Text

2020 DC Cook Ile As-Administered RO SRO Written Examination
	ML20356A257
Person / Time
Site:	Cook
Issue date:	08/25/2020
From:	; Indiana Michigan Power Co
To:	; NRC/RGN-III/DRS/OLB
	Reeser D
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ML19121A512	List: ML20356A257; ML20356A259; ML20356A265; ML21005A431;
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	Download: ML20356A257 (204)
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EXAMINATION ANSWER KEY RO32NRC-SRO 1 ID: CM-8693A Points: 1.00 The plant has experienced an ATWS condition. Boration is in progress in accordance with FR-S.1, Response to Nuclear Power Generation/ATWS.

Which of the following meets the criteria to transition from FR-S.1?

Wide Range Log Power Wide Range SUR A. 3% -0.1 dpm B. 5% +0.0 dpm C. 6% -0.3 dpm D. 4% +0.1 dpm Answer: A Answer Explanation:

The procedure checks the reactor is Subcritical by verifying WR log power < 5% and WR start up rate is Negative.

A. Correct - see above B. Incorrect - SUR is not negative.

C. Incorrect - Wide Range power > 5%.

D. Incorrect - SUR is not negative.

Question ID (Status) CM-8693A (Active)

External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: F/2 Comments:

Reference:

OHP-4023-FR-S-1 Response to Nuclear Power Generation / ATWS Source: Bank KA - 000007 EA2.01 Reactor Trip - Stabilization Ability to determine and interpret the following as they apply to a reactor trip:

Decreasing power level, from available indications RO - 4.1 SRO - 4.3 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - The candidate must determine the decreasing power indications DC COOK OPS Page: 1 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO that show that the reactor has been shut down.

Associated objective(s):

(RO-C-EOP04-E18) For each of the FR-S series procedures identify the Procedure Transitions.

DC COOK OPS Page: 2 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 2 ID: NRCAUDIT07-0129 Points: 1.00 Which ONE of the following describes the procedural actions in response to addressing a leaking Pressurizer (PRZ) PORV?

A. 1. PORV block valves are closed one at a time.

2. Temperature on the tailpipe is monitored by the operator.

3. Leakage is determined by a lowering of tailpipe temperature after each PORV block valve is closed.

B. 1. PORV block valves are closed one at a time.

2. Temperature in the Pressurizer Relief Tank (PRT) is monitored by the operator.

3. Leakage is determined by a lowering PRT temperature after each PORV block valve is closed.

C. 1. All PORV block valves are initially closed to lower tailpipe temperature.

2. One PORV block valve is opened at a time.

3. Leakage is determined by a rise in tailpipe temperature after each PORV block valve is re-opened.

D. 1. All PORV block valves are initially closed to stabilize Pressurizer Relief Tank (PRT) temperature.

2. One PORV block valve is opened at a time.

3. Leakage is determined by a rise in PRT temperature after each PORV block valve is re-opened.

Answer: C Answer Explanation:

A. Incorrect - All Block Valves are initially closed, and leakage is determined by a rise in tailpipe temperature after each PORV block valve is re-opened.

B. Incorrect - All Block Valves are initially closed. PRT conditions are checked but not used to determine leaky valves. Leakage is determined by a rise in tailpipe temperature after each PORV block valve is re-opened.

C. Correct - The procedure requires that all PORV Block Valves be initially closed.

Once tailpipe temperature is lowering, the block valves are opened 1 at a time to check for a rise in tailpipe temperature.

D. Incorrect - PRT conditions are checked but not used to determine leaky valves.

Question ID (Status) NRCAUDIT07-0129 (Active)

DC COOK OPS Page: 3 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: H/3 Comments:

Reference:

OHP-4022-002-009, Leaking Pressurizer Power Operated Relief Valve Source: Bank Previous NRC exam: 2014 KA - 000008 2.1.7 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Conduct of Operations Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

RO - 4.4 SRO - 4.7 CFR - 41.5 / 43.5 / 45.12 / 45.13 KA Justification - The candidate must determine the correct method and indication used to assess plant conditions to determine if the PORV has been isolated.

Associated objective(s):

(RO-C-0020920412-E3) Given a set of plant conditions which includes a Leaking Pressurizer Power Operated Relief Valve, explain the procedural mitigation strategy for a Leaking Pressurizer Power Operated Relief Valve, in accordance with applicable Annunciator Response Procedures and OHP-4022-002-009 Leaking Pressurizer Power Operated Relief Valve.

DC COOK OPS Page: 4 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 3 ID: RO32NRC-003 Points: 1.00 A LOCA occurred on Unit 1.

Core Exit Thermocouples are 300°F and slowly lowering.

RCS pressure is slightly higher than containment pressure.

SG pressures are 675 psig and slowly lowering.

(1) Heat removal is occurring by (1) .

(2) Heat removal will be continued by maintaining (2) .

A. (1) break flow ONLY; (2) at least the minimum necessary ECCS injection flow B. (1) break flow ONLY; (2) SG NR level using aux feed, and reducing steam pressure using SG PORVs or Steam Dumps to the condenser if available C. (1) reflux cooling ONLY; (2) SG NR level using aux feed, and reducing steam pressure using SG PORVs or Steam Dumps to the condenser if available D. (1) reflux cooling ONLY; (2) at least the minimum necessary ECCS injection flow Answer: A Answer Explanation:

The maximum containment pressure under any circumstances during a LBLOCA is expected to be less than 50 PSIG. Under these conditions, the RCS has completely blown down, and ECCS is expected to be cooling the core then spilling out of the break.

Two-phase cooling will only occur if the SGs are still a heat sink. In this case the SG saturation temperatures are about 200°F greater than the Core exit temperatures, such that the SGs are actually a heat source. The only operator response to aid in core cooling is to continue to follow the ERGs and maintain core cooling using ECCS.

A. Correct - See explanation above.

B. Incorrect - Plausible if the applicant has a misconception about the conditions given, and procedural guidance as to the basis and timing for reducing SG pressure following a LBLOCA.

C. Incorrect - Plausible if the applicant has a misconception of the mechanisms for two-DC COOK OPS Page: 5 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO phase cooling. Could be correct if reflux cooling were occurring.

D. Incorrect - Plausible if the applicant has a misconception of the mechanisms for two-phase cooling, or core cooling during a LBLOCA vs. SBLOCA.

Question ID (Status) RO32NRC-003 (Active)

External Topic ID:

Level/Difficulty: H/2 Comments:

Reference:

RO-C-EOP09 pg. 34-41 Source: Bank Previous NRC exam: 2016 Prairie Island K/A 000011 EK1.01 Large Break LOCA Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA:

Natural circulation and cooling, including reflux boiling RO - 4.1 SRO - 4.4 CFR - 41.8 / 41.10/45.3 KA Justification - Candidate must demonstrate knowledge of how natural circulation changes based on the SG conditions during a Large Break LOCA.

Associated objective(s):

(RO-C-EOP02-E8) Describe the methods of RCS heat removal available during post-accident conditions.

DC COOK OPS Page: 6 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 4 ID: 2008NRC-0425A Points: 1.00 A plant startup is in progress with Reactor Power at 12% and rising.

Given the following conditions on RCP #3:

Calculated #2 Seal Leak Rate is 1.6 gpm.

Lower Bearing water temperature is 200°F and rising.

Motor Bearing temperature is 174°F and stable.

Seal Leakoff temperature is 179°F and rising.

Seal Injection Flow is 10 gpm.

Vibrations are at 16 mils and stable.

After commencing a plant shutdown IAW OHP-4021-001-003, Power Reduction the RO reports #1 Seal Leakoff flow is indicating 6.1 gpm and slowly rising Which ONE of the following operator actions MUST be taken based upon these conditions?

A. Manually trip the reactor, enter OHP-4023-E-0, Reactor Trip or Safety Injection, perform immediate actions, and then trip the No. 3 RCP.

B. Initiate reactor shutdown per OHP-4021-001-003, Power Reduction, and trip the No. 3 RCP after the reactor is shutdown.

C. Do NOT trip the reactor. Trip the No. 3 RCP and be in Hot Shutdown in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

D. Do NOT trip the reactor. Trip the No. 3 RCP and close the No. 1 seal leakoff valve.

Answer: A Answer Explanation:

A. Correct - The #1 seal is failing, and a Reactor trip is required, based on step 2 RNO 2.c. #1 Seal leakoff > 6gpm requires a reactor Trip per Step 15. The RCP is tripped and isolated following the Reactor trip.

B. Incorrect - See A above.

C. Incorrect - See A above.

D. Incorrect - See A above.

Question ID (Status) 2008NRC-0425A (Active)

DC COOK OPS Page: 7 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: H/2 Comments:

Reference:

OHP-4022-002-001 Malfunction of A Reactor Coolant Pump Source: Modified KA - 000015 AK2.10 Reactor Coolant Pump (RCP) Malfunctions Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions and the following: RCP indicators and controls RO - 2.8 SRO - 2.8 CFR - 41.7 / 45.7 KA Justification - The candidate must determine the correct action and failure based on the RCP indications.

Associated objective(s):

(RO-C-AOP01404012-E3) Given plant conditions explain the procedural mitigation strategy for a Malfunction of a RCP in accordance with plant procedures.

DC COOK OPS Page: 8 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 5 ID: NRCAUDIT07-0779 Points: 1.00 The following plant conditions exist:

The plant has been operating at 100% power for 300 days.

An AEO reports to the control room that charging line containment isolation valve QMO-200 has a significant body to bonnet leak.

The Shift Manager directs the control room crew to isolate letdown and charging using the guidance in OHP-4022-002-020, Excessive Reactor Coolant Leakage, and place excess letdown in service.

Based on these conditions, why is excess letdown required?

