD DRIVE HYDRAULIC SYSTEM will have on following: Control rod drive mechanisms. (CFR: 41.7 / 45.4)

Proposed Question: RO Question 54 The plant is operating at 100% reactor power when a loss of air to the in service CRD Flow Control Valve occurs.

Which one of the following describes the impact on the CRD System?

A. Control Rods may drift B. Control Rod drive temperatures will remain unchanged C. CRD Accumulators would begin to discharge D. CRD Mechanism seal degradation would accelerate Proposed Answer: D Explanation: Flow Control Valves CV-1821 and 1822 fail closed (AOP-518) With CV-1821 /

1822 closed cooling flow to the CRD drives is decreased and drive temperatures rise. 1C05A (E-6) CRD Drive Mechanism HI Temp alarms at 250°F. Although the drives can function without cooling water, seal and bushing life is shortened by long term exposure to reactor temperatures.

A. Incorrect - the flow control valve fails closed on a loss of air reducing the pressure on the underside of the CRDM. This is plausible in the if the valve failed open it would increase the pressure on the underside of the CRDM piston resulting in the potential for rods drifting in.

B. Incorrect - The flow control valve fails shut on loss of air. Cooling water would stop to the mechanisms. Plausible if the student believes the FCVs fail as is and there is no change in cooling.

C. Incorrect - The charging line is upstream of the FCV. Plausible if the candidate believes that the charging water is downstream and that the system would not maintain the HCU charged.

D. Correct - the FCV fails shut on loss of air. Cooling water flow to the drives is reduced causing increased CRDM temperatures. Continued operation with elevated CRDM temperatures could result in CRDM seal degradation.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 118 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET AOP-518 Rev 42 page 2 Technical Reference(s): (Automatic Actions) (Attach if not previously provided)

SD 255 Rev 9 (page 15, & 23)

Proposed References to be provided to applicants during examination: N 10.01.01.05 given a CRDH system operating mode and various plant conditions, predict how the CRDH Learning Objective: (As available) system will be impacted by the failure of the following support systems:

f. Instrument and service air Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 4 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 119 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 204000 K1.11 Importance Rating 3.5 204000 (SF2 RWCU) Reactor Water Cleanup: K1.11 - Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER CLEANUP SYSTEM and the following: PCIS/NSSSS. (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Proposed Question: RO Question 55 The PCIS isolation demonstrated by the CIMS panel affected which system?

A. TIP Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 120 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET B. RHR C. RWCU D. Main Steam Proposed Answer: C Explanation:SD 959.1 Primary Containment Isolation System Group 5 affects the RWCU system.

A. Incorrect this would be Group 2 B. Incorrect, this would be Group 4 C. Correct this is affected by the Group 5 isolation D. Incorrect this would be Group 1 SD 959.1 Rev 13 page 10 Table 1 PCIS Isolation Signals Technical Reference(s): (Attach if not previously provided)

PCIS Status Board picture in SD 959.1 page 35 (rev 13)

Proposed References to be provided to applicants during examination: N 50007.05.05 Describe the indications Learning Objective: and meaning of the signals provided (As available) by the CIMS panel Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2-9 55.45 7-8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 121 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 219000 K2.02 Importance Rating 3.1*

RHR/LPCI: Torus/Suppression Pool Cooling Mode: Knowledge of electrical power supplies to the following: Pumps. (CFR: 41.7)

Proposed Question: RO Question 56 The RHR System is in torus cooling mode with A RHR Pump and B RHR Pump operating when the following is observed at 1C08:

What is the status of the A RHR and B RHR pumps based upon the following indications?

A. A RHR pump is operating B RHR pump is operating B. A RHR pump is operating B RHR pump is NOT operating Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 122 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. A RHR pump is NOT operating B RHR pump is operating D. A RHR pump is NOT operating B RHR pump is NOT operating Proposed Answer: C Explanation: Bus 1A3 Lockout Trip Automatic Actions: Breakers !a301 and 1A302 trip open and are interlocked from manually or automatically closing AND Bus 1A3 Load Sheds; A SBDG 1G31 auto starts on Bus 1A3 undervoltage, then runs up to speed and frequency. Breaker 1A311 does not auto close and cannot be manually closed.

AOP 301 Load Shedding of the following loads: 1P-229A and C RHR Pumps.

A. Incorrect: 1A3 lockout will cause a loss of voltage to 1A3, prevent 1A311 SBDG output breaker from closing onto the bus. 1P229A A RHR pump will not be running B. Incorrect: 1P229B would still be running 1A4 still has power and no trip conditions are given.

1A3 lockout will cause a loss of voltage to 1A3, prevent 1A311 SBDG output breaker from closing onto the bus. 1P229A A RHR pump will not be running C. Correct D. Incorrect: 1P229B would be running, no trip signal is present for it.

1A3 lockout will cause a loss of voltage to 1A3, prevent 1A311 SBDG output breaker from closing onto the bus. 1P229A A RHR pump will not be running ARP 1C08A (A-5) rev 100 page 18 AOP 301 Loss of 1A3 rev 75 Technical Reference(s): (Attach if not previously provided)

(page 2) Load Shedding information Proposed References to be provided to applicants during examination: N 2.01.01.06 given an RHR system operating mode and various plant conditions, predict how the RHR system will be impacted by operation, Learning Objective: (As available) or failure of the following support systems:

a. Essential 4160/480 VAC electrical power supplies.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 123 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 124 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 233000 G2.4.18 Importance Rating 3.3 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup: Generic K/A 2.4.18 - Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13)

Proposed Question: RO Question 57 What is the basis for the EOP entry condition for Spent Fuel Pool temperature?

The EOP entry condition for Spent Fuel Pool temperature is based upon the ______.

A. maximum Keff of the Spent Fuel in the Spent Fuel Pool B. capabilities of the Spent Fuel Pool cooling System piping C. environmental qualification of the Spent Fuel Pool Level Indicators D. Net Positive Suction Head limits for the Spent Fuel Pool Cooling pumps Proposed Answer: B Explanation: According to the EOP 3 Bases, the limiting design factor of the FPCCU System is the piping designed to accommodate 150°F water at 200psig. Spent fuel pool design temperature is 150°F The spent fuel pool design temperature is based on the capability of the FPC system piping, which bounds the temperature limitations of the pool liner A. Incorrect: Keff approaches the maximum value only when temperature lowers.

Moderation lessens with higher temperature resulting is less opportunity for inadvertent criticality. Plausible with the concept of moderator coefficient of reactivity.

B. Correct: According to the bases for EOP 3, the design limit temperature for the Spent Fuel Pool is the design limit of 150°F Actions will be taken by plant personnel to increase all available cooling to prevent exceeding this value. This is the basis for the entry condition to the EOP.

C. Incorrect: There is no mention of environmental qualification limitations of the Fuel Pool Level indicators as it relates to High Pool Temperature. Plausible in that the student may think that since other level instrumentation is affected by temperature and may invalidate readings this may be the basis for the High Temperature entry condition.

D. Incorrect: SFPC pumps trip on low skimmer surge tank level. Per EOP 3 the minimum safe operating spent fuel pool level is 25.17 ft (10 ft above the top of the fuel racks) well below the SST low level trips.

EOP-3 bases rev 13, (pages 24-Technical Reference(s): (Attach if not previously provided) 25)

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 125 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET UFSAR 9.1 Proposed References to be provided to applicants during examination: N 95.00.00.15 Explain the Bases of each Learning Objective: (As available) of the EOP Curves and Limits Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43/45 1/13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 126 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 239001 A4.11 Importance Rating 3.1 239001 (SF3, SF4 MRSS) Main and Reheat Steam: A4.11 - Ability to manually operate and/or monitor in the control room: Alternate methods of verifying valve positions. (CFR: 41.7 / 45.5 to 45.8)

Proposed Question: RO Question 58 The plant was operating at power with the Startup Transformer removed from service.

• The crew inserted a manual reactor scram

• The main turbine has tripped

• Reactor pressure is 940 psig and lowering While performing IPOI 5 actions the Operator observes:

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 127 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET What, if any, actions should the Operator perform in accordance with IPOI 5, Scram?

A. Close MO-1054 and MO-1055 B. Close MO-1362A and MO-1362B C. Close MO-9147 and MO-9148 D. No action is required for the observed condition Proposed Answer: A Explanation: Turbine Trip with the Startup Transformer out of service will cause a loss on nonessential electrical power.

MO9147 is powered by 1B1234 (non-essential)

MO9148 is powered by 1B2224 (non-essential)

MO-1362A is powered by 1B3705 (essential power)

MO1362B is powered by 1B3706 (essential power)

MO1054 is powered by 1B3708 (essential power)

MO-1055 is powered by 1B3707 (essential power)

IPOI-5 QRC step 9 is monitor RPV pressure, maintain RPV pressure below 1110 psig.

• Verify close MO-9147 / 9148 Close MO-1054 / 1055 if necessary IPOI-5 If steam flow is still indicated on either FI-1080A or FI-1081A, then close MO-1054 /

1055 MN STM to MSR Second Stage A. Correct: as directed by IPOI-5 and QRC. Flow is still indicated on FIs after scram. On loss of non-essential power caused by the turbine trip MO-9147 and 9148 will have lost power. It is necessary to close MO-1054 and 1055 B. Incorrect: Plausible if the student believes, with the loss of power additional steam loads should be removed from service to prevent excessive cooldown. Closing the SJAE steam supplies secures a large steam draw and conserves steam header pressure.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 128 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Incorrect: Loss of non-essential power will prevent these valves from closing or being closed. Plausible if the student believes operators should be sent to the field to CLOSE these valves. This is not directed to be done.

D. Incorrect: As seen by picture steam flow is present requiring MO-1054 and 1055 to be closed. Plausible if the student does not recognize the direction from IPOI 5 to close 1054 and 1055 with the conditions provided.

IPOI-5 rev 62 (page 6 of 15)

Technical Reference(s): (Attach if not previously provided)

IPOI-5 QRC rev 12 (page 1 of 1)

Proposed References to be provided to applicants during examination: N 93.00.00.14 Contrast the different methods of cooling down the reactor Learning Objective: (As available) when the Main Condenser is and is not available Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 5-8 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 129 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 241000 K5.04 Importance Rating 3.3 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating: K5.04 - Knowledge of the operational implications of the following concepts as they apply to REACTOR/TURBINE PRESSURE REGULATING SYSTEM: Turbine inlet pressure vs. reactor pressure. (CFR: 41.5 /

45.3)

Proposed Question: RO Question 59 The plant is operating at 98% reactor power with the "A" EHC Pressure Regulator in service.

• The "A" Pressure Regulator malfunctions such that "Steam Throttle Pressure A" is slowly failing DOWNSCALE Assuming NO operator action is taken, which one of the following describes the effect on reactor pressure?

Reactor Pressure will __________.

A. slowly rise until the reactor scrams on either high flux or high pressure B. stabilize a few psig lower controlled by the "B" EHC Pressure Regulator C. slowly lower resulting in a reactor scram on an automatic MSIV closure D. stabilize a few psig higher controlled by the "B" EHC Pressure Regulator Proposed Answer: D Explanation: As the A steam throttle pressure falls, the A side pressure error signal will go down. This will cause the CVs to begin to close. As the CVs close, reactor pressure (and hence throttle pressure) will begin to rise. This will be seen by the B regulator. So as the A side pressure error signal goes down, the B pressure error will go up. This will eventually cause the B regulator to take over at a slightly elevated reactor and throttle pressure A. Incorrect: As the A Side pressure error signal goes down, the B pressure error signal will go up. This will eventually cause the B regulator to take over and control pressure slightly higher. No scram will occur B. Incorrect: Due to the -5 psig bias the B regulator will take over at a slightly elevated reactor and throttle pressure Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 130 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Incorrect: As the A Side pressure error signal goes down, the B pressure error signal will go up. This will eventually cause the B regulator to take over and control pressure slightly higher. No scram will occur This would occur if the A regulator failed high D. Correct: As the A steam throttle pressure falls, the A side pressure error signal will go down. This will cause the CVs to begin to close. As the CVs close, reactor pressure (and hence throttle pressure) will begin to rise. This will be seen by the B regulator. So as the A side pressure error signal goes down, the B pressure error will go up. This will eventually cause the B regulator to take over at a slightly elevated reactor and throttle pressure AOP-262 rev 10 Automatic Actions page 2 Technical Reference(s): SD-693.2A EHC Logic rev 7 (61) (Attach if not previously provided)

Figure 17 EHC Logic Control System page Proposed References to be provided to applicants during examination: N 52.01.01.02 Given an EHC system operating mode and various plant conditions, predict how the EHC Learning Objective: (As available) system will be impacted by failures in the following support systems:

a. Logic Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 3 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 131 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 132 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 245000 A3.05 Importance Rating 3.0 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary: A3.05 - Ability to monitor automatic operations of the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS including:

Control valve operation. (CFR: 41.7 / 45.7)

Proposed Question: RO Question 60 The plant is operating at 20% power when a grid disturbance occurs and the following conditions are present:

• The Main Generator TRIPS WITHOUT a Main Turbine TRIP

• The Main Turbine speed rises to 1910 RPM What is the status of the Control and Intercept Valves at this speed?

Control Valves Intercept Valves A. OPEN CLOSED B. CLOSING CLOSING C. CLOSED CLOSING D. CLOSED CLOSED Proposed Answer: C Explanation: Based on either the EHC Logic Control System diagram on page 61 of SD-693.2 Rev or the Control Valve Intercept Valve Sequencing diagram on page 33 of SD-693.2 show the sequence of the turbine valves. From 1800 RPM to 1890 RPM the Turbine Control Valves ramp close, then at greater than 1890 RPM the intercept valves begin to ramp close to protect the turbine from overspeed. The Intercept Valves go from Fully OPEN to Fully Closed between 1890 RPM and 1926 RPM. From the EHC logic control system diagram, in this condition the rest of the response is that the Bypass Valves will continue to maintain RPV pressure.

A. Incorrect: Control valves close from 100% to 105% speed (1800-1890 RPM). The Intercept Valves would not be fully Closed. Plausible if the student doesnt understand the sequence of valve operation to protect the turbine from overspeed.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 133 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET B. Incorrect Control valves close from 100% to 105% speed (1800-1890 RPM). Plausible if the student doesnt understand the sequence of valve operation to protect the turbine from overspeed.

C. Correct D. Incorrect: IVs will ramp closed from 105% to 107% (1890 to 1926 RPM). Plausible if the student doesnt understand the sequence of valve operation to protect the turbine from overspeed.

SD 693.2A rev 7 Overspeed and Technical Reference(s): overpressure conditions (page 32- (Attach if not previously provided)

33) explanation and graph Proposed References to be provided to applicants during examination: N 51.00.00.04 Describe the operation of the following principle main turbine system components:

Learning Objective: (As available)

d. Main stop valves

e. Control valves

f. Combined Intermediate valves Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 134 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 256000 A2.14 Importance Rating 3.3 256000 (SF2 CDS) Condensate: A2.14 - Ability to (a) predict the impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low Condensate storage tank level. (CFR: 41.5 / 45.6)

Proposed Question: RO Question 61 The plant is operating at 95% reactor power when the CST level lowered to the low level alarm set point.

(1) What is the operational impact at this level in the CST AND (2) How is the CST level restored?

A. (1) 1P-12A and B Condensate Service Water Pumps and 1P-11 Condensate Service Water Jockey Pump TRIP (2) Transfer Demin Water to the CSTs B. (1) 1P 12-A and B Condensate Service Water Pumps and 1P-209 A and B CRD Pumps TRIP (2) Align Core Spray to the CST C. (1) 1P-12A and B Condensate Service Water Pumps and 1P-11 Condensate Service Water Jockey Pump TRIP (2) Align Core Spray to the CST D. (1) 1P 12-A and B Condensate Service Water Pumps and 1P-209 A and B CRD Pumps TRIP (2) Transfer Demin Water to the CST Proposed Answer: A Explanation:Low CST level at 6Ft will cause an automatic trip of 1P-12A and B the Condensate Service Water Pumps, and 1P-11 Condensate Service Water Jockey Pump 1P-209A/B do not trip on low CST level Align to the CSTs is incorrect, this is used to fill the Torus. Torus height is not sufficient to fill the CSTs Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 135 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Demin water storage tank 1T45 is the normal makeup source to the CSTs via a Makeup Demin trailer and 1P13A/B demin water transfer pumps.

A. Correct B. Incorrect, Aligning core spray is incorrect. This is used to fill the Torus not the CST.

The student may incorrectly assume Torus height is sufficient to effect the same method of filling the Torus from the CST C. Incorrect D. Incorrect. The CRD pumps do not TRIP from a low level in the CSTs, the suction head to the CRD pumps is provided by the reject line from Condensate Discharge. This was not provided to the student as a condition in the stem.

ARP 1C06A (B-8 and B-9) rev 81 Automatic actions page 48 & 50 Technical Reference(s): OI-537 Condensate/Demin (Attach if not previously provided)

Service Water Section 6.0 Filling the CSTs rev 54 (page 16)

Proposed References to be provided to applicants during examination: N 45.01.01.01 relate the precautions and limitations, operating cautions, or procedural notes of OI-639, and any Learning Objective: (As available) applicable ARPs to any component or feed and condensate system operating status.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 136 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 259001 A1.06 Importance Rating 2.7 259001 (SF2 FWS) Feedwater: A1.06 - Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: Feedwater heater level. (CFR: 41.5 / 45.5)

Proposed Question: RO Question 62 The plant is operating at 100% power when annunciator 1C06B (A-9) LP FDWTR HTR 1E-5A HI-HI LEVEL is received.

How will the feedwater heater level control system respond to this condition?

A. 1E-5A Feedwater Heater Drain Valve will OPEN and the 1E-5A Feedwater Heater Dump Valve will CLOSE B. 1E-5A Feedwater Heater Drain Valve will CLOSE and the 1E-5A Feedwater Heater Dump Valve will OPEN C. 1E-6A Feedwater Heater Drain Valve will CLOSE and the 1E-5A Feedwater Heater Dump Valve will OPEN D. 1E-6A Feedwater Heater Drain Valve will CLOSE and the 1E-5A Feedwater Heater Dump Valve will CLOSE Proposed Answer: C Explanation: On the 1E-5A feedwater heater hi-hi level the following occurs:

• The drain valve for the 6A feedwater heater fails closed

• The dump valve for the 5A feedwater heater fails open The drain valve for the 5A feedwater heater will be open A. Incorrect: The first part is correct. The 1E-5A feedwater heater dump valve will fail open. Plausible for a feedwater heater high level alarm.

B. Incorrect: The 1E-5A feedwater heater drain valve will be open due to the rising heater level. The second part is correct. Plausible if the candidate believes the hi-hi level is to be corrected by the dump valve alone to prevent downstream feedwater heater level perturbations.

C. Correct:

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 137 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET D. Incorrect: The first part is correct. The 1E-5A feedwater heater dump valve will fail open. Plausible if the candidate believes the 1E-6A drain valve closing will correct the 1E-5A feedwater hi-hi level alone.

AOP-646 Loss of Feedwater heating rev 25 Technical Reference(s): (Attach if not previously provided) 1C06B (A-9) LP FDWTR HTR 1E-5A HI-HI LEVEL Rev 66 Proposed References to be provided to applicants during examination: N 46.00.00.07 evaluate plant conditions and control room indications to determine if the extraction steam and feedwater heating system is operating Learning Objective: as expected, and identify any actions (As available) that may be necessary to place the extraction steam and feedwater heating system or the plant in the correct condition.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 5 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 138 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 271000 K6.11 Importance Rating 3.2 271000 (SF9 OG) Offgas: K6.11 - Knowledge of the effect that a loss or malfunction of the following will have on the OFFGAS SYSTEM: Condenser vacuum.

(CFR: 41.7 / 45.7)

Proposed Question: RO Question 63 How will excessive Main Condenser air inleakage affect operation of the Offgas System?

Offgas process flowrate will rise and ___________.

A. result in recombiner temperatures exceeding 875°F high temperature limit B. CV-4108, Offgas Outlet Isolation, will isolate C. CV-4151, Offgas Steam Jet Air Outlet, will isolate D. Offgas Loop Seals will isolate Proposed Answer: D Explanation: Offgas isolation signals are

>/= 4 psig at inlet to SJAE hold up line isolates all loop seal drains (CV-1379, 4126, 4179, 4106, 4107A/B)

.>/= 4.5 psig at inlet to 30 minute holdup line isolates all loop seal drains

</= 4000 lbm/hr motive steam flow to offgas jet compressor AND ,/= 270 psig steam supply pressure to the offgas jet compressor closes MO-4151 offgas jet compressor inlet valve (MO position interlock closes CV-1379 and CV-4126 and isolate HWC.

11-27-03 Plant OE: During a reactor startup a Condenser butt weld failed resulting in excessive in-leakage and high off gas flows and pressures. Offgas Loop Seals isolated. Offgas flow rose to above 300 scfm requiring charcoal adsorber beds to be bypassed.

A. Incorrect: Offgas recombiner temperatures would lower not rise in this case. Student may apply assume recombination would increase in this case raising recombiner temperatures.

B. Incorrect: CV-4108 Offgas Outlet Isolation no longer has automatic isolation signals.To ensure scram frequency reduction, CV-4108 fails open on a loss of instrument air or electrical power to the valve. A valve position interlock from CV-4108 causes all loop seals to isolate when CV-4108 closes. Plausible if the student believes the offgas outlet isolation closes on an isolation signal. This is not the case.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 139 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Incorrect: CV-4151 does not have an isolation however MO-4151 Offgas Jet Compressor inlet valve does CLOSE on an isolation signal. Plausible if the student believes the offgas outlet isolation closes on an isolation signal. This is not the case.

D. Correct: Excessive air inleakage would cause system pressure to rise and isolate the loop seals Technical Reference(s): ARP 1C34 (C-5) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 47.01.01.14 evaluate plant conditions and control room indications to determine if the Offgas and recombiner system is operating as Learning Objective: (As available) expected, and identify any actions that may be necessary to place the Offgas and recombiner system in the correct lineup.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 140 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 286000 K4.06 Importance Rating 3.4 286000 (SF8 FPS) Fire Protection: K4.06 - Knowledge of FIRE PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Fire suppression capability that does not rely on the displacement of oxygen (Halon): Plant-Specific. (CFR: 41.7 / 45.7)

Proposed Question: RO Question 64 The Jockey Fire Pump, 1P-47, fails and fire main header pressure is lowering slowly. Which one of the following is the first expected automatic response of the fire protection system?

The ____(1)____ Fire Pump will initiate at ___(2)___ psig.

A. (1) 1P-49, Diesel (2) 85 B. (1) 1P-49, Diesel (2) 95 C. (1) 1P-48, Electric (2) 85 D. (1) 1P-48, Electric (2) 95 Proposed Answer: D Explanation: Jockey fire pump normally maintains fire header pressure and cycles between psig120 psig and 130 psig 1P48 the electric fire pump starts at a system pressure of 95 psig 1P49 the diesel fire pump starts at a system pressure of 85 psig A. Incorrect : although this is the correct pressure for the diesel fire pump auto start the electric would start at 95 psig and prevent system pressure from lowering to 85 B. Incorrect: The electric fire pump 1P48 starts at 95 psig C. Incorrect: the electric fire pump starts at 95 psig. Listed pressure is for the 1P49 diesel fire pump auto start D. Correct: 1P48 electric fire pump would start at 95 psig ARP 1C40 (rev 77) J-1 and J-5 Technical Reference(s): (Attach if not previously provided)

(page 120 and 130 of 133 and )

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 141 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET OI-513 Fire Protection rev 139 (page 21and 22 of 123 Note)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.45 7 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 142 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 288000 K3.04 Importance Rating 3.2 288000 (SF9 PVS) Plant Ventilation: K3.03 - Knowledge of the effect that a loss or malfunction of the PLANT VENTILATION SYSTEMS will have on following: Secondary containment pressure. (CFR: 41.5 / 45.3)

Proposed Question: RO Question 65 The plant is operating at 100% reactor power and receives a valid Group 3 signal.

During verification of the Group 3 isolation the following is observed at 1C23:

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 143 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET What, if any, could be the impact of this indication?

A. Unmonitored ground level release B. SBGT train inlet relief valve will open C. Reactor Building air temperature will lower D. No impact this is normal for a Group 3 Isolation Proposed Answer: A SD 733 Revision 8 page 36. Operator should determine that the picture represents a failure ot isolate secondary containment which could lead to an unmonitored ground level release.

A. Correct Failure of both isolation dampers in one line could lead to unmonitored release Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 144 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET B. Incorrect: SBGT inlet relief would lift due to venting Primary Containment with a high pressure present C. Incorrect: Group 3 isolation will cause reactor building supply and exhaust fans to trip and will cause building temperatures to rise.

D. Two dampers within the same penetration have failed to close, this is abnormal and not an expected result IPOI-7 Group 3 verification Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 50007.05.02 Explain the isolation signals with respect to setpoints, Learning Objective: (As available) components affected and the reason for each isolation signal.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 3 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 145 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.1.34 Importance Rating 2.7 Knowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12)

Proposed Question: RO Question 66 Why is the reactor coolant chloride limit higher in Mode 1 than in Modes 2 and 3 for a Boiling Water Reactor?