A. To maintain constant reactor coolant system inventory.

B. To restore Volume Control Tank hydrogen control capability.

C. To provide a means of purifying reactor coolant by aligning flow through the mixed bed demineralizers.

D. To ensure an adequate cooling water flowrate through the reactor coolant pump seals while charging flow is isolated.

Answer: A Answer Explanation:

Without normal letdown in this situation there is no charging flow path to the regenerative HX, excess letdown is required to maintain PRZ level constant as it is designed to remove 20 gpm. This amount is exactly equal to seal injection flow that leaks into the RCS. Without excess letdown, PRZ level would rise and the plant would eventually trip.

A. Correct - see explanation above.

B. Incorrect - because although the return flow path for excess letdown is to the suction of the charging pumps, there is little impact on the hydrogen capability of the VCT. Plausible since hydrogen is added to the VCT.

C. Incorrect - because the return flow path in on the RCP seal return line which bypasses the demineralizers. Plausible if candidate does not know that return line bypasses demineralizers.

D. Incorrect - because seal injection can be supplied with charging isolated regardless DC COOK OPS Page: 9 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO of status of excess letdown. As long as thermal barrier HXs are supplied by CCW, seal cooling can be provided with no seal injection. Plausible since charging is heated with regen heat exchanger.

Question ID (Status) NRCAUDIT07-0779(Active)

External Topic ID:

Level/Difficulty: F/2 Comments:

Reference:

OHP-4022-002-020 Excessive Reactor Coolant Leakage, RO-C-00300 Chemical Volume Control System Source: Bank KA - 000022 AK3.03 Loss of Reactor Coolant Makeup Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: Performance of lineup to establish excess letdown after determining need RO - 3.1 SRO - 3.3 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - The candidate must determine which is the correct reason for using excess letdown.

Associated objective(s):

(RO-C-00300-E1) Explain the purpose(s) and/or function(s) of the CVCS.

DC COOK OPS Page: 10 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 6 ID: RO32NRC-006 Points: 1.00 The following conditions exist:

Unit 2 is in Mode 4 during cooldown per OHP-4021-001-004, Plant Cooldown from Hot Standby to Cold Shutdown

East RHR Pump is operating with the return aligned to the Normal Cooldown Path to Loops 2 & 3.

RCS temperature is 220°F and stable

RCS pressure is 250 psig and stable

2-IRV-310, East RHR Hx Outlet Valve is 50% open

2-IRV-311, RHR Hx Bypass Valve is 50% open The air supply line to 2-IRV-311, RHR Hx Bypass Valve, breaks, causing a complete loss of Instrument Air to the valve.

Which of the following describes the specific indications that would be used to diagnose this failure?

A. 2-IFI-335 RHR Return flow would RISE. 2-ITR-335 Loop Return Temperature would RISE.

B. 2-IFI-310 RHR HX Outlet flow would LOWER. 2-IPA-310 Pump Discharge Pressure would LOWER.

C. 2-IFI-335 RHR Return flow would LOWER. 2-IPA-310 Pump Discharge Pressure would RISE.

D. 2-IFI-310 RHR HX Outlet flow would RISE. 2-ITR-335 Loop Return Temperature would LOWER.

Answer: A Answer Explanation:

2-IRV-311 fails open on loss of air. This will raise total RHR flow while keeping the flow through the HX about the same since it is throttled with 2-IRV310. 2-IFI 335 indicates total RHR flow back to the RCS. Since the 2-IRV311 has failed open more flow will bypass the HX so the Return temperature will rise (Less cooling). The Pump will see a slight discharge pressure loss.

A. Correct IRV-311 fails open so flow will raise, and more will bypass HX rising temperature.

B. Incorrect IFI -310 does not monitor flow in this flow path so it will not change.

DC COOK OPS Page: 11 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Plausible since RHR HX flow does lower and the pump may see a slight pressure drop but the flow is not indicated on this meter.

C. Incorrect IRV-311 fails open so flow will raise, and more will bypass HX rising temperature. Plausible if 2-IRV-311 failed closed.

D. Incorrect IFI -310 does not monitor flow in this flow path so it will not change and the ITR temperature will rise. Plausible is they are not familiar with the flow path.

Question ID (Status) RO32NRC-006(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

02-OHP-4021-017-002, Placing in Service the RHR System; 02-OHP-4022-064-002, Loss of Control Air Recovery (Att. B-9)

Source: New KA - 000025 AA1.09 Loss of Residual Heat Removal System (RHRS)

Ability to operate and/or monitor the following as they apply to the Loss of Residual Heat Removal System: LPI pump switches, ammeter, discharge pressure gauge, flow meter, and indicators RO - 3.2 SRO - 3.1 CFR - 41.7 / 45.5 / 45.6 KA Justification: Candidate must determine which indications must be used to determine the failure of the RHR System. While this is not a complete loss of RHR is does address the malfunction of components which degrade the system to the point that cooldown control is lost.

Associated objective(s):

(RO-C-01700-E4) Explain how a loss of each will affect the operation of the RHR System:

a. CCW

b. Control air

c. ESF Ventilation

d. Control Power DC COOK OPS Page: 12 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 7 ID: RO32NRC-007 Points: 1.00 Which one of the following describes the limits for the Spent Fuel Pool Cooling System and the ability to supply CCW to the Spent Fuel Pool Heat exchanger following a Unit 1 CCW Miscellaneous Header Break?

A. SFP normal limit of 120F with Analysis limit of 159F. EITHER the North OR South SFP HX can be supplied from Unit 2.

B. SFP normal limit of 120F with Analysis limit of 159F. ONLY the South SFP HX can be supplied from Unit 2.

C. SFP normal limit of 105F with Analysis limit of 120F. EITHER the North OR South SFP HX can be supplied from Unit 2.

D. SFP normal limit of 105F with Analysis limit of 120F. ONLY the South SFP HX can be supplied from Unit 2.

Answer: B Answer Explanation:

A. Incorrect - These are the correct temperature limits, but only the south HX can be fed.

B. Correct - SFP temperature less than 120F (Admin Limit) and 159.5F (Analysis limit),

Nuclear Engineering is informed when temp exceeds 150F and report generated. The North SFP HX is located on the unit 1 miscellaneous header and so cannot be fed if the header is isolated.

C. Incorrect - These are the temperature limits of the CCW System and only the south HX can be fed.

D. Incorrect - These are the temperature limits of the CCW System Question ID (Status) RO32NRC-007(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

OHP-4021-016-003 Component Cooling Water System Operation, 12-OHP-4021-018-002 Placing in Service and Operating the Spent Fuel Pit Cooling and Cleanup System DC COOK OPS Page: 13 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Source: New KA - 000026 AA2.04 Loss of Component Cooling Water (CCW)

Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The normal values and upper limits for the temperatures of the components cooled by CCW RO - 2.5 SRO - 2.9 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification: Candidate must identify the SFP temperature limits to identify the component / alignment that is a concern due to the loss of CCW.

Associated objective(s):

(RO-C-AOP0420412-E2) Explain the required operator actions to stabilize plant conditions after Malfunctions of the CCW System prior to formal procedure implementation in accordance with plant procedures, and standards and expectations for performance.

DC COOK OPS Page: 14 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 8 ID: RO32NRC-008 Points: 1.00 Given the following conditions:

Unit 1 is operating at 100% power.

Pressurizer Pressure Channel #1 has failed and the associated bi-stables have been placed in the tripped condition per 1-OHP-4022-013-009 Pressurizer Pressure Instrument Malfunction.

Which one of the following describes the Reactor Trip and Safety Injection setpoints and the required coincidence of the REMAINING Pressure Channels?

Reactor Trip Safety Injection A. 1/3 < 1950 psig 1/2 < 1775 psig B. 1/2 < 1910 psig 2/3 < 1775 psig C. 1/2 < 1950 psig 2/3 < 1825 psig D. 1/3 < 1910 psig 1/2 < 1825 psig Answer: A Answer Explanation:

The Low Pressure Reactor Trip is 2/4 channels < 1950 psig. The Low Pressure SI is 2/3

< 1775 psig. These numbers move to 1 of remaining 3 and 1 of remaining 2. The knowledge that all the bi-stables are tripped within the procedure is also required.

A. Correct - 1/3 < 1950 and 1/2 < 1775 is the remaining coincidence and correct setpoints.

B. Incorrect - The SI Coincidence and Rx trip setpoint is wrong. This would be the correct answer if NPS-153 Channel 4 had failed.

C. Incorrect - This is the coincidence and setpoint for P-11 (1910 on 2/3). The Second setpoint is the value used during the SI to open and close the CCP ELO valve, but they are driven by only 1 channel.

D. Incorrect - This is the coincidence but the setpoints are for P-11 and the CCP ELO.

Question ID (Status) RO32NRC-008(Active)

External Topic ID:

DC COOK OPS Page: 15 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Level/Difficulty: H/3 Comments:

Reference:

OHP-4022-013-009 Pressurizer Pressure Instrument Malfunction Source: New KA - 000027 2.4.2 Pressurizer Pressure Control (PZR PCS) Malfunction Emergency Procedures/Plan Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

RO - 4.5 SRO - 4.6 CFR - 41.7 / 45.7 / 45.8 KA Justification: This requires the Operator to determine the remaining coincidence and correct setpoints for the Pressurizer Pressure system Reactor Trip and Safety Injections which are entry conditions for E-0.

Associated objective(s):

(RO-C-AOP0330412-E3) Given a set of plant conditions including a Pressurizer Pressure Instrument Malfunction or Controller Failure explain the procedural mitigation strategy for the malfunction in accordance with OHP-4022-013-009, Pressurizer Pressure Instrument Malfunction.

DC COOK OPS Page: 16 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 9 ID: CM-8679 Points: 1.00 Unit 1 was operating at 100% power when:

A condensate system transient caused both Main FW pumps to trip.

The turbine and reactor failed to trip automatically.

The Operator at the Controls performs the immediate actions of 1-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS.

Which ONE of the following describes the bases for these immediate actions in 01-OHP-4023-FR-S.1?

The safeguards systems are designed assuming that the only heat being added to the RCS is from . For an ATWS event with a loss of normal Feedwater, a Turbine trip within 30 seconds will .