A. Higher zinc concentration B. Lower oxygen concentration C. Higher noble metal concentration D. Lower hydrogen concentration Proposed Answer: B Explanation: The higher coolant temperatures and de-aeration of the condensate in Mode 1 reduce the oxygen concentration and the effect of chloride is not as great with low oxygen Therefore the upper limit on chloride is allowed to be higher in Mode 1.

A. Incorrect. The Zinc Injection process converts the corrosion layer in the recirc piping by replacing CO-60, which in turn reduces Drywell dose from the recirculation piping.

Plausible if the student identifies this process as a means of reducing Chloride concentration in the Reactor Coolant System B. Correct: The higher coolant temperatures and de-aeration process of the condensate in Mode 1 reduce the oxygen concentration and the effect of chloride is not as great with low oxygen Therefore the upper limit on chloride is allowed to be higher in Mode 1 C. Incorrect: The Noble Metal Chemical Addition (NMCA) process is in place to further enhance the effectiveness of HWC and lowers H2 addition rates lowering plant dose.

Plausible if the student believes the NMCA process is the reason for higher Chloride limits during MODE 1.

D. Incorrect: H2 is injected to the coolant to raise the concentration and force O2 scavenging to prevent excessive oxygen concentration and IGSCC. Plausible if the student does not understand the addition of Hydrogen to prevent this condition. The H2 concentration is not why the coolant cl- limit is higher at higher powers.

SD-563 Hydrogen Water Chemistry Rev. 12 Technical Reference(s): (Attach if not previously provided)

SD-644 Feedwater System, Rev 16 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 146 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43/45 5/12 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 147 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.1.3 Importance Rating 3.7 Knowledge of shift or short-term relief turnover practices. (CFR: 41.10 / 45.13)

Proposed Question: RO Question 67 With the plant at full power, the following conditions exist:

• You are an NSOE, unassigned to the shift, working in the Work Control Center

• You last stood watch 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> ago One of the Control Room NSOEs is required to attend a briefing, and you have been asked to relieve him for about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(1) How far back are you required to read the Station Log prior to taking the watch AND (2) Is the NSOE and ANSOE Turnover Form, NG-016K, required to be used?

A. (1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2) The Turnover form must be used B. (1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2) The Turnover form is NOT required C. (1) 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> (2) The Turnover form must be used D. (1) 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> (2) The Turnover form is NOT required Proposed Answer: A Explanation:

Per Conduct of Operations OP-AA-100-1000 Rev 25. Attachment 8 Shift Relief and Turnover, the on-coming watchstander shall review the station log back to the last time the individual stood the watch or three days (whichever is less)

Per ACP 1410.10 Shift Turnover / Shift Brief, The applicable shift turnover form should be utilized for the watchstanders position being relieved..

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 148 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET A. Correct. 1st part correct, 2nd part correct. According to OP-AA-100-1000 Conduct of Operations, the on-coming watchstander shall review the Station Log back to the last time the individual stood the watch or three days (whichever is less). Since the operator last stood watch 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> ago, the requirement is to review back for ONLY 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

According to ACP 1410.10 (section 3.3 ) Watchstander Relief during the shift the applicable Shift Turnover Form should be used for the watchstanders position being relieved In the event that only one position (Shift Manager, CRS, STA, NSOE, or ANSOE) is relieved, then an N/A should be placed where appropriate on the shift turnover form for the non-relieved crew members.

B. Incorrect. 1st part correct, 2nd part wrong. This is plausible because the operator may incorrectly believe that the form does not need to be used because the relief is in the middle of a shift and of a temporary nature.

C. Incorrect. 1st part wrong, 2nd part correct. This is plausible because the operator last stood the watch 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> ago, and four hours have elapsed since the start of the work day. The operator may incorrectly believe that they are required to read the log back to the time that they last held the shift.

D. Incorrect. 1st part wrong, 2nd part wrong. This is plausible because the operator may incorrectly believe that they are required to read the log back to the time that they last held the shift; and because the operator may incorrectly believe that the form does not need to be used because the relief is in the middle of a shift and of a temporary nature.

OP-AA-100-1000 Conduct of Operations rev 26 Attachment 8 (pages 62) section 3.1 step 12 Technical Reference(s): (Attach if not previously provided)

ACP 1410.10 Shift Turnover/Shift Brief rev 49 Section 3.3 relief during the shift Proposed References to be provided to applicants during examination: N 96.05.05.03 Explain the requirements and instructions of ODI-009, Reactor Learning Objective: Operator, Senior Reactor Operator, (As available) and Shift Technical Advisor Qualification requirements Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: PDA 2009 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 149 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.45 13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 150 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.1.39 Importance Rating 3.6 Knowledge of conservative decision making practices. (CFR: 41.10 / 43.5 / 45.12)

Proposed Question: RO Question 68 Xenon is still changing while shift turnover is in progress.

The offgoing CRS directs the offgoing OATC to lower power by 1 MWe using Recirc How should this evolution take place?

A. The offgoing OATC should make the power change prior to completing the turnover; a peer check is required.

B. The offgoing OATC should make the power change prior to completing the turnover; a peer check is NOT required.

C. The oncoming OATC should make the power change prior to completing the turnover; peer check is required.

D. The oncoming OATC should make the power change prior to completing the turnover; a peer check is NOT required.

Proposed Answer: A Explanation: OP-AA-100-1000 When the plant or a particular watch station is operating in other than steady state conditions the offgoing and oncoming shift supervision shall determine which watch stations if any may relieve the watch and which watch stations shall not relieve until the plant has been placed in a stable condition. In cases when reactivity changes are directed, they are to be avoided at shift turnover.

A. Correct:

B. Incorrect: A peer check is required to be performed for reactivity manipulations.

Plausible if the student does not understand the expectations to perform a peer check for reactivity manipulations.

C. Incorrect: Shift turnover practices allow for turning over the shift to the oncoming operator if changes in the equipment are understood by the oncoming operator. In cases when reactivity changes are directed, they are to be avoided at shift turnover.

Plausible if the student does not fully understand the expectations for shift turnover of individual watchstations.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 151 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET D. Incorrect: Shift turnover practices allow for turning over the shift to the oncoming operator if changes in the equipment are understood by the oncoming operator. In cases when reactivity changes are directed, they are to be avoided at shift turnover.

Plausible if the student does not fully understand the expectations for shift turnover of individual watchstations. A peer check is required to be performed for reactivity manipulations. Plausible if the student does not understand the expectations to perform a peer check for reactivity manipulations.

Conduct Of Operations OP-AA-Technical Reference(s): (Attach if not previously provided) 100-1000 Proposed References to be provided to applicants during examination: N 96.07.01.20 Explain the requirements for equipment manipulations and Learning Objective: (As available) control of plant equipment for on-shift and/or off-shift operations personnel Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/12 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 152 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.2.14 Importance Rating 3.9 Knowledge of the process for controlling equipment configuration or status. (CFR: 41.10 / 43.3 /

45.13)

Proposed Question: RO Question 69 Which of the following is an approved method to control equipment configuration of the plant?

A. System Descriptions B. Operating Instruction Lineups C. Operations Equipment Database D. Archived clearance order restoration directions Proposed Answer: B Explanation: Operating Instruction Lineups are controlled documents and are used to maintain plant configuration control.

OP-AA-101-1000 rev 19 Clearance and tagging, Work complete and tag removal 4.8.3.C clearance removals shall be prepared using controlled references when available Preparing Clearances The clearance preparer shall prepare a clearance document using available references and/or walk-downs. Controlled references should be used, when available.

A. Incorrect: The system descriptions do contain simplified drawings and are controlled documents however they are not approved for configuration control B. Correct C. Incorrect: The Operations equipment database is uncontrolled and can only be used for information only.

D. Incorrect: Archived clearances do contain component identifications and as left configuration however they should not be used to control configuration of plant components OP-AA-101-1000 Rev 19 page 35 of 193 Technical Reference(s): (Attach if not previously provided)

OP-AA-100-1002 Plant Status Control Management Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 153 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 96.07.01.19 Describe the processes Learning Objective: by which configuration control is (As available) maintained at the station Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43/45 3/13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 154 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.2.13 Importance Rating 4.1 Knowledge of the Tagging and Clearance program. (CFR: 41.10 / 45.13)

Proposed Question: RO Question 70 Which of the following work tasks requires a Clearance Order?

A. Replacing a SBGT train Roughing Filter B. Changing fuses in a 250 VDC system C. Changing the Domestic Water Filter D. Liquid nitrogen tank delivery/refill Proposed Answer: A Explanation: OP-AA-101-100 Clearance and Tagging Rev. 24 Attachment 10 General Pre-Approved Work A. Correct: This task is on radiologically controlled equipment and the worker is exposed to rotating equipment. OP-AA-101-100 Clearance and Tagging Rev. 24 Attachment 10 General Pre-Approved Work does not allow work on components near rotating equipment. In addition, the system filters are potentially contaminated. A clearance would be required.

B. Incorrect: Changing fuses on systems with less than 600 volts does not require a clearance when using properly rated fuse pullers therefore no clearance is required.

C. Incorrect: This task is controlled by OI 528 Domestic Water section 6.2 Changing the disposable Pre-Filter 1F97, and no clearance is required. Valve positions are procedurally controlled and the system is low energy.

D. Incorrect: The N2 tank is vendor installed and maintained equipment. Liquid N2 receipt is controlled by OI 573 Attachment 6 Liquid N2 receipt. Filling the tank does not require a clearance.

OP-AA-101-1000 Clearance and Technical Reference(s): (Attach if not previously provided)

Tagging pg. 177 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 155 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43/45 2/13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 156 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.2.38 Importance Rating 3.6 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

Proposed Question: RO Question 71 Reactor Power is 1900 MWth and rising 1 MWth/minute. At what time will the facility license limit be reached first?

A. 5 minutes B. 11.5 minutes C. 12 minutes D. 15 minutes Proposed Answer: C Explanation: Technical Specification Renewed Facility Operating License, DPR-49, Maximum Power Level: Nextera Energy Duane Arnold, LLC is authorized to operate the Duane Arnold Energy Center at steady state reactor core power levels not in excess of 1912 megawatts (thermal). Amendment 306 At current rate of rise 1 MWth/minute the facility can operate 12 minutes prior to reaching its licensed thermal power limit.

IPOI-3 P&L #2 The licensed power limit of 1912 MWTh shall not be intentionally exceeded.

Operators shall take prompt action if rated thermal power (RTP) exceeds the limit of 1912 MWTh.

A. Incorrect Plausible if candidate believes 1905 limit as stated in IPOI-3 P&L 3(f)

Prior to the PPC being shutdown for scheduled maintenance that could last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or longer, reactor power should be lowered with recirc flow to 1905 mwth. This will allow sufficient margin for maintaining the 8 hr average (C177) for core thermal power below 1912 mwth while the heat balance is unavailable. Plausible if the student mistakes the administrative limit as the license thermal limit.

B. Incorrect IPOI-3 1911.5 is not the facility licensed limit.

IPOI-3 P&L 3(b) states if the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 1911.5 MWth as deter4mined by computer point NSS063, take prompt action to lower reactor power as necessary to prevent the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average from exceeding 1912 MWt. h Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 157 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET C. Correct 1912 MWth is the facility licensed limit D. Incorrect 1915 is above the facility licensed limit Tech Spec licensed thermal power Technical Reference(s): limit (Attach if not previously provided)

IPOI-3 P&L 2 (page 3) rev 160 Proposed References to be provided to applicants during examination: N 94.03.01.03 Relate how each step and Learning Objective: its performance meets the mitigation (As available) strategies of AOP 255.2 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7,10 55.43/45 1/13 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 158 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.3.7 Importance Rating 3.5 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(CFR: 41.12 / 45.10)

Proposed Question: RO Question 72 Who can suspend radiation work permit requirements during a station emergency when the Technical Support Center is not manned?

A. Operations Shift Manager B. Emergency Planning Manager C. On-shift Health Physics Technician D. On-shift Nuclear Station Operating Engineer (NSOE)

Proposed Answer: A Explanation: In the above example, with the TSC not yet manned. In the event of a declared plant emergency, RWP usage may be suspended. All work in Radiological Areas will be performed under the direction of the Site Radiation Protection Coordinator and the Emergency Response Organization. IAW EPIP 2.5 Control Room Emergency Response Operation, these positions are assumed by the OSM until the TSC is manned.

A. Correct: During a declared emergency, the Emergency Response Organization (Emergency Coordinator,) is the responsibility of the Operations Shift Manager, when the TSC is not manned. The OSM can suspend RWP requirements prior to responsibility turnover to the TSC.