A. decay and RCP heat; prevent challenging the PZR PORV's.

B. decay and RCP heat; maintain S/G inventory.

C. 4% fission heat; maintain S/G inventory.

D. 4% fission heat; prevent challenging the PZR PORV's Answer: B Answer Explanation:

The safeguards systems that protect the plant during accidents are designed assuming that only decay heat and pump heat are being added to the RCS. For an ATWS event where a loss of normal feedwater has occurred, analyses have shown that a turbine trip is necessary (within 30 seconds) to maintain SG inventory.

A. Incorrect - Heat input is correct, but reason is incorrect.

B. Correct - Heat input and reason are both correct.

C. Incorrect - Heat input is incorrect, but reason is correct.

D. Incorrect - Heat input and reason are both incorrect.

DC COOK OPS Page: 17 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) CM-8679(Active)

External Topic ID: PROC-4: PROC-4, Procedure Step Basis Level/Difficulty: F/3 Comments:

Reference:

12-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS Step 1-3 Background Source: Bank Previous NRC exam: 2010 KA - 000029 EK1.01 Anticipated Transient Without Scram (ATWS)

Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermo-hydraulics behavior RO - 2.8 SRO - 3.1 CFR - 41.8 / 41.10 / 45.3 KA Justification - Requires the candidate to understand the implications of performing the immediate actions during an ATWS to address the power production, decay heat production and removal and secondary heat sink requirements.

Associated objective(s):

(RO-C-EOP04-E15) For each of the FR-S series procedures discuss the basis or reason for all Steps.

DC COOK OPS Page: 18 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 10 ID: NRC2010-33 Points: 1.00 Operators are performing 2-OHP-4023-ECA-2.1, Uncontrolled Depressurization of All Steam Generators due to a steam leak inside containment along with failure of all SG stop valves to close.

Given the following plant conditions:

Containment pressure is 3 psig.

The crew has taken action to minimize the plant cooldown.

Steam Generator AFW flow indicates 25x103 pph each SG.

T-hots are slowly lowering.

The following alarms are received:

Ann. 213 Drop 5, STEAM GEN #1 WATER LEVEL LOW-LOW Ann. 213 Drop 35, STEAM GEN #2 WATER LEVEL LOW-LOW Ann. 214 Drop 5, STEAM GEN #3 WATER LEVEL LOW-LOW Ann. 214 Drop 35, STEAM GEN #4 WATER LEVEL LOW-LOW Which ONE of the following actions is required in accordance with 2-OHP-4023-ECA-2.1?

A. Adjust AFW flow to 60x103 pph on each Steam Generator until NR level is 28%.

Adjust flow as necessary to stabilize level.

B. Maintain AFW flow at its current value until WR drops to 26%. Adjust AFW flow to maintain WR level > 17%.

C. Maintain AFW flow at its current value. If T-hot starts to rise, raise AFW flow to stabilize RCS temperature.

D. Maintain AFW flow at its current value. If SG levels continue to lower, raise AFW flow to maintain NR levels >13%.

Answer: C Answer Explanation:

A. Incorrect - After throttling to minimize RCS cooldown, even if levels are low, AFW remains throttled until Thot begins to rise. At that point, AFW is throttled just enough to stabilize temperature. Credible because 28% is the lower limit that level is maintained for Adverse Containment. (Containment is NOT Adverse)

B. Incorrect - After throttling to minimize RCS cooldown, even if levels are low, AFW remains throttled until Thot begins to rise. At that point, AFW is throttled DC COOK OPS Page: 19 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO just enough to stabilize temperature. Credible because 50% is the upper limit that level in most of the EOPs.

C. Correct - AFW Flow is maintained at a minimum amount since the levels are low and T-hot is lowering.

D. Incorrect - Flow is maintained per at 25x103 pph per the procedure. If the transition to FR H.1 is reached Step 1 will return the Crew to ECA-2.1. Plausible as Operator should know they have the ability to avoid a Red path on heat sink but avoiding this red path is in violation of ECA 2.1 requirements and the red path addresses the intentional reduction in AFW flow.

Note:

- Level setpoint for SG Low-Low is 22% - TDAFP start setpoint.

- Since this an operator induced reduction of AFW flow, FR-H. 1 actions would not be performed even if the transition was made.

Question ID (Status) NRC2010-33(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

2-OHP-4024-213 & 214 Drops 5 & 35, 2-OHP-4023-ECA-2.1 Uncontrolled Depressurization of All Steam Generators Source: Bank Previous NRC exam: 2010 KA - 00WE12 EK2.1 Uncontrolled Depressurization of all Steam Generators Knowledge of the interrelations between the Uncontrolled Depressurization of all Steam Generators and the following: Components, and functions of control and safety systems, including Instrumentation, signals, interlocks, failure modes, and automatic and manual features RO - 3.4 SRO - 3.7 CFR - 41.7 / 45.7 KA Justification - This requires the candidate to understand and evaluate the components available and the conditions and determine the correct manual actions for the current situation.

Associated objective(s):

(RO-C-EOP07-E13) For E-2, E-1, and ECA-2.1 discuss the basis or reason for all Steps.

DC COOK OPS Page: 20 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 11 ID: NRCAUDIT07-0297A Points: 1.00 Unit 2 is performing 2-OHP-4023-FR-H.1, Response to A Loss of Secondary Heat Sink and is unable to start any AFW pumps.

2-OHP-4022-055-003, Loss of Condensate to Auxiliary Feedwater Pumps is being performed to establish AFW crosstie from Unit 1 Attachment A of 2-OHP-4022-055-003 directs the operators to open 2-FW-129, Unit 2 East MDAFP Discharge to Unit 1 Crosstie Valve, and start the Unit 1 West MDAFP.

Which ONE of the following describes the actions that will be required to feed the Unit 2 SGs aligned as a result of performing the actions in Attachment A?

A. Throttle open 2-FMO-212 and 2-FMO-222 to feed SG 21 and 22 B. Throttle open 2-FMO-232 and 2-FMO-242 to feed SG 23 and 24 C. Throttle open 2-FMO-212 and 2-FMO-242 to feed SG 21 and 24 D. Throttle open 2-FMO-222 and 2-FMO-232 to feed SG 22 and 23 Answer: D Answer Explanation:

The SGs are grouped 21 & 24 and 22 & 23. This is maintained during the crosstie but the train feeding AFW is swapped.

A. Incorrect - SG 21 is not fed in this arrangement.

B. Incorrect - SG 24 is not fed in this arrangement.

C. Incorrect - This would be correct if the Unit 1 East AFW pump was aligned.

D. Correct - The Unit 1 West AFW feeds the Unit 2 East Train Valves.

Question ID (Status) NRCAUDIT07-0297A (Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

2-OHP-4022-055-003, Loss of Condensate to Auxiliary Feedwater Pumps Attachment A Step 2 DC COOK OPS Page: 21 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Source: Modified KA - 000054 AK3.03 Loss of Main Feedwater (MFW)

Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW): Manual control of AFW flow control valves RO - 3.8 SRO - 4.1 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - Question addresses Loss of Heat Sink and the operators knowledge of which and why the AFW valves need to be opened to provide flow to the SGs.

Modified from NRCAUDIT07-297. Changed to restore the West AFW pump from unit 1 which changes correct answer to D vs. C Associated objective(s):

(RO-C-AOP0460412-E2) Given a set of plant conditions including a Loss of Condensate to the Auxiliary Feedwater Pumps, describe the required operator actions to correct, control, or mitigate the plant in accordance with OHP-4022-055-003, Loss of Condensate to Auxiliary Feedwater Pumps.

DC COOK OPS Page: 22 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 12 ID: NRCAUDIT07-0253 Points: 1.00 A 250 VDC station battery is supplying DC bus loads without a battery charger online. If the equipment loaded on the DC bus does NOT change, as the battery is expended the current draw (amps) will:

A. be fairly constant until the design battery capacity (amp-hours) is exhausted and then will rapidly lower.

B. rise steadily until the design battery capacity (amp-hours) is exhausted.

C. lower steadily until the design battery capacity (amp-hours) is exhausted.

D. initially lower until approximately 50% design capacity (amp-hours) had been expended and then rise until the battery has been exhausted.

Answer: B Answer Explanation:

A. Incorrect - The current draw rises. This is a typical response for many design systems.

B. Correct - Recall that Power=Voltage x Current. As the battery discharges the voltage will drop. To maintain a constant power output the current draw must rise.

C. Incorrect - The current draw rises. The effect of lowering voltage on current draw has been reversed.

D. Incorrect - The current draw rises. Incorrect battery theory.

Question ID (Status) NRCAUDIT07-0253(Active)

External Topic ID: 8204: 8204, 250 VDC Distribution Level/Difficulty: H/3 Comments:

Reference:

RO-C-BE01, Basic Electricity Source: Bank KA - 000055 EA1.05 Loss of Offsite and Onsite Power (Station Blackout)

Ability to operate and/or monitor the following as they apply to a Station Blackout:

Battery, when approaching fully discharged DC COOK OPS Page: 23 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO RO - 3.3 SRO - 3.6 CFR - 41.7 / 45.5 / 45.6 KA Justification - Question requires candidate to evaluate response of battery as the charge lowers.

Associated objective(s):

(RO-C-08204-E4) Describe how the 600 Volt AC System provides a support function for the 250 Volt DC system.

DC COOK OPS Page: 24 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 13 ID: NRCAUDIT07-0743 Points: 1.00 The operator incorrectly opens the breaker labeled "7.5 KVA Static Inverter Channel IV" on 250 VDC distribution panel "MCAB". The operator realizes the mistake and immediately recloses the breaker.

Which ONE of the following describes the effect of these actions?

A. The alternate power source to the CRID Inverter will be lost when the breaker is reclosed. The CRID will transfer to the 120 VAC from the Regulating Transformer.

B. The alternate power source to the CRID Inverter will be lost. No automatic action will occur when the breaker is reclosed. The auto transfer lockout must be reset at the inverter.

C. The normal power source to the CRID Inverter will be lost so it will auto transfer to the alternate source. When the breaker is reclosed, it will auto transfer to the normal source.

D. The normal power source to the CRID Inverter will be lost so it will auto transfer to the alternate source. When the breaker is reclosed, the auto transfer lockout must be reset at the inverter.

Answer: C Answer Explanation:

A. Incorrect - The Alternate source will not be lost. The normal DC supply will be restored, and the Inverter will re-transfer to the normal source.

B. Incorrect - The Alternate source will not be lost. The normal DC supply will be restored, and the Inverter will re-transfer to the normal source.

C. Correct - The static transfer switch provides a virtual zero-time transfer to the alternate source in case of inverter failure. Thirty seconds after the static switch transfer event ceases and all system parameters are normal, the static switch automatically re-transfers the load to the inverter, without power interruption.

D. Incorrect - The normal DC supply will be restored when the breaker is closed, and the Inverter will re-transfer to the normal source.

Question ID (Status) NRCAUDIT07-0743(Active)

External Topic ID: 8203: 8203, Instrument Electrical Distribution DC COOK OPS Page: 25 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Level/Difficulty: H/3 Comments:

Reference:

RO-C-08203, Instrumentation Electrical System Rev 10 pg. 27 2-OHP-4024-219 Drop 29, CRID 3 Inverter Abnormal Source: Bank Previous NRC exam: 2006 KA - 000057 AA2.14 Loss of Vital AC Electrical Instrument Bus Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: That substitute power sources have come on-line on a loss of initial AC RO - 3.2 SRO - 3.6 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Question tests ability of the operator to differentiate between normal and alternate sources and the automatic transfers when a source is lost or restored.

Original Question Source: Cook 2006 NRC Exam: Bank 01082C0303-2 Original KA - 063 K4.01 (2.7 / 3.0)

Associated objective(s):

(RO-C-08203-E3) Describe the function and operation of the following Instrumentation Electrical System Major Components:

a. AMSAC Inverter

b. Lighting Transformers

c. Auto Bus Transfer Switches: ABT-1 though ABT-5, 12-TSC-UPS-ABT

d. Inverters

e. Static Switches

f. Emergency Lighting Packs

g. Constant Voltage Transformers DC COOK OPS Page: 26 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 14 ID: RO-C-05200-E9-1 Points: 1.00 Open Reference A Steam Generator Tube Rupture (SGTR) has occurred and 2-OHP-4023-E-3, SGTR procedure, is being performed. The crew has isolated the ruptured Steam Generator and has completed the cooldown to a target temperature of <475.6°F.

The Unit Supervisor directs you to stabilize Reactor Coolant System (RCS)

Temperature and set up Steam Dump Pressure Controller to maintain RCS temperature at approximately 470°F.

Which ONE of the following is the correct Steam Dump Pressure Controller setpoint required to maintain RCS temperature at approximately 470°F?

A. 500 psig B. 514 psig C. 530 psig D. 543 psig Answer: A Answer Explanation:

A. Correct - 514.67 psia is Psat for 470°F. The Controller would need to be set at 500 psig. (514.67 -14.7 = 500).

B. Incorrect - 514.67 psia is Psat for 470°F C. Incorrect - 530 psig is derived from adding 14.7 psi to 515 psia.

D. Incorrect - 543 psia is Psat for 475.6°F. (Absolute not PSIG)

Question ID (Status) RO32NRC-014(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

REFERENCE:

(Provided) Steam Tables, SOD-05200-001 DC COOK OPS Page: 27 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Source: Bank NRC 2012 KA - 000038 K1.01 Steam Generator Tube Rupture Knowledge of the operational implications of the following concepts as they apply to the SGTR: Use of Steam Tables RO - 3.1 SRO - 3.4 CFR - 41.8 / 41.10 / 45.3 EPE 038 EK1.01 KA Justification - Requires operator knowledge of how to use the steam tables to determine the required steam dump controller setpoint.

Associated objective(s):

(RO-C-05200-E9) Given Steam Tables, determine the RCS temperature and SG pressure associated with the following:

a. SG PORV opening setpoint

b. No-load Tavg

c. Various temperature-pressure relationships DC COOK OPS Page: 28 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 15 ID: 2007-0406 Points: 1.00 Given the following conditions:

Unit 1 is at 100% power.

The crew has entered 1-OHP-4022-019-001, ESW System Loss/Rupture, due to a large leak just downstream of the U1 East ESW Pump Discharge Valve, WMO-701.

The control room crew has closed WMO-707, Unit 2 ESW Header Crosstie, as directed by the procedure.

The 1E ESW pump is NOT running.

Which ONE of the following sets of components have completely lost ESW flow capability due to these actions?

A. DG1CD Cooling Water Supply East MDAFP Emergency Suction North Control Room Air Conditioning ESW Supply East CCW Hx Cooling Water Supply B. DG1AB Cooling Water Supply West MDAFP Emergency Suction South Control Room Air Conditioning ESW Supply West CCW Hx Cooling Water Supply C. West MDAFP Emergency Suction East MDAFP Emergency Suction North Control Room Air Conditioning ESW Supply East CCW Hx Cooling Water Supply D. TDAFP Emergency Suction West MDAFP Emergency Suction South Control Room Air Conditioning ESW Supply West CCW Hx Cooling Water Supply Answer: C Answer Explanation:

A. Incorrect - DG1CD normal supply is lost, but the alternate supply is still available.

B. Incorrect - DG1AB alternate lost but the normal supply is still available. N CRAC and E CCW Hx are affected.

DC COOK OPS Page: 29 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO C. Correct - East and West MDAFPs, N CRAC, and E CCW Hx, are supplied by the 1E ESW Pump Header.

D. Incorrect - TDAFP is supplied by opposite train. N CRAC and E CCW Hx are affected.

Question ID (Status) 2007-0406(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

SOD-01900-001, 12-OHP-4021-019-001 Operation of the Essential Service Water System Line-Up Sheet #1 Source: Bank Previous NRC exam: 2014 KA - 000062 AA1.07 Loss of Nuclear Service Water Ability to operate and/or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): Flow rates to the components and systems that are serviced by the SWS, interactions among the components RO/SRO Value - (2.9 / 3.0) CFR - 41.7 / 45.5 / 45.6 KA Justification - Question requires candidate to monitor which components have lost flow due to the isolation of the ESW header.

Associated objective(s):

(RO-C-AOP0590412-E1) Given an ESW System Loss or Rupture, identify the event and predict the response of the plant to ESW System Loss or Rupture, including final plant configuration.

DC COOK OPS Page: 30 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 16 ID: RO32NRC-016 Points: 1.00 1-OHP-4023-E-3, Steam Generator Tube Rupture is being implemented and the Crew has reset SI and Phase A Containment Isolation and is restoring Control Air to Containment per step 12.

After opening the Control Air to Containment isolation valves, the Containment Ctrl Air Header Pressure Low annunciator (Panel 104, Drop 33) is still LIT.

Which one of the following describes the required actions and the concern with this alarm being lit?

A. Close the Control Air Isolation valves. RCS Depressurization will need to be performed using the Pressurizer PORVs.

B. Close the Control Air Isolation valves. Normal charging flow will NOT be able to be restored so BIT flow will need to be throttled.

C. Start an additional Control Air Compressor to restore air pressure. Wait until Control Air Pressure has been restored to containment to perform the RCS depressurization.

D. Start an additional Control Air Compressor to restore air pressure. If Control Air Pressure to Containment is not restored, transition to 1-OHP-4023-ECA-3.3, SGTR Without Pressurizer Pressure Control.

Answer: A Answer Explanation:

A. Correct - The Containment Air pressure check is performed whenever Air is restored to Containment to see if air leakage into containment is suspected. If the alarm is lit the isolation valves are reclosed. Normal spray will be unavailable but 2 PORVs have backup supplies.

B. Incorrect - Charging flow will be available, but letdown flow will be lost is the air is isolated.

C. Incorrect - The Control air compressor would have been started prior to opening the air valves is system pressure was low. The procedure doesn't wait for air pressure prior to continuing (it does wait for the cooldown to complete.)

D. Incorrect - The Control air compressor would have been started prior to opening the air valves is system pressure was low. The PORVs have a backup supply DC COOK OPS Page: 31 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO and so a transition should not be required at this time.

Question ID (Status) RO32NRC-016(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

1-OHP-4023-E-3 Steam Generator Tube Rupture Step 12 and

Background

Source: New KA - 000065 AK3.08 Loss of Instrument Air Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Actions contained in EOP for loss of instrument air RO - 3.7 SRO - 3.9 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification: Question requires candidate knowledge of the reason for the AOP/EOP step for checking the containment air pressure alarm and the consequences if the alarm is lit.

Associated objective(s):

(RO-C-AOP0170412-E3) Given a set of plant conditions and the occurrence of an abnormal event, without use of references, explain the procedural mitigation strategy for a Steam Generator Tube Leak in accordance with plant procedures, and standards and expectations for performance.

DC COOK OPS Page: 32 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 17 ID: RO32NRC-017 Points: 1.00 Given the following sequence of events:

Unit 1 was at 100% power.

ESF bus T11D is de-energized due to a bus fault.

Subsequently:

A Manual reactor trip and SI occurred due to a pressurizer safety valve stuck partially open.

The CR receives a report that a large amount of water is leaking in the U-1 Aux building.

The crew transitions to OHP-4023-ECA-1.2, LOCA OUTSIDE CONTAINMENT, with ALL B train ECCS pumps still operating.

The following indications are CURRENTLY noted:

RCS wide range pressure is 1700 psig and stable.

West RHR discharge flow is 0 gpm.

South SI pump discharge flow is 200 gpm.

CCP flow is 300 gpm.

The leak in the Aux Building can be reduced by closing...