B. Incorrect: The Emergency Planning Manager, who can be qualified as the Emergency Coordinator, can suspend the RWP requirements. Since, the TSC is not manned, this individual CANNOT suspend RWP requirements.

C. Incorrect: During normal operations, the On-shift Health Physics Technician, will direct which RWP requirements are necessary for given plant conditions.

D. Incorrect: During a declared emergency, the On-shift Nuclear Station Operating Engineer (NSOE) can provide the radiological brief to responding Operators but CANNOT suspend the RWP requirements.

HPP 3101.05, Admin of RWPs, Technical Reference(s): (Attach if not previously provided)

Rev. 61 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 159 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET EPIP 2.5 Control Room Emergency Response Operation, Rev 23 section 3 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.45 10 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 160 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.3.15 Importance Rating 2.9 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.12 / 43.4 / 45.9)

Proposed Question: RO Question 73 Which of the following is the minimum number of ARMs on the refuel floor which are required to meet the 10CFR70.24 requirements for criticality monitoring?

A. 1 B. 2 C. 3 D. 4 Proposed Answer: B Explanation: OI-879.2 Area radiation monitoring system.

While the system is in operation, each power supply provides power to 10 channels. At no time will any power supply be turned off except for servicing of that power supply.

To comply with the Code of Federal Regulations 10CFR70.24 part (a)(1), the ARMs on the refuel floor (RM9153, RM9163, RM9164, and RM9178) are to serve as criticality monitors with 2 of the 4 monitors remaining in service at all times. In the event that more than 2 of the monitors must be removed from service at one time, alternate monitoring with an audible alarm must be provided such that at least 2 monitors are always in service. The alarm setpoint of the alternate monitor must be less than or equal to the setpoint of the monitor it is replacing.

The correct answer cannot be determined by the system description. The answer is derived from the Code of Federal Regulation 10CFR70.24 part (a)(1).

A. Incorrect 4 refuel floor ARM detectors are provided, but only 2 are required B. Correct Two is the minimum required number of ARM detectors for refuel floor criticality monitoring C. Incorrect 4 refuel floor ARM detectors are provided, but only 2 are required D. Incorrect 4 refuel floor ARM detectors are provided, but only 2 are required Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 161 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET OI-879.2 Area Radiation Monitoring System rev 27 P&L 4 Technical Reference(s): (Attach if not previously provided)

(page 3 of 14)

AOP-317 rev 109 Note (page 5)

Proposed References to be provided to applicants during examination: N 86.04.01.01 Relate the precautions and limitations, operating cautions, or Learning Objective: procedural notes of OI-879.2 to any (As available) component or ARM system operating status Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 4 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 162 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.4.12 Importance Rating 4.0 Knowledge of general operating crew responsibilities during emergency operations. (CFR:

41.10 / 45.12)

Proposed Question: RO Question 74 During Emergency Operating Procedure (EOP) implementation, which one of the following describes the proper use of Abnormal Operating Procedures (AOP) during this time?

A. All AOPs are suspended until EOPs are exited B. The Only AOP actions that can be taken are those directed by EOPs C. AOP actions can be taken as long as they do not contradict EOP actions D. All AOP actions are required to be completed regardless of the EOP actions Proposed Answer: C Explanation: EOPs can be used in conjunction with other operating procedures (OIs, ARPs, AOPs, etc). However, EOPs are higher tier documents and shall direct the primary response to operational transients that require their use. The decision to utilize other approved procedures during EOP execution rests with the Shift Supervisor / Manager. It other procedures are used while executing EOPs, actions specified in these procedures shall not contradict or subvert actions described in EOPs or degrade the operability of equipment critical to EOP strategies.

A. Incorrect AOPs Can be used in conjunction with EOPs provided they do not conflict with the EOPs as a higher tier document. Since EOPs are higher tier the student may think they are not to be performed concurrently B. Incorrect Any AOPs can be used in conjunction with the EOPs as long as they do not conflict with the EOPs. Some AOPs are not directed by the EOPs and should be performed.

C. Correct D. Incorrect As stated above only those AOP actions that do not conflict with the EOPs should be performed as plant conditions require.

Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 163 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET ACP 1410.1 Operations Working Technical Reference(s): Standards rev 109 (step 7 on page (Attach if not previously provided) 16 of 28)

Proposed References to be provided to applicants during examination: N 96.06.06.06 For any given plant operating condition when the Emergency Operating Procedures Learning Objective: (As available)

(EOPS) are entered, determine when the lower tier documents may be used.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.45 12 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 164 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.4.26 Importance Rating 3.1 Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage. (CFR: 41.10 / 43.5 / 45.12)

Proposed Question: RO Question 75 Which one of the following should be present in the Diesel Generator Room until loaded conditions have been established and temperatures stabilized?

A. A charged Fire Hose B. A portable CO2 Extinguisher C. A portable Halon Extinguisher D. A portable Water Mist Extinguisher Proposed Answer: D Explanation: When performing a standby diesel generator surveillance test or other manual startup and loading of the SBDG, the plant operator should have a portable water mist extinguisher in the SBDG Room until unit is loaded, operating temperatures have stabilized, and exhaust lagging is not smoking excessively. In this case, should is considered SHALL unless management override is given (AD-AA-10-1006 Procedure Work Instruction Use and Adherence)

A. Incorrect A fire hose station is located outside the north end between the SBDGs, but not required to be charged and stationed when running a SBDG.

B. Incorrect CO2 is not required, Dry chemical and water mist extinguishers are listed on PFP-TB-757 C. Incorrect A halon extinguisher is not required, not listed on the PFP-TB-757 D. Correct OI-324 Rev123, P&L 24 Technical Reference(s): (Attach if not previously provided)

PFP-TB-757, Rev 5 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 165 TR-AA-230-1000-F07, Revision 0

EXAM COVER SHEET Proposed References to be provided to applicants during examination: N 19.01.01.01 Relate the precautions and limitations, operating cautions, or Learning Objective: (As available) procedural notes of OI-324 to any component or SBDG operating status.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 50007_PDA OPS 19-1 NRC Exam RO as given 4-11-19.docx 166 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295004 G2.4.41 Importance Rating 4.6 295004 (APE 4) Partial or Total Loss of DC Power / 6: Generic K/A 2.4.41 - Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 / 45.11)

Proposed Question: SRO Question 1 Given the following:

• The plant has shutdown and is commencing a cooldown following major flooding of the Cedar River

• Reactor Coolant Temperature is 275°F

• 1D10 the Division-1 125 VDC Distribution Panel is de-energized as a result of the flooding

• Output voltage on Division-2 125VDC is 102VDC and stable

`

Which Emergency Action Level should be declared?

A. SA4.1 B. SS6.1 C. SS1.1 D. SS3.1 Proposed Answer: D Explanation:

A. Incorrect: Plausible since this condition could also result in a loss of most of the annunciator power in the plant resulting in this EAL. SA4.1 states an unplanned loss of 75% of control room annunciators with a transient in progress should result in the EAL.

Provided in the stem, a significant transient is NOT in progress thus making this choice wrong. In addition, compensatory indications remain available throughout the evolution.

B. Incorrect: Plausible since the plant is in a transient and has lost or is close to losing all 125VDC power which supplies annunciator power.

C. Incorrect: Plausible if the candidate assumes the loss of control power supplied by 125VDC could lead to a loss of AC power from protective relaying or breaker control power.

D. Correct: Less than 105 VDC bus voltage on BOTH Div 1 and Div 2 125 VDC busses for 15 minutes or longer. With Div1 125VDC Bus deenergized and Div2 125VDC Bus reading less than 105VDC, both DC Electrical sources resulting in the Site Area Emergency.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 1 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

EPIP FORM EAL-01, Rev. 11 EAL Bases Document System Technical Reference(s): (Attach if not previously provided)

Malfunction Category EBD S, Rev.11 page 24 EAL Board EAL-01 Proposed References to be provided to applicants during examination:

Rev 11N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/11 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 2 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295005 AA2.03 Importance Rating 3.1 295005 (APE 5) Main Turbine Generator Trip / 3: AA2.03 - Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Turbine valve position.

(CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 2 The plant is operating at 60% power. TBPV testing is in progress. While depressing the test pushbutton for the No. 1 Bypass valve, the valve fails to move.

What is required for this failure?

Declare the No. 1 Bypass INOPERABLE and ______________.

A. insert a MCPR limit for an inoperable Main Turbine Bypass Valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. restore the Bypass Valve to OPERABLE within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. no further action is required since only 1 Bypass Valve is REQUIRED to be OPERABLE D. lower reactor power to < 21.7% power within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Proposed Answer: B Explanation:

A. Incorrect Required time is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This is plausible since some action requirements for similar conditions are to be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

B. Correct LCO 3.7.7 The Turbine Bypass System shall be Operable , Requirements of the LCO not met Satisfy the LCO within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. Incorrect Tech Spec Bases 3.7.7 An operable main turbine bypass valve requires the bypass valves to open in response to increasing main steam line pressure. No action is plausible since the applicant may believe that 1 TBPV is sufficient for the LCO D. Incorrect LCO 3.7.7 required action A.1 Satisfy the requirements of the LCO within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Action B.1 if the required action and completion time are not met then B.1 Reduce Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 3 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0) thermal Power to < 21.7% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This is plausible since the applicant may determine that we have exceeded the action time of the previous condition.

Tech Spec 3.7.7 Turbine Bypass System (page 3.7-16) Amendment 243 Technical Reference(s): (Attach if not previously provided)

Tech Spec 3.2.2 Minimum Critical Power Ratio (MCPR) page 3.2-2 Amendment 280 Proposed References to be provided to applicants during examination: L3.2.2 and L3.7.7 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/13 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 4 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295016 G2.2.44 Importance Rating 4.4 295018 (APE 18) Partial or Complete Loss of CCW / 8: Generic K/A 2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)

Proposed Question: SRO Question 3 The plant is operating at 100% power. The crew observes the following during panel walkdowns at panel 1C06.

• 1B4208 MO 4841B RBCCW RETURN HEADER ISOLATION breaker has been found in the trip free position and will not reset.

Which one of the following is correct for the given conditions?

A. NO action is required. The RBCCW system remains in service. MO 4841A RBCCW DRYWELL OUTLET ISOLATION is still able to perform the isolation function.

B. Isolate the affected penetration flow path within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and verify the affected penetration flowpath remains isolated once every 31 days C. Isolate the affected penetration flow path within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and verify the affected flowpath remains isolated once every 31 days.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 5 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

D. Isolate the affected penetration flow path within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify the affected flowpath remains isolated once every 31 days.

Proposed Answer: D Explanation: The given conditions makes LCO 3.6.1.3 Condition C applicable. MO4841B is a Type C Containment Isolation Valve. As such with one or more penetration flow paths with one PCIV Inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. Required action C.1 must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

A. Incorrect: The given conditions make MO4841B inoperable. MO4841B is a Type C Containment Isolation Valve. No action is plausible in that the system remains in service and the student may incorrectly conclude there is no isolation requirement since the other valve in the line remains operable.

B. Incorrect: MO 4841B is a Type C Containment Isolation valve and the actions listed are for containment penetration flow paths with 2 PCIVs and one of the PCIVs is INOPERABLE.

C. Incorrect: MO 4841B is a Type C Containment Isolation valve and the actions listed are for containment penetration flow paths with 2 PCIVs and both of the PCIVs are INOPERABLE.

D. Correct: The given conditions makes LCO 3.6.1.3 Condition C applicable. MO4841B is a Type C Containment Isolation Valve. As such with one or more penetration flow paths with one PCIV Inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. Required action C.1 must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

TS LCO 3.6.1.3 Ammendment 234 Technical Reference(s): TSCR 104(Bases) (Attach if not previously provided)

ACP 1410.7 rev. 27 Proposed References to be provided to applicants during examination: LCO 3.6.1.3 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 6 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43/45 5/12 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 7 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295024 EA2.02 Importance Rating 4.0 295024 High Drywell Pressure / 5: EA2.02 - Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Drywell temperature. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 4 The crew has just completed a Fast Power Reduction to 78% power after a steam leak in the drywell was detected.

Current plant conditions:

• Drywell air temperature is 180°F and rising 1°F/ 1 minute

• Drywell pressure is being maintained between 1.0 and 1.5 psig through venting

• The crew is briefing continuation of plant shutdown by inserting control rods What procedure directs the actions the crew takes to address these conditions and what action should be taken next?