A. ICM-311, East RHR to RC Loops 1&4 Hot Legs Containment Isolation Valve.

B. QRV-251, CVCS Centrifugal Charging Pumps Discharge Flow Control Valve C. ICM-111, RHR TO COLD LEGS ISOL VLV.

D. ICM-265, South SI Pump Discharge Containment ISOL VLV.

Answer: D DC COOK OPS Page: 33 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Answer Explanation:

Shut off head for the ECCS pumps are as follows: RHR-200 psid, SI-1500 psid, CCP-2600 psid. At the current RCS pressure of 1700 psid the only pump that should have indicated flow is the CCP pump. Therefore, if the SI pump indicates 200 gpm and assuming the leak is RWST water, it is logical to conclude that the leak is somewhere on the SI pump discharge line. The bus T11D fault is in the stem to limit ECCS flows to one train for more accurate assessment capabilities of the individual pumps.

A. Incorrect - RHR pump flow is normal for current RCS pressure.

B. Incorrect - CCP pump flow is normal for current RCS pressure.

C. Incorrect - RHR pump flow is normal for current RCS pressure.

D. Correct - SI pump flow is abnormal for current RCS pressure.

Question ID (Status) RO32NRC-017(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Source: Bank Previous NRC exam: 2016 Braidwood K/A: W/E04 LOCA Outside Containment EK2.2 Knowledge of the interrelations between the LOCA Outside Containment and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility RO 3.8 SRO 4.0 CFR - 41.7 / 45.7 KA Justification: The question meets the K/A because the candidate must understand the system operations and their interrelation with the leak flow and expected flows.

Associated objective(s):

(RO-C-EOP09-E36) For each of the E-1 Series procedures, discuss the basis or reason for all Steps.

DC COOK OPS Page: 34 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 18 ID: RO32NRC-018 Points: 1.00 A LOCA is in progress and the crew has entered OHP-4023-ECA-1-1, Loss of Emergency Coolant Recirculation from OHP-4023ECA-1.2, LOCA Outside Containment when the leak could not be isolated.

The following conditions exist:

RWST level is 18% and lowering.

East and West CCPs, North and South SI & East and West RHR pumps are running.

RCS Pressure is 340 psig.

The crew has initiated a RCS Cooldown to Cold Shutdown at Step 9. While establishing only One Train of ECCS pumps in service the BOP reports the RWST level is 10.9%

and lowering.

What actions are required based on the BOP report?

A. Trip all running ECCS pumps as RWST level is below 11%

B. Trip the running SI and CCPs ONLY as RWST is below 11%.

C. Trip the running RHR pumps as RWST level is below 11% and trip the running SI and CCPs when RWST is below 7%.

D. Trip the running SI and CCPs pumps as RWST level is below 11% and trip the running RHR pumps when RWST is below 7%.

Answer: C Answer Explanation:

OHP-4023-ECA-1.1 fold out page ECCS / CTS Pump trip Criteria contains the following directions:

IF RWST level lowers to less than 11%, THEN trip any running RHR/CTS pump aligned to the RWST.

IF RWST level lowers to less than 7%, THEN trip CCPs/SI pumps aligned to the RWST.

A. Incorrect -Only the RHR pumps would be tripped when RWST is at 11%. See above.

B. Incorrect - The CCP and SI pumps are tripped when RWST is at 7%. See above DC COOK OPS Page: 35 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO C. Correct - See above D. Incorrect - The RHR pumps would be tripped when RWST is at 11%. The CCP and SI pumps are tripped when RWST is at 7%. See above.

Question ID (Status) RO32NRC-018(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

OHP-4023-ECA-1-1, Loss of Emergency Coolant Recirculation Source: New KA - 00WE11 EK1.1 Loss of Emergency Coolant Recirculation Knowledge of the operational implications of the following concepts as they apply to the Loss of Emergency Coolant Recirculation:

Components, capacity, and function of emergency systems RO - 3.7 SRO - 4.0 CFR - 41.8 / 41.10 / 45.3 KA Justification: Question requires candidate to know the limits of the ECCS component suction source and required action.

Associated objective(s):

(RO-C-EOP09-E36) For each of the E-1 Series procedures, discuss the basis or reason for all Steps.

DC COOK OPS Page: 36 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 19 ID: CM-8482 Points: 1.00 Unit 2 has been at 60% power for 3 days.

The following conditions exist:

Control Bank D is at 174 steps withdrawn.

Rod Control is in automatic.

RCS Auctioneered High Tavg is 564o F and steady.

Control Bank D rods start withdrawing at 8 steps per minute. With Control Bank D at 177 steps, Rod Control is placed in Manual, and all rod motion stops.

Which ONE of the following describes why control rods will be inserted in manual to 174 steps withdrawn in accordance with 2-OHP-4022-012-003, Continuous Control Bank Movement?

A. To prevent reduced charging flow.

B. To restore Tavg to programmed band.

C. To prevent exceeding QPTR Limits.

D. To restore RCS pressure to normal.

Answer: B Answer Explanation:

A. Incorrect - Plausible because with Tavg higher, programmed pressurizer level will be higher. This would cause charging flow to increase, though, rather than lower.

B. Correct - Returning rods to prior position will Correct Tavg-Tref difference as directed by procedure.

C. Incorrect - Rod movement in banks affects axial power not radial. Plausible if student confuses the two power spectra.

D. Incorrect - Rod insertion will reduce temperature and hence pressure, but this is not the reason given in the procedure. Pressure will still be controlled via automatic pressure control. Plausible for student to select this choice based on parameter response.

DC COOK OPS Page: 37 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) CM-8482(Active)

External Topic ID:

Level/Difficulty: H/2 Comments:

Reference:

02-OHP-4022-012-003, Continuous Control Bank Movement step 3 Source: Bank Previous NRC exam: 2004 KA: APE001 AK3.01 Continuous Rod Withdrawal Knowledge of the reasons for the following responses as they apply to the Continuous Rod Withdrawal: Manually driving rods into position that existed before start of casualty RO-3.2 SRO-3.6 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - Question requires candidate to know the reason rods are returned to the pre event position.

Associated objective(s):

(RO-C-AOP0200412-E3) Given a set of plant conditions including continuous control bank movement, describe the procedure transition to the proper mitigating procedure, in accordance with 1/2-OHP-4022-012-003, Continuous Control Bank Movement.

DC COOK OPS Page: 38 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 20 ID: CM-2780A Points: 1.00 Given the following conditions:

The unit is at EOL with an RCS boron concentration of 212 ppm.

After having been at 80% power for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reactor power was raised to 90% about 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ago.

Slow and intermittent boration is being used to minimize rod motion while compensating for the related changes in xenon concentration.

The rod control system is in automatic.

A rod control URGENT FAILURE alarm is received, which is followed shortly by a TAVG LO TAVG < TREF DEVIATION alarm.

Which ONE of the following methods will successfully clear the TAVG LO TAVG <

TREF DEVIATION alarm?

A. Transfer rod control to manual and withdraw rods.

B. Allow rods to auto withdraw when a 1.5 °F mismatch occurs.

C. Lower Main Turbine load.

D. Raise the boration rate.

Answer: C Answer Explanation:

A. Incorrect - Rod motion in Manual is inhibited with an URGENT FAILURE. Plausible if student believes manual rod control can occur.

B. Incorrect - Automatic rod motion is inhibited with an URGENT FAILURE. Plausible if student believes automatic rod motion can occur.

C. Correct - Tavg is low, most likely because the boration rate was too high for the xenon reduction rate or because the turbine loading rate was too high. Lowering turbine load will raise Tavg.

D. Incorrect - Xenon is still reducing, adding positive reactivity, following the power change. By raising boration, this adds negative reactivity and Tavg will lower faster.

Question ID (Status) CM-2780A (Active)

External Topic ID: 1200-2: 1200-2, Rod Control - Control DC COOK OPS Page: 39 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Level/Difficulty: H/3 Comments:

Reference:

OHP-4024-110 Drop 26 Rod Control Urgent Failure Source: Bank KA - 000005 AA1.04 Inoperable/Stuck Control Rod Ability to operate and/or monitor the following as they apply to the Inoperable/Stuck Control Rod: Reactor and turbine power RO - 3.9 SRO - 3.9 CFR - 41.7 / 45.5 / 45.6 KA Justification - Question requires candidate to determine the effects of reactor power and turbine power on the temperature following a stuck rod and select the correct operation to restore.

Associated objective(s):

(RO-C-AOP0240412-E2) Given a set of plant conditions and the occurrence of an abnormal event, without use of references, explain the required operator actions to stabilize plant conditions after Dropped Rod, Misaligned Rod, or RPI Failure prior to formal procedure implementation in accordance with plant procedures, and standards and expectations for performance.

DC COOK OPS Page: 40 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 21 ID: RO26-0037 Points: 1.00 Given the following plant conditions:

From 100% power, Unit 2 is being reduced to 50% to remove a Main Feedwater Pump from service

Pressurizer Level Control is selected to Channel 1-2

The Pressurizer Level Master Controller setpoint fails at its current value

The load reduction is initiated Which ONE of the following describes the impact of this failure and the action(s) required to address the failure?

A. Actual pressurizer level will be higher than program level for the power level as load is decreased. Change the pressurizer level channel input to the master controller.

B. Actual pressurizer level will be higher than program level for the power level as load is decreased. Take manual control of Charging flow.

C. Actual pressurizer level will be lower than program level for the power level as load is decreased. Take manual control of Charging flow.

D. Actual pressurizer level will be lower than program level for the power level as load is decreased. Change the pressurizer level channel input to the master controller.

Answer: B Answer Explanation:

A. Incorrect - Plausible because this action could remedy a controlling channel failure.

Incorrect because changing input will not affect the controller because the setpoint has failed, not the input.

B. Correct - With the setpoint failed erroneously high, as power lowers, the PRZ level control system will attempt to maintain level higher than the programmed level (program level lowers as Tave lowers during power reduction).

C. Incorrect - Plausible if student does not correctly correlate the down power to a lowering of Tavg and therefore a lowering of Pressurizer Level Setpoint.

Incorrect because charging flow must be reduced because level will be artificially high.

D. Incorrect - Plausible if student does not correctly correlate the down power to a lowering of Tavg and therefore a lowering of Pressurizer Level Setpoint.

DC COOK OPS Page: 41 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Incorrect because changing input will not affect the controller because the setpoint has failed.

Question ID (Status) RO26-0037(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

2-OHP-4022-013-010, Pressurizer Level Instrument Malfunction, SOD-00202-003 Source: Bank KA - 000028 AA2.08 Pressurizer (PZR) Level Control Malfunction Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: PZR level as a function of power level RO - 3.1 SRO - 3.5 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Justification - Question requires candidate to determine how pressurizer level changes with regards to power and the actions required to address the failure.

Associated objective(s):

(RO-C-AOP0340412-E1) Given a set of plant conditions including a Pressurizer Level Instrument Malfunction or Controller failure, predict the plant response with no operator intervention, without error.

DC COOK OPS Page: 42 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 22 ID: RO32NRC-022 Points: 1.00 Given the following plant conditions on Unit 2:

The unit is in Mode 6 with refueling activities in progress.

The SRO-CA reports that the refueling water level is lowering and he is performing actions and requests the Control Room enter 12-OHP-4022-018-002, Loss of Refueling Water Level During Refueling Operations - Local Actions Which ONE of the following is an action required by the procedure and the reason for it?

A. Verify Containment Purge is in service to limit the airborne radiation levels as water level lowers.

B. Verify Containment Purge is in service to assist in closing the Airlock doors.

C. Verify Containment Purge is NOT in service to prevent a safety hazard as Airlock doors are closed.

D. Verify Containment Purge is NOT in service to minimize the spread of contamination outside of containment.

Answer: C Answer Explanation:

A. Incorrect - Containment Purge is verified removed from service in OHP-4022-018-002 to prevent a safety hazard as Airlock doors are closed.

B. Incorrect - Containment Purge is verified removed from service the reason is prevent a safety hazard as Airlock doors are closed.

C. Correct - Containment Purge is verified removed from service in OHP-4022-018-002 to prevent a safety hazard as Airlock doors are closed.

D. Incorrect - Action is Correct but reason is to prevent a safety hazard as Airlock doors are closed.

Question ID (Status) RO32NRC-022(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

DC COOK OPS Page: 43 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

12-OHP-4022-018-002, Loss of Refueling Water Level During Refueling Operations - Local Actions Note prior to and Step 3 Source: New KA - 000036 2.1.20 Fuel Handling Incidents Conduct of Operations Ability to interpret and execute procedure steps.

RO - 4.6 SRO - 4.6 CFR - 41.10 / 43.5 / 45.12 KA Justification - Question tests candidates knowledge of actions required (to limit personnel hazard) per the procedure for an accident with irradiated fuel (loss of refueling water level).

Associated objective(s):

(RO-C-AOP0130412-E2) Given a set of plant conditions including a loss of Refueling Cavity Water level, explain the required operator actions to stabilize plant conditions, in accordance with 1/2-OHP-4022-002-006, Loss of Refueling Level during Refueling Operations.

DC COOK OPS Page: 44 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 23 ID: NRC2010-14 Points: 1.00 Given the following conditions:

An accidental spill of the Monitor Tank has occurred in the Aux Building.

Radiation levels in the area of the spill are 40 mrem per hour at 30 cm.

Contamination levels based on smear on the floor around the tank are 1.2 x104 dpm/100 cm2 beta-gamma.

Which ONE of the following describes how the area will be posted in accordance with PMI-6010, Radiation Protection Plan?

A. Radiation Area ONLY.

B. Contamination Area ONLY.

C. Radiation Area AND Contamination Area.

D. High Radiation Area AND Contamination Area.

Answer: C Answer Explanation:

A. Incorrect - Greater than >1000 dpm /100 cm2 is a contamination area. Plausible because this answer is only partially correct.

B. Incorrect - >5 mrem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and <100 mrem / hr is a radiation area. Plausible because this answer is only partially correct.

C. Correct - This area should be posted as a radiation area (>5 mrem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and

<100 mrem / hr) and a contamination area (>1000 dpm /100 cm2).

D. Incorrect - Greater than 100 mrem / hr is a high radiation area. Area is <100 mrem/hr.

Question ID (Status) NRC2010-14(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

PMI-6010, Section 4.7, RO-C-RP02 Source: Bank DC COOK OPS Page: 45 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Previous NRC exam: 2010 KA - 000059 AK1.01 Accidental Liquid Radwaste Release Knowledge of the operational implications of the following concepts as they apply to Accidental Liquid Radwaste Release: Types of radiation, their units of intensity and the location of the sources of radiation in a nuclear power plant RO - 2.7 SRO - 3.1 CFR - 41.8 / 41.10 / 45.3 KA Justification - Requires knowledge of the posting requirements for an accidental release (spill of Monitor Tank contents) in the Aux Building. In addition, requires knowledge of the units and acceptable levels of radiation/contamination for these conditions.

Original Question # - NRC 2010 Exam CALLAWAY2007-59 Original Question KA - 000059 AK1.02 Associated objective(s):

DC COOK OPS Page: 46 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 24 ID: NRCAUDIT07-0542 Points: 1.00 Which one of the following correctly describes the Steam Generator PORV Monitors (MRA-1600, 1700, 2600, 2700)?

Two Radiation Monitors per plant Unit with (1) detectors.

They provide accurate indication and provide valid alarms (2) .

A. (1) 4 (2) only when SG PORVs are Open.

B. (1) 4 (2) when SG PORVs are Open or Closed.

C. (1) 2 (2) only when SG PORVs are Open.

D. (1) 2 (2) when SG PORVs are Open or Closed.

Answer: A Answer Explanation:

A. Correct - Steam Generator PORV Monitors (MRA-1600, 1700, 2600, 2700) monitors each UNITs four secondary main steam line PORV line relief valve headers downstream of the PORVs. Monitors will not provide accurate Steam line readings when SG PORVs are closed.

B. Incorrect - Monitors will not provide accurate Steam line readings when SG PORVs are closed.

C. Incorrect - 2 Monitors per UNIT but they are fed from 4 detectors.

D. Incorrect - 2 Monitors per UNIT but they are fed from 4 detectors. Monitors will not provide accurate Steam line readings when SG PORVs are closed.

Question ID (Status) NRCAUDIT07-0542(Active)

External Topic ID:

Level/Difficulty: F/2 Comments:

Reference:

RO-C-01350 Radiation Monitoring System DC COOK OPS Page: 47 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Source: Bank KA - 000061 AK2.01 Area Radiation Monitoring (ARM) System Alarms Knowledge of the interrelations between the Area Radiation Monitoring (ARM)

System Alarms and the following: Detectors at each ARM system location RO - 2.5 SRO - 2.6 CFR - 41.7 / 45.7 KA Justification - Question requires operator knowledge of how the main steam area monitors are arranged and their location with respect to steam flow.

Associated objective(s):

(RO-C-01350-E4) Describe the function of the following Radiation Monitoring System Monitors including any automatic actions that occur on a high alarm:

a. 1-MRA-1600/1700 & 2-MRA-2600/2700, Steam Generator PORV Monitors DC COOK OPS Page: 48 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 25 ID: RO32NRC-025 Points: 1.00 The Control Room has been evacuated due to a fire. The crew has successfully implemented the first 15 steps of 2-OHP-4025-001-001 Emergency Remote Shutdown prior to leaving the control room.

Which ONE of the following describes the position the SG PORV is placed in and why?

A. Steam Generator PORVS are verified to be in Auto with the setpoint adjusted to 1005 psig to maintain Tavg at 547o F.

B. Steam Generator PORVS are placed in Auto with the setpoint adjusted to 1025 psig to ensure even SG cooling (to prevent an SI).

C. Steam Generator PORVS are placed in Manual with a zero demand to prevent spurious opening.

D. Steam Generator PORVS are placed in Manual with a 10% demand to begin an RCS Cooldown.

Answer: C Answer Explanation:

A. Incorrect - SG PORVs are placed in Manual. The 1005 is the setpoint used on the Steam Dumps to maintain 547F.

B. Incorrect - SG PORVs are placed in Manual. 1025 is the normal PORV setpoint and even cooling is a concern during PORV operation outside of the control room.

C. Correct - SG PORVs are placed in Manual with zero demand to prevent a fire short from causing the PORVs to open.

D. Incorrect - A cooldown will be started with local PORV operation but that is not how they are left with the first 15 steps of 4025-001-001.

Question ID (Status) RO32NRC-025(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

OHP-4025-001-001 Emergency Remote Shutdown, step 14 DC COOK OPS Page: 49 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Source: New KA - 000067 AK3.04 Plant Fire on Site Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site: Actions contained in EOP for plant fire on site RO - 3.3 SRO - 4.1 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - Question requires candidate knowledge of the position the SG PORV is placed in and why.

Associated objective(s):

(RO-C-EC02-E4) IDENTIFY actions necessary to stabilize plant conditions following control room evacuation.

DC COOK OPS Page: 50 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO NOTE: The words - be taken in answer choice B (blue italicized text) were found to be missing during the exam administration. All applicants were informed to insert the missing words into answer choice B as shown below.

26 ID: RO-C-EOP-E10-1 Points: 1.00 Unit 1 reactor has been manually tripped due to a secondary system malfunction. 1-OHP-4023-E-0 has been performed and a transition made to 1-OHP-4023-ES-0.1, Reactor Trip Response. The STA has identified a YELLOW path on the Heat Sink Status Tree for steam generator level.

The crew has entered 1-OHP-4023-FR-H.3, Response to Steam Generator High Level.

Given the following plant conditions:

Steam Generator #13 pressure is being maintained by cycling of the Steam Generator Safety Valves

Steam Generator #13 NR Level - 96%

Which ONE of the following is the action relative to steam release for Steam Generator #13?

A. Actions to terminate the steam release should be taken to prevent excessive RCS cooldown and depressurization that could potentially cause an unnecessary SI.

B. Actions to terminate the steam release should be taken to prevent two phase flow and water hammer that could potentially damage pipes and valves.

C. Steam may be released via the SG PORV ONLY since narrow range level is greater than 67%.

D. Steam may be released without restriction since narrow range level has been adequately established.

Answer: B Answer Explanation:

A. Incorrect - The higher level in the SG should have little effect on the RCS pressure drop. An SI would not be expected from a single SG steam release.