A. IPOI 4, Reactor Shutdown - Insert Control rods to get below 70% power B. AOP 573, Primary Containment Control - Insert a Manual Reactor SCRAM C. EOP 2, Primary Containment Control - Anticipate Emergency Depressurization per SEP 307 D. EOP 3, Secondary Containment Control - Continue controlled plant shutdown to cold shutdown conditions Proposed Answer: B Explanation: AOP-573 Follow up action 6 and 7 if drywell pressure cannot be maintained <1.5 psig or Drywell temperature cannot be maintained <180°F THEN reduce reactor power per IPOI-4 Fast power reduction to restore and maintain within limits.

If DW pressure still cannot be maintained <1.5 psig or DW temperature cannot be maintained

<180°F THEN Manually SCRAM A. Incorrect: AOP 573 states that following a fast power reduction, if Drywell air pressure rises to 1.5 psig or DW air temperature rises above 180°F a reactor scram is required.

Plausible if the applicant does not determine that power has already been reduced via fast power reduction and the direction at this point is to SCRAM.. IAW IPOI4 Fast Power reduction, If power is >70% and core flow has been reduced to 39MLB/hr, then continue Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 8 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0) to insert control rods to get below 70%. Then Recirculation flow would be adjusted to 27MLBM/hr.

B. Correct: AOP 573 Step 7 states If DW pressure still cannot be maintained < 1.5 psig OR DW Temp cannot be maintained <150°F Then manually Scram.

C. Incorrect: EOP 2 entry is met being above 150°F. However anticipate ED should not be performed until the reactor is shutdown. Plausible if the applicant fails to implement the step to SCRAM prior to reducing pressure D. Incorrect: The conditions described in the stem contain No EOP 3 entry conditions.

EOP 3 is not applicable based upon these indications. Plausible if the applicant incorrectly applies the guidance for secondary containment control vice primary containment control.

AOP-573 Rev 7(page 4 of 8)

Technical Reference(s): (Attach if not previously provided)

IPOI-4 Rev 141 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43/45 5/13 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 9 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295025 G2.2.40 Importance Rating 4.7 295025 (EPE 2) High Reactor Pressure / 3: Generic K/A 2.2.40 - Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3)

Proposed Question: SRO Question 5 With the plant operating at 20% power the following sequence of events occur:

• 10:00 The Balance of Plant (BOP) Operator reports that the B EHC Pressure Regulator is in service

• 10:02 The At the Controls Operator reports that Reactor Pressure is 1030 psig

• 10:17 Attempts to lower Reactor Pressure were unsuccessful Based upon the above conditions (1) The CRS will direct _________________ to mitigate this condition?

AND (2) What is the FIRST Technical Specification required action that must be completed?

A. (1) AOP 693 Main Turbine/EHC Failures (2) Be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B. (1) AOP 693, Main Turbine/EHC Failures (2) Be in MODE 2 within 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> C. (1) AOP 262, Loss of Reactor Pressure Control (2) Be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D. (1) AOP 262, Loss of Reactor Pressure Control (2) Be in MODE 2 within 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Proposed Answer: C Explanation: AOP-262 Rev 10 should be entered based upon the failure of the pressure control system. Only LCO 3.4.10 is applicable for the given conditions. Since Reactor Steam Dome Pressure cannot be maintained less than 1025 psig and 15 minutes has elapsed, Required Action B.1 is applicable and the Reactor shall be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 10 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

A. Incorrect - AOP 693 does not provide the guidance for a failed pressure regulator. It does provide guidance for EHC Hydraulic failures, not logic failures. LCO 3.4.10 is applicable and the Required Action to be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> should be taken.

This is plausible if the student incorrectly identifies AOP 693 as the procedure that is required to be entered for the given conditions.

B. Incorrect - AOP 693 does not provide the guidance for a failed pressure regulator.

Plausible if the student does not know the governing procedure for the given conditions.

While AOP 693 provides EHC Hydraulic system guidance it does not provide the guidance for inoperable pressure regulators. LCO 3.4.10 is applicable and the Required Action to be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> should be taken. Plausible if the student applies the wrong AOP to address the conditions and incorrectly ascertains that there is no Tech Spec guidance for the given conditions and enters LCO 3.0.3 which requires the plant to be brought to MODE 2 within 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

C. Correct - AOP 262 provides the guidance for a failed pressure regulator. LCO 3.4.10 is applicable and the Required Action to be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> should be taken.

D. Incorrect - AOP 262 provides the guidance for a failed pressure regulator. Plausible if the student incorrectly ascertains that there is no Tech Spec guidance for the given conditions and enters LCO 3.0.3 which requires the plant to be brought to MODE 2 within 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

AOP 262 Loss of Reactor Pressure Control, rev. 10 step 6,(page 4)

LCO 3.0.3, Technical Reference(s): (Attach if not previously provided)

LCO 3.4.10 Reactor Steam Dome Pressure Rev. 224 AOP 693 Main Turbine EHC Failures Proposed References to be provided to applicants during examination: L3.4.10 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 2 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 11 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295026 EA2.01 Importance Rating 4.2 295026 (EPE 3) Suppression Pool High Water Temperature / 5: EA2.01 - Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool water temperature. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 6 The following conditions are present following a Reactor SCRAM:

• Reactor pressure is 910 psig and stable

• Torus water level is 10.4 feet and rising slowly

• Torus water temperature is 160°F and rising slowly

• Drywell pressure is 6 psig and rising slowly

• Drywell air temperature is 220°F and rising slowly The Control Room Supervisor will direct ____________.

A. vent the Drywell B. emergency depressurization C. lowering reactor pressure to 200 psig D. anticipate emergency depressurization Proposed Answer: B Explanation: Heat Capacity Limit Graph 4 has been exceeded. Emergency Depressurization is required. Prior to exceeding Graph 4 EOP-1 PC/P-1 Continuous Recheck statement allows lowering RPV/pressure. If torus water temperature cannot be maintained below the Heat Capacity Limit (Graph 4) and reducing RPV pressure will not result in loss of injection required for adequate core cooling Then maintain RPV pressure below the limit (OK to exceed cooldown rate limit).

EOP-2 T/T-6 Wait until torus water temperature and RPV pressure cannot be maintained below the Heat Capacity limit (Graph 4) Emergency RPV Depressurization is Required.

A. Incorrect: A containment isolation signal precludes the Operators from venting primary containment. Plausible since venting the containment is a viable mitigating action if the student does not recognize the isolation precludes this action.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 12 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

B. Correct: As a result of exceeding the Heat Capacity Temperature Limit, EOP 2 requires an Emergency Depressurization when Torus water temperature cannot be maintained below the Heat Capacity Limit.

C. Incorrect Graph 4 is exceeded. Emergency Depressurization is required. This would be true if only HPCI or RCIC steam drive systems were available for RPV injection and adequate core cooling. No conditions were provided that AC driven injections pumps were not available.

D. Incorrect, Prior to exceeding the Heat Capacity Graph 4, allowances are provided to maintaining reactor pressure below the graph. Once the graph is exceeded Emergency Depressurization is required. Plausible if the student does not understand the interpretation of the step cannot be maintained below.

Technical Reference(s): EOP 2 Bases, Rev.16 page 29 (Attach if not previously provided)

EOP 1 RPV Control step RC/P-2, Rev.20 EOP-2 T/T-6 Rev 18 Proposed References to be provided to applicants during examination: EOP Graph 4 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 13 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 700000 G2.1.20 Importance Rating 4.6 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6: Generic K/A 2.1.20 -

Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

Proposed Question: SRO Question 7 The plant is operating at 100% reactor power. The BOP operator notices the following from the 1C08 panel:

• Generator MVARS are 200 LAGGING

• Essential Bus Voltages 4120V and lowering slowly The ITC has contacted the control room and informed them that storms in the region are expected to impact grid supply and that the grid is entering a limited reserve condition. 161KV Grid Voltage is expected to lower to 159.5KV or 99.07% DAEC post trip contingency voltage.

Based upon these reports and conditions, what should the CRS direct?

A. Enter AOP 304 GRID INSTABILITY, Declare ONLY the Startup Transformer inoperable in accordance with Technical Specification LCO 3.8.1 Condition A B. Enter AOP 304 GRID INSTABILITY, Declare both offsite sources inoperable in accordance with Technical Specification LCO 3.8.1 Condition C.

C. Enter AOP 301 LOSS OF ESSENTIAL ELECTRICAL POWER, Declare both offsite sources inoperable in accordance with Technical Specification LCO 3.8.1 Condition C D. Enter AOP 301 LOSS OF ESSENTIAL ELECTRICAL POWER, Declare ONLY the Startup Transformer inoperable in accordance with Technical Specification LCO 3.8.1 Condition A Proposed Answer: B Explanation: Any potential loss of a grid component (a contingency) other than the DAEC, which would lead to an undervoltage condition in the DAEC switchyard does NOT result in Technical Specification LCO actions.

DAEC actual 161KV switchyard voltages </= 99.2% when 1A3 and 1A4 are connected to the Startup Transformer 1X3 or 345 KV Switchyard voltages </= 98.2% when 1A3 and 1A4 are connected to the Standby Transformer 1X4 may not be an indication of inoperability of offsite power. A trip of the DAEC could potentially cause actual switchyard voltage to increase above minimum value in which case offsite power may be inoperable.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 14 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

If notified by ITC Midwest that contingency trip of the DAEC would lead to an undervoltage condition of </= 99.2% in the DAEC switchyard 161KV bus when 1A3 And 1A4 are connected to Startup Transformer 1X3. OR ITC Midwest confirms a trip of the DAEC would lead to an undervoltage condition of </= 98.2% in the DAEC switchyard 345 KV bus when 1A3 and 1A4 are connected to Standby Transformer 1X4. THEN Declare both Offsite Sources inoperable and enter TS LCO action as required by the mode of applicability.

A. Incorrect: No loss of essential power has ooccurred. 1A3/4 would trip on degraded voltage of 91.3% for 8.5 seconds (3798 volts) or SU or SB transformer output voltages of 65% (2704 volts.) Student may select this choice if they do not understand the tech spec requirement.

B. Correct: If notified by ITC that the contingency trip of DAEC would lead to an undrevoltage condition of < 99.2% in the DAEC switchyard 161KV bus when 1A3 and 1A4 are connected to the SU Xfmr then declare both offsite sources inoperable and enter TS LCO as required.

C. Incorrect: No loss of essential power has ooccurred. 1A3/4 would trip on degraded voltage of 91.3% for 8.5 seconds (3798 volts) or SU or SB transformer output voltages of 65% (2704 volts.) Student may select this choice if they do not understand the tech spec requirement to declare both sources inop.

D. Incorrect: If notified by ITC that the contingency trip of DAEC would lead to an undrevoltage condition of < 99.2% in the DAEC switchyard 161KV bus when 1A3 and 1A4 are connected to the SU Xfmr then declare both offsite sources inoperable and enter TS LCO as required. Student may select this choice if they do not understand the tech spec requirement to declare both sources inop.

AOP-304 Rev 51 Grid Instability Technical Reference(s): Follow up Action 1 page 3 (Attach if not previously provided)

LCO 3.8.1.C Amendment 270 Proposed References to be provided to applicants during examination: N 94.47.01.01 Apply the notes/cautions Learning Objective: of AOP 304 GRID INSTABILITY to (As available) plant conditions Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 15 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 16 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295002 AA2.02 Importance Rating 3.3 295002 (APE 2) Loss of Main Condenser Vacuum / 3: AA2.02 - Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM: Reactor power: Plant-Specific. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 8 The plant was operating at 78% power following a fast power reduction per IPOI 4 due to rising condenser back pressure. A manual reactor SCRAM was inserted. The following are plant conditions:

• Condenser Back Pressure has risen to 8.5Hg and continues to rise 0.2Hg/minute

• Main Turbine has tripped

• MSIVs are closed

• SRVs are repeatedly cycling on their setpoints for several minutes

• Drywell pressure is 5 psig and rising

• Torus water temperature is 90°F and rising rapidly What procedure(s) should the CRS direct FIRST to address these conditions?

A. AOP 693 MAIN TURBINE / EHC FAILURES B. EOP 1 RPV CONTROL and transition to ATWS C. EOP 2 PRIMARY CONTAINMENT CONTROL D. AOP 691 CONDENSER HIGH BACKPRESSURE Proposed Answer: B Explanation: ATWS indications are present. Power and pressure fluctuating with SRV actuation for several minutes. SRVs are not expected to cycle on their setpoints for several minutes following a normal Full SCRAM.

EOP-1 is entered (Scram required with power above 5% or unknown) Continuous Recheck Statement IF any rod is withdrawn past position 00 AND it has not been determined that the reactor will remain shutdown under all conditions without boron THEN Enter ATWS would be the correct action to address the conditions and lower the heat input to the containment prior to addressing degrading containment conditions.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 17 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

A. Incorrect: Although it appears that there is a pressure transient in progress, the pressure changes can be attributed to the SRV actuations on their setpoints in an ATWS condition with the Main Turbine TRIPPED. AOP 693 is plausible if the student believes the Turbine TRIP is the result of EHC failures.

B. Correct: EOP 1 and ATWS provide the proper guidance necessary to deal with the conditions provided.

C. Incorrect: EOP 2 alone does not provide the guidance to address the given conditions of an ATWS. EOP 2 will not provide mitigative steps to eliminate the heat source to the containment which is the most challenging parameter at this time. Plausible if the student chooses to implement this EOP and does not understand the significance of the energy deposition to the containment.

D. Incorrect: While this procedure does provide guidance to deal with the rising condenser backpressure condition, there are other matters for the crew to address that are more pressing, controlling reactor power under the ATWS EOP should be the top priority of the crew.

EOP-1, Rev 20 Technical Reference(s): (Attach if not previously provided)

ATWS EOP Rev 23 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 18 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295022 G2.1.27 Importance Rating 4.0 295022 (APE 22) Loss of Control Rod Drive Pumps / 1: Generic K/A 2.1.27 - Knowledge of system purpose and/or function.

Proposed Question: SRO Question 9 The A CRD Pump is out of service for gearbox oil replacement. The plant has experienced a transient and the CRS has directed a Reactor SCRAM. Reactor Power following the SCRAM is 70%. The following is observed on panel 1C05:

A lockout occurs on 1A4 4160VAC Essential Bus.

Based upon the above conditions which one of the following Rod Insertion Procedures should be DIRECTED to insert the control rods?

A. RIP 103.2 INCREASE CRD COOLING FLOW AND PRESSURE B. RIP 103.3 MANUALLY DRIVE CONTROL RODS C. RIP 102.1 REPEATED MANUAL SCRAM D. RIP 101.3 VENT SCRAM AIR HEADER Proposed Answer: C Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 19 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Explanation: In the stem the plant conditions result in a loss of both CRD pumps. The CRS is required to direct performance of Rod Insertion Procedures per EOP ATWS.

RIP 103.2 and RIP 103.3 require CRD Pumps to operate.

RIP 101.3 is not effective due to the SCRAM valves being OPEN and the SCRAM air header is already depressurized. This is a hydraulic ATWS RIP 102.1 is the only effective manner of inserting the control rods from the provided choices.

Even with the loss of one side of RPS, the procedure is allowed to continue as long as compensatory actions are taken for scram discharge volume draining (6 minutes)

A. Incorrect: This RIP requires CRD pump operation to be effective inserting control rods.

This is plausible if the student does not recognize the loss of both CRD pumps B. Incorrect: This RIP requires CRD pump operation to be effective inserting control rods.

This is plausible if the student does not recognize the loss of both CRD pumps C. Correct: RIP 102.1 is the only effective manner of inserting the control rods from the provided choices. Even with the loss of one side of RPS, the procedure is allowed to continue as long as compensatory actions are taken for scram discharge volume draining (6 minutes)

D. Incorrect: RIP 101.3 is not effective due to the SCRAM valves being OPEN and the SCRAM air header is already depressurized. This is a hydraulic ATWS. This is plausible if the student does not interpre ARP 1C05A Rev 90 (A-6 / A-7)

Technical Reference(s): (Attach if not previously provided)

A/B CRD Pump Trip Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 20 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295032 EA2.02 Importance Rating 3.5 295032 (EPE 9) High Secondary Containment Area Temperature / 5: EA2.02 - Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Equipment operability. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 10 The plant is operating at 100% reactor power when the following occurs:

• Steam Leak Detection high ambient temperature

• The Operator attempted to manually CLOSE MO2238 HPCI INBD STEAM LINE ISOLATION and MO2239 HPCI OUTBD STEAM LINE ISOLATION

• HPCI Room temperature is 185°F and rising Based upon the following conditions answer the following:

(1) What is the next required action?

AND (2) What is the basis for this action?

A. (1) SCRAM the reactor and enter IPOI 5 (2) To ensure equipment necessary for the safe shutdown of the facility will not fail B. (1) Emergency Depressurize when the same parameter exceeds its max safe operating limit in 2 or more areas (2) To ensure equipment necessary for the safe shutdown of the facility will not fail C. (1) SCRAM the reactor and enter IPOI 5 (2) To ensure the HPCI System remains operable D. (1) Emergency Depressurize when the same parameter exceeds its max safe operating limit in 2 or more areas (2) To ensure the HPCI System remains operable Proposed Answer: A Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 21 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Explanation: EOP 3 Bases SC-5 Before any parameter reaches its MAX SAFE Operating Limit Scram MSOL are defined as the highest parameter value at which neither (1) equipment necessary for the safe shutdown of the plant will fail nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. Room temperature is at 185°F and rising already above safe personnel entry.

A. Correct: Step SC-2 has failed Isolate all systems discharging into the area Max Normal Temp has been exceeded and continues to rise.

Step SC-4 Will RPV pressure reduction decrease leakage into secondary containment (Yes) HPCI is a primary system. So action at SC-5 is Scram prior to reaching MSOL would be correct. MSOL for HPCI room is 310°F B. Incorrect, waiting for the same parameter to exceed MSOL in same area is ED Criteria.

Student may incorrectly apply this step if previous step not implemented.

C. Incorrect. The HPCI system should be declared inoperable given the conditions in the stem.

D. Incorrect: waiting for the same parameter to exceed MSOL in same area is ED Criteria.

Student may incorrectly apply this step if previous step not implemented. The HPCI system should be declared inoperable given the conditions in the stem.

EOP 3, Rev.22 Step SC-5 Technical Reference(s): (Attach if not previously provided)

EOP-3 Bases Rev 13 (page 20)

Proposed References to be provided to applicants during examination: EOP-3 Table 6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 22 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 203000 G2.4.20 Importance Rating 4.3 203000 (SF2, SF4 RHR/LPCI) RHR/LPCI: Injection Mode: Generic K/A 2.4.20 - Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 11 Following Emergency Depressurization, plant conditions are as follows:

• RPV Water Level is -10 inches and stable

• ONLY one RHR Pump is available and is injecting at 4800 GPM

• Suppression pool water level is 7.5 feet and slowly rising

• Suppression pool temperature is 212°F and stable

• Drywell Pressure is 3.3 psig and stable

• Torus Pressure is 3.0 psig and stable Which one of the following actions is required?

A. Continue RPV Injection Flowrate B. Place Torus Cooling in service maximized C. Reduce RHR Flowrate to maintain within the NPSH limits for 1 RHR Pump operation D. Reduce RHR Flowrate to maintain within the VORTEX limits for 1 RHR Pump operation Proposed Answer: A Explanation: Where references to this caution (3) occur, the identified systems should be operated within the NPSH and Vortex limits if possible. If the situation warrants. However, the limits may be exceeded. A judgment as to whether a pump should be operated beyond its limit in a particular event should consider Immediate and catastrophic failure is not expected if a pump is operated beyond the NPSH or vortex limit. The undesirable consequences of uncovering the reactor core should thus outweigh the risk of equipment damage.

A. Correct: RPV/L is <+15 inches. Injection should continue to restore adequate core cooling.

B. Incorrect: RPV/L is <+15 inches. Inappropriate to divert RHR injection away from the core. Student may select this choice if they believe cooling the Torus takes precedence to injection.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 23 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

C. Incorrect: RPV/L is <+15 inches. Inappropriate to divert RHR injection away from the core. While this may prevent damage to the RHR pumps restoring adequate core cooling takes precedence.

D. Incorrect: RPV/L is <+15 inches. Inappropriate to divert RHR injection away from the core. While this may prevent damage to the RHR pumps restoring adequate core cooling takes precedence.

EOP Cautions Bases Rev 12 page 18.

EOP Bases Breakpoints page 7 of Technical Reference(s): (Attach if not previously provided) 14 (+15 inch ) loss of adequate core cooling through core submergence Top of active fuel EOP CAUTION Figure 19 (NPSH)

Proposed References to be provided to applicants during examination:

and Figure 23 (Vortex)

Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 24 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 206000 A2.05 Importance Rating 3.8*

206000 (SF2, SF4 HPCIS) High Pressure Coolant Injection: A2.05 - Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM (HPCIS);

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: D.C. electrical failure: BWR-2,3, and 4.

(CFR: 41.5 / 45.6)

Proposed Question: SRO Question 12 An electrical malfunction occurs on 1D4102 MO-2321 HPCI INBD TORUS SUCTION ISOLATION. Report from the field is that the breaker has visible damage and an acrid odor is present.

HPCI is in a normal system lineup.

(1) What is the effect, if any, on Tech Spec requirements?

AND (2) What would be the required action, if any?

A. (1) HPCI remains fully operable, only CST suction is required for operability.

(2) There are no required TS actions.

B. (1) HPCI is inoperable with the loss of the ability to transfer suction to the Torus.

(2) Restore HPCI to operable status in 14 days.

C. (1) 1D41 250V HPCI DC electrical power distribution panel is inoperable.

(2) Declare supported features inoperable immediately.

D. (1) HPCI Low CST and Torus High Level Instrumentation inoperable.

(2) Declare HPCI system inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Proposed Answer: B Explanation: HPCI is assumed in the accident analyses to have the suction path aligned to the Torus to ensure long term ECCS capability. The loss of motive power to MO2321 prevents automatic transfer from the normal lineup to the Torus suction path. Manual alignment to the Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 25 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Torus suction is an option, but the given conditions are not compliant with TS until realignment occurs.

A. Incorrect Torus suction path is required for long term ECCS. Plausible if candidate believes CST suction is the only required path.

B. Correct HPCI is assumed in the accident analyses to have the suction path aligned to the Torus to ensure long term ECCS capability. The loss of motive power to MO2321 prevents automatic transfer from the normal lineup to the Torus suction path.

C. Incorrect Loss of one breaker on any distribution panel only requires operability to be assessed for the individual load. Plausible if candidate believes a malfunction on the panel requires the entire panel to be considered inoperable.

D. Incorrect Inability of the end device to change state does not impact the operability status of the instrumentation. The instrumentation is still fully capable to perform its intended function. Plausible as the instrumentation that feeds the isolation valve.

Tech Spec 3.5.1 Amendment 305 Technical Reference(s): Bases page B3.5-3. (Attach if not previously provided)

OI152 HPCI System P&L 8 TS 3.3.5.1 w/Table 3.3.5.1 Proposed References to be provided to applicants during examination:

TS 3.5.1 TS 3.8.7 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 26 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 211000 G2.4.6 Importance Rating 4.7 211000 (SF1 SLCS) Standby Liquid Control: Generic K/A 2.4.6 - Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 13 HPCI and ACRD pump are tagged out for maintenance when a Loss of Coolant Accident AND a loss of offsite power has occurred.

The following conditions now exist:

• RPV water level is 105 inches and lowering slowly

• Reactor pressure is 850 psig and lowering slowly

• 1A4 Bus Lockout

• RCIC is operating and injection maximized For the current plant conditions, what alternate injection procedure should be directed to address the lowering RPV Water Level?

Direct alternate injection systems placed in service per _________.

A. AIP 401, Injection with RHRSW B. AIP 404, Injection with Fire Water C. AIP 406, Injection with SBLC D. AIP 407, Maximize CRD injection Proposed Answer: C Explanation: With the current RPV pressure the low pressure systems (Fire Water and RHRSW will be below their shutoff head and will not inject into the RPV).

No CRD pumps are available 1P209A is tagged out of service for maintenance, and 1P209B is lost with the 1A4 Bus Lockout 1P230A SBLC is available powered by 1B34 (1G31, A SBDG)

A. Incorrect Discharge pressure of RHRSW is 0-270 psig, well below the given RPV pressure. The student may select this since this is an alternate injection source.

B. Incorrect Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 27 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Fire system pressure is 0-125 psig, well below the given RPV pressure. . The student may select this since this is an alternate injection source.

C. Correct 1B34 available via 1G31 A SBDG and 1P230A A SBLC pump available Discharge pressure 0-1400 psig D. Incorrect 1P209 A CRD is tagged out for maintenance 1A4 is locked out so 1P-209B will have no power. While the discharge head would be sufficient for injection, the system is unavailable given the stem conditions. The student may select this if they do not make the correlation that the B CRD pump is unavailable due to the lockout of 1A4.

EOP-1 Rev 20 Alternate Injection Systems Table Technical Reference(s): 2A (Attach if not previously provided)

AOP 301 Rev 75 (page 49 and 50) 1P230A breaker 1B3445 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 28 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 217000 A2.12 Importance Rating 3.0 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling: A2.12 - Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve openings. (CFR: 41.5 / 45.6)

Proposed Question: SRO Question 14 The plant has experienced a loss of offsite power. The BOP Operator is directed to place RCIC in service per the OI 150 QRC 1 RCIC RAPID START to maintain RPV Water level. While starting RCIC the following is observed:

What Procedure and Action should be directed to start RCIC from this condition?