B. Correct - As discussed in 1-OHP-4023-FR-H.3 Step 1, with a high SG level, steam should not be released until the steam lines can be evaluated.

C. Incorrect - Steam release through the SG PORV could still lead to water hammer.

D. Incorrect - At this high of level, steam should not be released.

DC COOK OPS Page: 51 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) RO-C-EOP-E10-1(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

1-OHP-4023-FR-H.2, Response to Steam Generator Overpressure; 1-OHP-4023-FR-H.3, Response to Steam Generator High Level; 12-OHP-4023-FR-H.2 Background Document; 12-OHP-4023-FR-H.3 Background Document Source: Bank Previous NRC exam: 2004 KA - 00WE13 EA1.3 Steam Generator Overpressure Ability to operate and/or monitor the following as they apply to the Steam Generator Overpressure: Desired operating results during abnormal and emergency situations KA Justification - Question requires candidate knowledge of the concerns of water in and around the steam lines and the limitations to achieving the desired results.

Question tests RO knowledge as it can be answered by knowing the overall mitigating strategy of a procedure.

Associated objective(s):

(RO-C-EOP11-E10) For each of the FR-H series procedures discuss the basis or reason for all Steps.

NOTE: The following is a list of steps that are considered essential to understanding the purpose and overall strategy contained in the FR-H series procedures.

FR-H.2

Identifying affected SG

Checking affected SGs NR level

Trying to dump steam from the affected SG

Checking RCS hot leg temperatures (RO-C-EOP11-E11) For each of the FR-H series procedures discuss the basis or reason for all Cautions.

DC COOK OPS Page: 52 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 27 ID: CM-39624 Points: 1.00 Given the following events and conditions:

Unit 1 is responding to a steam break inside containment All systems operate as designed Narrow range Steam Generator (SG) level is 15% for each intact SG AFW flow is 40 X 103 lbm/hr to each intact SG All Reactor Coolant Pumps (RCP) were tripped The crew entered OHP-4023-FR-P-1, Response to Imminent Pressurized Thermal Shock Condition, due to low temperature:

RCS temperature is now stable RCS pressure is stable with only the control group of pressurizer heaters energized Letdown has been restored The crew has determined that a one-hour soak is required Which one of the following evolutions could be performed by the crew in the next hour while continuing through the procedures?

A. Start #13 RCP.

B. Place auxiliary spray in service.

C. Raise AFW flow to one intact SG to raise NR level to >25%.

D. Commence a 25°F/hour cooldown to Mode 5.

Answer: B Answer Explanation:

A. Incorrect - Starting a RCP will cause a pressure transient and could cause further cooldown.

B. Correct - Any actions that will NOT cause either a cooldown or a pressure rise and are specified by any other procedure in effect are permitted during this soak period.

C. Incorrect - Increases cooldown stressing the vessel.

D. Incorrect - Cooldown is NOT allowed.

Question ID (Status) CM-39624(Active)

DC COOK OPS Page: 53 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: H/3 Comments:

Reference:

1-OHP-4023-FR-P-1, Response to Imminent Pressurized Thermal Shock Condition, step 26 Source: Bank Previous NRC exam: 2012 KA - 00WE08 EA2.2 Pressurized Thermal Shock Ability to determine and interpret the following as they apply to the Pressurized Thermal Shock: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments RO - 3.5 SRO - 4.1 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Questions requires candidate knowledge of the procedural steps and impact on the associated cooldown restriction.

Original Question # - NRC EXAM 2012, RO22 AUDIT- BOTH-78, CM-39624 Original Question KA - WE 08 EK 3.3 Associated objective(s):

(RO-C-EOP12-E28) For each of the FR-P series procedures, identify the Major Action Categories and discuss the bases for each.

DC COOK OPS Page: 54 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 28 ID: NRCAUDIT07-0977 Points: 1.00 Given the following initial conditions on Unit 2:

QRV-251, CCP Discharge Flow Control Valve is in MANUAL.

QRV-200, Charging Header Pressure Control Valve is in MANUAL.

Pressurizer pressure is 2235 psig.

Reactor Coolant Pump seal injection flow is 32 gpm.

Charging line flow is 89 gpm.

If pressurizer pressure is raised to 2300 psig, which ONE of the following sets of system parameter changes is Correct?

A. Charging line flow lowers and total seal injection flow lowers.

B. Charging line flow lowers and total seal injection flow remains the same.

C. Charging pump discharge header pressure remains the same and total seal injection flow lowers.

D. Charging pump discharge header pressure rises and total seal injection flow remains the same.

Answer: A Answer Explanation:

A. Correct - Centrifugal pump laws require that the discharge header pressure raises and flow lowers as system pressure rises. Therefore, charging line flow and total seal flow will lower while charging line discharge pressure rises.

B. Incorrect - The RCP Seals will also see the pressure rise causing the seal flow to also lower.

C. Incorrect - Centrifugal pump laws require that the discharge header pressure raises and flow lowers as system pressure rises.

D. Incorrect - Centrifugal pump laws require that the discharge header pressure raises and flow lowers as system pressure rises.

Question ID (Status) NRCAUDIT07-0977(Active)

External Topic ID: 0300-1: 0300-1, CVCS Level/Difficulty: H/3 Comments:

DC COOK OPS Page: 55 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

SOD-00300-001, Charging and Letdown System Source: Bank Previous NRC exam: 2012 KA - 003000 A3.01 Reactor Coolant Pump System (RCPS)

Ability to monitor automatic operation of the RCPS, including: Seal injection flow RO - 3.3 SRO - 3.2 CFR - 41.7 / 45.5 KA Justification - The question requires the candidate to describe to expected response of the RCP Seal Injection flow based on changing conditions.

Associated objective(s):

(RO-C-00300-E19) Predict CVCS system response for various plant and system conditions.

DC COOK OPS Page: 56 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 29 ID: RO32NRC-029 Points: 1.00 Unit 2 is at 40% reactor power with the following conditions:

Loop 2 Thot instrument has failed HIGH.

Rod control is in MANUAL.

Indicated Tavg is 574.

Which one of the following describes the Pressurizer level control system response to this failure and required operator actions?

A. Pressurizer level control system begins raising pressurizer level to a programed level of 100%. Place QRV-251 in Manual and stabilize Pressurizer level between 52 and 56%.

B. Pressurizer level control system begins raising pressurizer level to a programed level of 54.1%. Place QRV -251 in Manual and stabilize Pressurizer level between 33 and 36%.

C. Pressurizer level control system begins raising pressurizer level to a programed level of 49.1%. Place QRV -251 in Manual and stabilize Pressurizer level between 31 and 35%.

D. Pressurizer level control system maintains pressurizer level at its present level.

Verify RCI has defeated the failed transmitter and removed it from the Tave calculation.

Answer: B Answer Explanation:

Pressurizer level setpoint is derived from Auctioneered High Tave instrument. With one Thot failed HIGH => Tave would go HIGH for that loop. This would cause the Level Control system to raise level to the 100% power level of 54.1% (22 to 54.1%). 4022-IFR-001 will direct the operator to take manual control of Pressurizer level or Charging Flow Control (QRV-251) and stabilize level at the desired program.

A. Incorrect - The Pressurizer level program has a high limit of the 100% power / Tave value which is 54.1%.

B. Correct - The Pressurizer level program value will rise to 54.1% and the QRV-251 will begin to open to raise level to 54% so it is placed in manual to maintain level. 40% power level programmed level is 34.8%.

C. Incorrect - This is the correct response for Unit 1. Unit 1 40% power level DC COOK OPS Page: 57 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO programmed level is 32.8%

D. Incorrect - Many of the RCI failures will cause the value to revert to a previous good value. This is not the case with this failure since the Tavg value is failed to the 100% value and this is what drives the programmed pressurizer level.

Question ID (Status) CM-0612A(Active)

External Topic ID: 0202-2: 0202-2, Pressure and Level Control Level/Difficulty: H/3 Comments:

Reference:

OHP-4022-IFR-001 Instrument Failure Response Source: Bank KA - 004000 A2.16 Chemical and Volume Control System (CVCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations: Tave and Tref deviations RO - 3.2 SRO - 3.6 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Question requires candidate to determine how the PZR level setpoint is impacted (CVCS Flow control) and determine the action that must be taken.

Associated objective(s):

(RO-C-AOP0310412-E1) Given a set of plant conditions including a RCS RTD Instrument Malfunction, predict the response of the plant, without error.

DC COOK OPS Page: 58 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 30 ID: CM-39781 Points: 1.00 When emergency boration is started by opening QMO-410, Emergency Boration Valve, the boric acid transfer pumps __________________ to inject boron at a correct rate.

A. will start automatically in slow speed B. will automatically shift to fast speed C. must be started in slow speed by the operator D. must be shifted to fast speed by the operator Answer: D Answer Explanation:

The BA Transfer Pumps are manually shifted to fast speed prior to opening QMO-410.

A. Incorrect - Plausible since several ECCS pumps start on emergency signals to deliver highly borated water to the RCS.

B. Incorrect - Plausible since several ECCS pumps start on emergency signals to deliver highly borated water to the RCS.

C. Incorrect - Plausible since the pumps are usually run in slow speed for normal operation.

D. Correct - The pumps must be manually shifted to fast speed to achieve the correct injection rate.

Question ID (Status) CM-39781(Active)

External Topic ID: 0300-2: 0300-2, CVCS Makeup System Level/Difficulty: F/2 Comments:

Reference:

OHP-4021-005-007 Operation of Emergency Boration Flowpaths Source: Bank KA - 004000 A4.18 Chemical and Volume Control System (CVCS)

Ability to manually operate and/or monitor in the control room: Emergency borate valve DC COOK OPS Page: 59 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO RO - 4.3 SRO - 4.1 CFR - 41.7 / 45.5 to 45.8 KA Justification - Question requires candidate to monitor the opening of the Emergency Boration valve and the method required to ensure the correct flowrate is obtained.