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 29 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

A. ARP 1C04C A-5 RCIC MO-2405 Turbine Trip. Close MO-2405 and then OPEN MO-2405 B. ARP 1C04C A-5 RCIC MO-2405 Turbine Trip. Close MO-2405. MO-2405 will OPEN Automatically once fully closed C. SAMP 703 RCIC OPERATION FOLLOWING LOSS OF ELECTRICAL POWER. Close MO-2405 and then OPEN MO-2405 D. SAMP 703 RCIC OPERATION FOLLOWING LOSS OF ELECTRICAL POWER. Close MO-2405. MO-2405 will OPEN Automatically once fully closed Proposed Answer: A Explanation: 1C04B (A-5) RCIC MO-2405 Turbine Trip provides guidance to restart RCIC MO2405 does not have an auto re-open. The action is to manually CLOSE MO-2405 RCIC TURBINE STOP VALVE motor control till it indicates fully CLOSED, then throttle OPEN MO-2405 RCIC TURBINE STOP VALVE motor control as necessary for required injection.

A. Correct: ARP 1C04C A-5 RCIC MO-2405 Turbine Trip provides guidance to restart RCIC B. Incorrect: ARP 1C04C A-5 RCIC MO-2405 Turbine Trip provides guidance to restart RCIC. MO-2405 will not auto re-open. Student may select if they think MO 2405 will automatically open. It will not.

C. Incorrect: SAMP 703 is to operate RCIC when Div 1 125 VDC batteries are no longer available. Batteries are still available. Student may select this however the SAMP is not authorized for this condition.

D. Incorrect: SAMP 703 is to operate RCIC when Div 1 125 VDC batteries are no longer available. Batteries are still available. Student may select this however the SAMP is not authorized for this condition.

SAMP 703 Rev 8 Technical Reference(s): (Attach if not previously provided)

ARP 1C04C (A-5) rev 53 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 30 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 31 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262001 G2.2.42 Importance Rating 4.6 262001 (SF6 AC) AC Electrical Distribution: Generic K/A 2.2.42 - Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7 / 41.10 / 43.2

/ 43.3 / 45.3)

Proposed Question: SRO Question 15 While operating at 100% reactor power the following annunciator is received:

• 1C08A (C-5), LC 1B3 BREAKER 1B301, 1B302, 1B303, 1B304 TRIP As a result of the conditions associated with the alarm above, the following equipment loses power:

• 125VDC Battery Charger 1D12

• 250VDC Battery Charger 1D43

• A Standby Filter Unit

• ESW Pump 1P99A

• RWCU MO2700 Which one of the following is correct regarding TS actions required to address these conditions?

A. 3.8.1 Condition B, restore the SBDG to OPERABLE status within 7 days B. 3.7.3 Condition A, restore the ESW subsystem to OPERABLE status within 7 days C. 3.8.4 Condition A, restore 125VDC Electrical Power Subsystem to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D. 3.8.7 Condition A restore AC Electrical Power Distribution System to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Proposed Answer: D Explanation: TS 3.8.7 bases. The AC and DC electrical power distribution subsystems listed in the LCO are required to be Operable. Based on the number of safety significant electrical loads associated with each bus listed in Table 3.8.7-1, if one or more of the buses become inoperable, entry into the appropriate ACTIONS of LCO 3.8.7 is required. Action A.1 the required AC buses, load centers, motor control centers, and distribution panels must be restored to Operable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 32 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

A. Incorrect Important electrical power is lost 1G312, 1V-SF-020 supply fan are lost. Bus 1B32 and or 1B34 must be restored within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per 3.8.7.A. This may be selected if the student thinks they should cascade. They should not.

B. Incorrect 1B32 is lost as well as 1B34. 3.8.7.A requires restoration within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This may be selected if the student thinks they should cascade. They should not.

C. Incorrect 1B32 is lost and 1D12 125VDC charger. 125 VDC distribution is still maintained via battery. TS 3.8.7.still is applicable. This may be selected if the student thinks they should cascade. They should not.

D. Correct: The equipment listed lost power based upon the loss of the AC Bus. Thus, the applicant should know to address Distribution Systems operating and not cascade to the individual equipment specifications unless directed to do so.

TS 3.8.7 / bases page 3.8-65 Technical Reference(s): (LCO 3.0.6) (Attach if not previously provided)

LCO 3.0.6 page B 3.0-9 LCO 3.8.7 LCO 3.8.4 Proposed References to be provided to applicants during examination:

LCO 3.7.3 LCO 3.8.1 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7,10 55.43 2-3 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 33 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 201003 A2.06 Importance Rating 3.1 201003 (SF1 CRDM) Control Rod and Drive Mechanism: A2.06 - Ability to (a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Suction strainer(s) becoming plugged. (CFR: 41.5 / 45.6)

Proposed Question: SRO Question 16 The plant is operating at 100% reactor power. The previous shift established more frequent monitoring of CRD Suction Filter pressure due to a shift in suction pressure trends. While reviewing Logs, the CRS notices CRD Suction Filter pressures as follows:

00:00:00 9 Hg vacuum 01:00:00 10 Hg vacuum 02:00:00 12 Hg vacuum 03:00:00 14 Hg vacuum Per OI 255 CRD System, assuming the current trend, what action should the CRS direct for this condition prior to 0500?

A. START the standby CRD Pump B. Place the standby Suction Filter in service C. Throttle CLOSED V-17-24 Charging Water Isolation D. Align Condensate Service for CRD Control Rod Assemblies Proposed Answer: B Explanation: OI-255 Section 4.0 Normal Operation, CRD suction pressure high limit is listed as 16 Hg vacuum. ARP 1C05A (B-6 / B-7) A/B CRD Pump Lo Suct Pressure, setpoint is 18Hg.

Automatic action is CRD pump trips after 15 second time delay. Crew would place standby filter into service prior to the low suction pressure trip A. Incorrect: Starting the standby CRD pump would further challenge degrading suction pressure. This action would be non-conservative and could cause a loss of the CRD system. Section 6.1 Student may choose if they determine additional CRD flow is necessary to overcome conditions.

B. Correct: see explanation above.

C. Incorrect: Although this would raise suction pressure, it is not authorized with the plant operating at power. Conditions to throttle closed this valve are provided during outage Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 34 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0) or emergency conditions while implementing Rod Insertion Procedure 103.3 Manually Drive Control Rods.

D. Incorrect: This requires both CRD pumps tagged out and is an outage only option.

Plausible in that this action can be performed in the plant. This is performed in accordance with OI 255 section 5.2 Providing Condensate Service for CRD Flush.

OI-255 rev 97 Section 4.0 Normal Operation Technical Reference(s): (Attach if not previously provided)

ARP 1C05A (B-6/B-7) rev 90 CRD Pump Lo Suction Pressure Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.45 6 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 35 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 202001 G2.4.30 Importance Rating 4.1 202001 (SF1, SF4 RS) Recirculation: Generic K/A 2.4.30 - Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)

SRO Question 17 The plant is operating at 100% reactor power. A sequence exchange is scheduled to be conducted.

• Reactor Power is required to be lowered to 55%

Just prior to this adjustment, who (by title) is required to be informed?

A. Real time desk and the NDDO B. Real time desk and NRC resident C. Operations Director and the NDDO D. RX Engineering Supervisor and NRC resident Proposed Answer: A Explanation: REDP 8 Sequence exchange requirements prerequisites require Real Time Desk notification and NDDO notification. Page 9 of 32 rev 77 A. Correct B. Incorrect: NRC resident not required for normal power reduction. If a major power adjustment is required NRC resident should be informed.

C. Incorrect: Operations Director not required for normal power reduction. If a major power adjustment is required Operations Director should be informed.

D. Incorrect: RX Engineer developed the plan and the NRC resident not required for normal power reduction Prior Reactivity Management Technical Reference(s): REDP 8 rev 77 (page 9) Plan with prerequisites page (Attach if not previously provided)

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 36 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 37 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 234000 K1.04 Importance Rating 3.3 234000 (SF8 FH) Fuel Handling Equipment: K1.04 - Knowledge of the physical connections and/or cause-effect relationships between FUEL HANDLING EQUIPMENT and the following:

Reactor manual control system. (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Proposed Question: SRO Question 18 The Mode Switch is in REFUEL and all control rods are inserted. The Refueling bridge Operator grappled a fuel bundle, partially raised the grapple, and commenced moving the bundle from the spent fuel pool towards the core.

Which one of the following describes what will result as the Refueling Bridge moves towards the core?

The Refueling Bridge __________.

A. continues over the core and initiates a control rod block B. continues over the core and causes NO other protective actions C. stops before it reaches the core and initiates a control rod block D. stops before it reaches the core and causes NO other protective actions Proposed Answer: A Explanation: Rod Out Block = Fuel Grapple Not Full Up and Refuel Platform Over or Near the Core AND Mode Switch in Refuel Road Block = Fuel Grapple Not Full Up, AND Not All Rods Full In and Refuel Platform Over or Near the Core AND Mode Switch in Refuel A. Correct All rods are full in therefore a Road block does not occur. The refuel bridge moves unimpeded in this case. A Rod block is initiated due to the bridge being over the core with the mode switch in REFUEL B. Incorrect A rod block is initiated with these conditions. Student may choose if they do not understand the rod block circuitry.

C. Incorrect The refuel bridge is not impeded with these conditions. Student may choose if they do not understand the Refuel Bridge movement protective circuitry.

D. Incorrect Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 38 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

The refuel bridge is not impeded with these conditions. Student may choose if they do not understand the Refuel Bridge movement protective circuitry. A rod block is initiated with these conditions. Student may choose if they do not understand the rod block circuitry.

Technical Reference(s): SD-281, Rev 9 (page 8 and 9) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 NRC ILT Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2-9 55.45 7 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 39 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.1.23 Importance Rating 4.4 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)

SRO Question 19 Following a rise in condenser backpressure, reactor power has been lowered. The OATC determined that load line has exceeded 100.64%.

(1) Which of the following should be directed?

AND (2) Is a reactivity plan required?

A. (1) Reactor Recirc Flow shall be lowered (2) A reactivity plan is required B. (1) Reactor Recirc Flow shall be lowered (2) A reactivity plan is NOT required C. (1) Control Rods shall be inserted (2) A reactivity plan is required D. (1) Control Rods shall be inserted (2) A reactivity plan is NOT required Proposed Answer: D Explanation: IPOI-3 Section 5.0 Lowering Power to 35%, step 6 Continuous Recheck Statement If inadvertent entry into the buffer or exclusion areas of the power to flow map, Then comply with the requirements of AOP-255.2 Power/Reactivity Abnormal Change. This will direct control rod insertion, no reactivity plan is required.

AOP-255.2 step 8 follow up actions In the event of inadvertent entry into the area above the power to flow map (ie exceeding the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average load line of 100.64%) exit this area by inserting control rods If unable to reduce power below the MELLA limit within one hour, manually scram the reactor A. Incorrect Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 40 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

This would cause Load Line to increase further. While this may lower power the power to flow conditions would worsen.

B. Incorrect This would cause Load Line to increase further. While this may lower power the power to flow conditions would worsen.

C. Incorrect, a reactivity plan is not required because the direction is provided with the AOP for the given conditions. Student may select if they believe a reactivity plan is required for all power changes.

D. Correct: Control rod insertion is the only viable method for correcting this condition.

This action will reduce the Load Line giving more operating margin to any limits.

Direction is provided within the AOP to perform this action. A reactivity plan is not required.

IPOI-3 Rev 160 P&L 10 and 11 (page 7 of 34)

Technical Reference(s): (Attach if not previously provided)

AOP-255.2 rev 48 step 8 (page 6 of 13).

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2013 NRC Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 41 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.1.43 Importance Rating 4.3 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. (CFR: 41.10 / 43.6 / 45.6)

Proposed Question: SRO Question 20 The plant is operating at 100% reactor power when the B MSR Drain tank dump valve CV1077A fails OPEN. Reactor Power rises ~ 6 MWth and stabilizes.

Based upon these indications what procedure(s) should be entered and what action should be directed to address these conditions?

A. AOP 255.2 Power Reactivity Abnormal and AOP 646 Loss of Feedwater Heating.

Perform a fast power reduction to 39 MLBM/hr to maintain within Modified Exhaust Pressure Limit B. AOP 255.2 Power Reactivity Abnormal and AOP 646 Loss of Feedwater Heating.

Lower recirc flow as necessary to maintain reactor power within licensed power limitations C. AOP 255.2 Power Reactivity Abnormal and IPOI 5 SCRAM, remove the unit from service due to entering unanalyzed loss of feedwater heating region D. SCRAM the Reactor in accordance with IPOI 5 SCRAM. CLOSE the MSIVs to stabilize RPV pressure Proposed Answer: B Explanation: AOP 646 Rev 25. Immediate Action is If reactor thermal power exceeds 1912 MWTh, Then lower reactor power with recirc as necessary to maintain core thermal power less than 1912 MWTh.

AOP 255.2 Power Reactivity Abnormal Rev 48. Step 1 is Take any necessary steps to bring the reactor/reactivity transient under control, including, but not limited to

• Adjusting recirculation flow as necessary to bring power within specified operating limits.

• Scramming the reactor

• Inserting control rods per the current rod pull sheet

• Assuming manual control of a malfunctioning system A. Incorrect Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 42 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Fast power reduction is not required. This action is too severe for the given condition.

A normal reactivity adjustment is warranted to lower power within the license thermal limit. In addition the action is not performed to maintain below the MEPL.

B. Correct Restore power to <1912 MWth with recirc flow reduction. This is an immediate action with AOP 646.

C. Incorrect Prompt action is taken to reduce power to <1912 MWth. The required action is to lower power to maintain within the Licensed power limit. There is no guidance to shutdown the reactor in this case.

D. Incorrect Scram is not required, closing the MSIVs is not required. Student may select this action to isolate the MSR from the Main Steam system.

AOP-255.2 Rev 48 Technical Reference(s): (Attach if not previously provided)

AOP 646 Rev 25 (page 2)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 6 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 43 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.2.25 Importance Rating 4.2 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2)

Proposed Question: SRO Question 21 A reactor startup is in progress following a refueling outage. Control rod 30-19 was declared inoperable and inserted to position 00. The control room supervisor has directed control rod 30-19 be disarmed.

Which of the following correctly describes how this task can be accomplished?

Per the TS 3.1.3 Bases, Control rod drive 30-19 can be electrically disarmed by removing

___(1)____ or it can be hydraulically disarmed by ___(2)____

A. (1) amphenol connectors from all four directional control valve solenoids for HCU 30-19 (2) closing drive water, exhaust water, and cooling water isolation valves for HCU 30-19 B. (1) fuses at the SCRAM SOLENOID FUSE PANEL for SV-1856 and SV-1855 HCU SCRAM PILOT SOLENOID VALVES for HCU 30-19 (2) closing the drive water, exhaust water, cooling water isolation for HCU 30-19 C. (1) amphenol connectors from all four directional control valve solenoids for HCU 30-19 (2) closing drive water and exhaust water isolation valves for HCU 30-19 D. (1) fuses at the SCRAM SOLENOID FUSE PANEL for SV-1856 and SV-1855 HCU SCRAM PILOT SOLENOID VALVES for HCU 30-19 (2) closing the drive water, exhaust water, isolation for HCU 30-19 Proposed Answer: C In accordance with TS Bases 3.1.3: (C.1 and C.2) The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids.

(A.1, A.2, A.3, and A.4) The control rod isolation method should also ensure cooling water to the CRD is maintained.

A. Incorrect: Closing the cooling water isolation to the HCU would cause the Drive to heat up and could result in seal degradation if not corrected. Student may select if they think complete isolation of the CRDM is required.

B. Incorrect: Fuses are not required to be pulled for electrical isolation of the CRDM. This action would fail open the SCRAM inlet and outlet valves. The amphenols for the Directional Control Valves are required to be isolated. Closing the cooling water Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 44 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0) isolation to the HCU would cause the Drive to heat up and could result in seal degradation if not corrected. Student may select if they think complete isolation of the CRDM is required.

C. Correct: Per TS Bases disconnecting the amphenols and isolating the Drive hydraulically is a means of disarming the CRD.

D. Incorrect: Fuses are not required to be pulled for electrical isolation of the CRDM. This action would fail open the SCRAM inlet and outlet valves. The amphenols for the Directional Control Valves are required to be isolated.

Tech Spec 3.1.3 Bases page Technical Reference(s): (Attach if not previously provided)

(B3.1-18 and 19) Amendment 223 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Monticello Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 2 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 45 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.2.6 Importance Rating 3.6 Knowledge of the process for making changes to procedures. (CFR: 41.10 / 43.3 / 45.13)

Proposed Question: SRO Question 22 While performing a section of Operating Instruction (OI) 410, River Water Supply System, the reactor operator comes across this step:

(27) Place Traveling Screen Wash Pump 1P-112A[B] in service by placing handswitch HS-2906A[B] to the AUTO position on breaker 1B9106 [1B2106].

This step is unable to be performed. The Reactor Operator (RO) has determined performance of the step is NOT required.

Which one of the following meets the approval requirements to place an NA in the block for this step?

A. Two Senior Reactor Operators B. Only the Operations Shift Manager C. The STA and the RO performing the procedure D. The Control Room Supervisor and the RO performing the procedure Proposed Answer: A Explanantion: AD-AA-100-1006 Procedure Work Instruction Use and Adherence, page 28 section 4.8 step 5D, A. Correct: Two (2) SROs are required to review, approve, and initial an NA to a step in a work instruction or procedure that are for safety related equipment or equipment that supports technical specifications.

B. Incorrect: The OSM would need a second SRO. Plausible in that many processes require just the OSM permission to complete. In this case, an additional SRO is required to approve.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 46 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

C. Incorrect: Two SROs are required to perform this action. Plausible in that many processes require only one SRO to satisfy requirements. This case however requires 2 SROs.

D. Incorrect: The CRS would need a second SRO. Plausible in that many processes require only one SRO to satisfy requirements. This case however requires 2 SROs.

AD-AA-100-1006 Procedure Work Technical Reference(s): Instruction Use and Adherence, (Attach if not previously provided)

Rev 16 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: 2017 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 47 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.3.11 Importance Rating 4.3 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)

Proposed Question: SRO Question 23 The plant is shutdown in response to an event. While attempting to close the MSIVs due to elevated Main Steam Line radiation levels, the B MSL inboard and outboard MSIVs failed to CLOSE. The following conditions exist:

• Reactor Building ventilation isolated due to HI radiation levels

• Reactor Building KAMAN on EMS is indicating HI-HI

• Turbine Building KAMAN reading 3.5e-02 ci/cc and rising

• North Refuel Floor Radiation Monitor RI-9163 reads 100 mr/hr and rising slowly

• CRD repair room radiation levels are reading 1000 mr/hr and rising slowly What procedure and action should be directed to address the conditions above?

A. EOP 3 SECONDARY CONTAINMENT CONTROL, Restart the Reactor Building ventilation systems by installing Defeat 9 B. EOP 3 SECONDARY CONTAINMENT CONTROL, perform an Emergency Depressurization based upon 2 areas above max safe C. EOP 4, perform an Emergency Depressurization based upon exceeding Site Area Emergency Radiation Release levels D. EOP 4, Stop the Main Plant Exhaust Fans, Stop TB Supply Fans, verify at least 1 TB Exhaust Fan operating at High Speed Proposed Answer: D Explanation:

A. Incorrect EOP-3 CRS states if all the following conditions apply RB Vent shaft RIM-7606A(B) is below 8 mR/hr. This is not true. Defeat 9 is not allowed.

Student may select if they do not understand the limitation of use of the defeat.

B. Incorrect:

EOP-3 SC-4 decision will RPV pressure reduction decrease leakage into secondary containment. Stem does not indicate that a primary system leak is present in secondary Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 48 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0) containment. SC-4 would be answered No and continue to step SC-8 shutdown per IPOI-3,4, or 5.

C. Incorrect:

TB radiation levels are at the Alert, conditions are not met to ED per EOP-4. Student may select if they do not determine the appropriate threshold to ED.

D. Correct EOP-4 Continuous Recheck Statement (CRS) provides the guidance to perform these actions. In the stem we provided a Group 3 isolation due to the high radiation level, and a RX Bldg KAMAN HI-HI level. This triggers the CRS.

EOP-3 Rev 22 Technical Reference(s): (Attach if not previously provided)

EOP-4 Rev 22 EAL-01 Rev 11 EAL Board EAL-01 Proposed References to be provided to applicants during examination:

EOP-3 Table 6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11 55.43 4 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 49 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.4.32 Importance Rating 4.0 Knowledge of operator response to loss of all annunciators.(CFR: 41.10 / 43.5 / 45.13)

Proposed Question: SRO Question 24 The plant is operating at 100% reactor power when the following conditions are observed:

• All control rods SCRAM full in

• All systems respond as designed to the SCRAM and are indicating properly

• No annunciators from 1C08, 1C05, 1C04, or 1C03 alarm

• Maintenance has determined that repairs will take at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> What (if any) EAL is required to be declared?

A. NO EAL is required B. Unusual Event C. Alert D. Site Area Emergency Proposed Answer: C Explanation: EAL SA4.1 Unplanned Loss of approximately 75% or more of ANY of the following for 15 minutes or more:

• 1C03, 1C04, and 1C05 annunciators

• 1C03, 1C04, and 1C05 indications

• Radiation monitor indications AND Either of the following

• A SIGNIFICANT Transient is in progress

• Compensatory indications are unavailable EAL Bases Document EBD S (system malfunction) describes a significant transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10%

or greater.

A. Incorrect AN EAL threshold has been met. Student may select if they do not believe there are enough annunicators lost to declare.

Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 50 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

B. Incorrect An Alert should be declared for the given conditions. If the student does not determine a significant transient in progress they may select this choice.

C. Correct Alert threshold has been met D. Incorrect Alert threshold has been met Site threshold would be met if Compensatory indications were not unavailable. Student may select if they believe compensatory indications are not available either.

AOP-302.2 LOSS OF ALARM PANEL POWER EAL-01 Rev 11 Technical Reference(s): (Attach if not previously provided)

EAL Bases Document EBD S Rev 10 (page 17 of 31) describes significant transient.

Proposed References to be provided to applicants during examination: EAL Board EAL-01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 51 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.4.50 Importance Rating 4.0 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 / 43.5 / 45.3)

Proposed Question: SRO Question 25 The plant was operating at 100% reactor power with Torus Cooling in service.

• Day 1 of 7 per TS 3.5.1.B The following annunciator is received:

• 1C03A(A-8), A Core Spray System AUTO Initiation Thirty (30) seconds later the following annunciator is received:

• 1C03A(A-9), A Core Spray Pump 1P-211A Trip or Motor Overload What is the FIRST required action for this condition?

A. Restore Drywell Spray Subsystem to FUNCTIONAL within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. Restore the RHR Subsystem to OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C. Restore the A Core Spray Subsystem to OPERABLE within 7 days D. Enter LCO 3.0.3 IMMEDIATELY Proposed Answer: D Explanation: Tech Specification 3.5.1 B One Low Pressure ECCS subsystem inoperable for reasons other than Condition A (One RHR Pump inoperable).

3.5.1.N Two or more low pressure ECCS subsystems inoperable for reasons other than Conditions C or D (Enter LCO 3.0.3)

Condition C One CS subsystem inoperable and one or two RHR pumps inoperable (RHR pumps are not inoperable in this situation, the LPCI mode is inoperable due to being in Torus Cooling.

Condition D Both CS subsystems inoperable. Only the A" CS subsystem is inoperable.

OI-149 Section 5.4 Normal Torus Cooling Continuous Recheck Statement If Torus Cooling is operating when LPCI is required to be Operable Then LPCI shall be declared inoperable and the Technical Specification for ECCS-Operating and RPV Water Inventory Control complied with Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 52 TR-AA-230-1000-F07, Revision 0

EXAM (60006_PDA OPS 19-1 NRC Exam, Rev. 0)

A. Incorrect:

Drywell Spray remains operable in these conditions. Student may incorrectly assume the LPCI inoperability due to Torus Cooling also makes Drywell Spray inoperable.

B. Incorrect:

3.5.1.B was entered for the LPCI subsystem inop while in Torus cooling. Condition C is not applicable for the given conditions.

C. Incorrect:

Although Core Spray is inoperable in the given condition, student must realize that two low pressure ECCS subsystems are inoperable and correct condition would be N enter LCO 3.0.3 D. Correct: 3.5.1.N Two or more low pressure ECCS subsystems inoperable for reasons other than Conditions C or D (Enter LCO 3.0.3 IMMEDIATELY)

TS 3.5.1 rev 262 Technical Reference(s): OI-149 rev 169 Section 5.4 Normal (Attach if not previously provided)

Torus Cooling LCO 3.5.1 and TRM Proposed References to be provided to applicants during examination:

3.5.1 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Filename: 60006_PDA OPS 19-1 NRC Exam SRO as given 4-12-19.docx 53 TR-AA-230-1000-F07, Revision 0