Associated objective(s):

(RO-C-00300-E15) Describe the purpose/function of the following instruments used to verify proper operation of the CVCS:

a. Temperature

b. Pressure

c. Flow

d. Level DC COOK OPS Page: 60 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 31 ID: RO32NRC-031 Points: 1.00 Given the following plant conditions:

The plant has sustained a Large Break Loss of Coolant Accident.

The Train A ECCS and CTS pumps failed to start

CCP BIT total flow is 450 gpm.

Residual Heat Removal flow is 3600 gpm.

Containment Spray flow is 3000 gpm.

Safety Injection Pump flow is 300 gpm.

19 minutes ago, Annunciator 105 Drop 22 RWST Level at 375,500 Gal was received.

Annunciator 105 Drop 23 RWST Level High OR Low has just alarmed.

Did this Drop 23 annunciate sooner than, later than, or when expected and what actions are now required?

A. Sooner than expected.

Secure Pumps taking suction from the RWST until the sump can be aligned.

B. Sooner than expected.

Immediately transition to 1-OHP-4023-ES-1.3, Transfer to COLD Leg Recirculation.

C. When expected.

Immediately transition to 1-OHP-4023-ES-1.3, Transfer to COLD Leg Recirculation.

D. Later than expected.

Secure Pumps taking suction from the RWST until the sump can be aligned.

Answer: B Answer Explanation:

ANSWER: Flow rate is 7350 gpm. TS Min. is 375,500 gallons. 105 Drop 23 is at 30%

(134,300 gallons)

Tank volume change is 375,500 gallons - 134,300 gallons (30%) for a change of 241,200 gallons. (241,200/7359) = 32.8 minutes.

A. Incorrect. The expected time to reach 30% RWST level is 32.8 minutes. The ES-1.3 must be entered to begin alignment to the recirc sump. All pumps taking suction from RWST are stopped upon a low-low level in the RWST.

DC COOK OPS Page: 61 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO B. Correct. The expected time to reach 30% RWST level is 32.8 minutes. The ES-1.3 must be entered to begin alignment to the recirc sump.

C. Incorrect. If flow rate for both trains is used and full tank to a 20% level is used, then 19 minutes is the expected time D. Incorrect. If flow rate for both trains is used and full tank to a 30% level is used, then 16 minutes is the expected time. All pumps taking suction from RWST are stopped upon a low-low level in the RWST.

Question ID (Status) RO32NRC-031(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

Technical Data Book 12-Figure 18.10, Rev 1 -

Calculation 1- 2-19-03 CALC 6, Rev. 1, determined 4140 gal/%/ span Source: New KA - 005000 2.4.46 Residual Heat Removal System (RHRS)

Emergency Procedures/Plan Ability to verify that the alarms are consistent with the plant conditions.

RO - 4.2 SRO - 4.2 CFR - 41.10 / 43.5 / 45.3 / 45.12 KA Justification - Question requires candidate to determine if the alarm is Correct given the RHR and ECCS flows.

Associated objective(s):

(RO-C-EOP09-E36) For each of the E-1 Series procedures, discuss the basis or reason for all Steps.

DC COOK OPS Page: 62 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 32 ID: CM-7056 Points: 1.00 Select ALL automatic actions that occur in the Chemical and Volume Control System (CVCS) directly from the initiation of Safety Injection signal.

Note: Exclude Phase 'A' actions.

1) RWST to Charging Pump Suction Valves (IMO-910/911) Open

2) Both Centrifugal Charging Pumps Start

3) Charging Header Flow Control Valve (QRV-251) Closes

4) Charging Header Isolation Valves (QMO-200/201) Close

5) Seal Water Return Containment Isolations (QCM-250/350) Close

6) Letdown Containment Isolation Valves (QCR-300/301) Close

7) VCT Outlet Valves (QMO-451/452) Close

8) RHR Supply to CVCS (IRV-300) Closes A. 1, 2, 4, 7 B. 1, 3, 5, 8 C. 2, 3, 6, 8 D. 4, 5, 6, 7 Answer: A Answer Explanation:

Upon receipt of a Safety Injection signal, the CVCS system realigns to support rapid injection of highly borated water into the primary system. This includes alignment to the RWST as a suction source (1), auto-start of both CCPs (2), closing the normal charging flow path (4), and isolating the VCT as a suction source when the RWST is aligned (7).

The answer with numbers 1, 2, 4, 7 is fully correct.

1) RWST to Charging Pump Suction Valves (IMO-910/911) Open (CORRECT)

2) Both Centrifugal Charging Pumps Start (CORRECT)

3) Charging Header Flow Control Valve (QRV-251) Closes (INCORRECT - Plausible, as downstream valves QMO-200/201 do close on an SI signal)

4) Charging Header Isolation Valves (QMO-200/201) Close (CORRECT)

DC COOK OPS Page: 63 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

5) Seal Water Return Containment Isolations (QCM-250/350) Close (INCORRECT - Plausible, the valves do close on containment isolation signals)

6) Letdown Containment Isolation Valves (QCR-300/301) Close (INCORRECT - Plausible, the valves do close on containment isolation signals)

7) VCT Outlet Valves (QMO-451/452) Close (IF IMO-910 or IMO-911 is Open)

(CORRECT)

8) RHR Supply to CVCS (IRV-300) Closes (INCORRECT - Plausible charging suction supply valves do realign upon an SI signal)

Question ID (Status) CM-7056(Active)

External Topic ID: 0300-1: 0300-1, CVCS Level/Difficulty: F/3 Comments:

Reference:

OHP-4030-132-217A DG1CD Load Sequencing & ESF Testing, Data Sheet 5 Verification of train A Safety injection Equipment positions NOTE: QMO-451/452 receive a signal from SI actuation but will not close until either IMO-910 or IMO-911 have fully opened. OHP-4030-132-217A DG1CD Load Sequencing & ESF Testing Data Sheet 5 Verification of train A Safety injection Equipment positions Source: Bank KA - 006000 K1.08 Emergency Core Cooling System (ECCS)

Knowledge of the physical connections and/or cause-effect relationships between the ECCS and the following systems: CVCS RO - 3.6 SRO - 3.9 CFR - 41.2 to 41.9 / 45.7 to 45.8 KA Justification - Question requires candidate to identify the valve alignments for the connection of CVCS to the ECCS system.

Associated objective(s):

(RO-C-00300-E13) Describe the automatic response of the CVCS components to a Safety Injection/Phase A Containment Isolation actuation signal.

DC COOK OPS Page: 64 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 33 ID: RO32NRC-033 Points: 1.00 The Crew is attempting to align the RHR supply to the CCP and SI pumps per 1-OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation.

Both RHR pumps are running with suctions aligned to the Recirculation Sump.

IMO-340, Charging Pump Suction from East RHR Hx is OPEN.

IMO-350, SI Pp Suction from West RHR Hx will NOT OPEN.

IMO-360, SI Pp Suction X-Tie to Charging Pumps Suction is OPEN.

IMO-262, OR IMO-263 SI PPS Recirc to RWST are CLOSED.

Which ONE of the following describes the actions required to complete this alignment?

A. OPEN IMO-361 AND IMO-362, SI Pp Suction X-Tie to Charging Pumps Suction.

Both CCPs and SI pumps may continue to operate.

B. OPEN IMO-361 AND IMO-362, SI Pp Suction X-Tie to Charging Pumps Suction.

Shutdown one CCP and one SI pump due to the restricted flowpath.

C. CLOSE IMO-360, SI Pp Suction X-Tie to Charging Pumps Suction to make up the interlock to OPEN IMO-350, SI Pp Suction from West RHR Hx. After IMO-350 is open Both CCPs and SI pumps may continue to operate.

D. Stop all running SI pumps since their suction flowpath is limited and both CCPs are running.

Answer: A Answer Explanation:

Answer Explanation:

A. Correct - The East RHR pump provides suction to the Charging pumps which may also share suction with the SI pumps through the IMO-361 and 362 valves. This flow path is sufficient to run all of the SI and CCP pumps.

B. Incorrect - This flow path is sufficient to run all of the SI and CCP pumps.

C. Incorrect - The East RHR pump provides suction to the Charging pumps which may also share suction with the SI pumps but through the IMO-361 and 362 valves AND IMO-360. IMO-262 OR IMO-263 SI Pumps Recirc to RWST valves must be closed to allow opening IMO-350. Plausible if student thinks that suction ring header valve IMO-360 must be closed to open DC COOK OPS Page: 65 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO IMO-350 D. Incorrect - Flow is provided to the SI pumps through the CCP suction crosstie. This is aligned per the procedure. If the crosstie could not be aligned this would be the correct answer.

Question ID (Status) RO32NRC-033(Active)

External Topic ID:

Level/Difficulty: H Comments:

Reference:

1-OHP-4023-ES-1.3 Transfer to Cold Leg Recirculation Step 14 Source: New KA - 006000 K6.10 Emergency Core Cooling System (ECCS)

Knowledge of the effect of a loss or malfunction of the following will have on the ECCS:

Valves RO - 2.6 SRO - 2.8 CFR - 41.7 / 45.7 KA Justification - Question requires candidate knowledge of the impact and required actions to take for the failed SI pump suction supply from RHR.

Associated objective(s):

(RO-C-EOP09-E36) For each of the E-1 Series procedures, discuss the basis or reason for all Steps.

DC COOK OPS Page: 66 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 34 ID: NRCAUDIT07-0805 Points: 1.00 Which ONE of the following describes the adverse effects of NO operator action with a leaking pressurizer PORV?

A. There are NO adverse effects. The PRT is designed to handle continuous in-leakage.

B. The cyclic temperature stresses in combination with inner wall erosion on the PORV tailpipe may lead to premature piping failure.

C. The PRT rupture disc may break with subsequent elevated radiation, temperature and pressure indications in containment.

D. Mechanical breakdown of the PORV seating surface may cause the PORV to fail when needed for overpressure protection.

Answer: C Answer Explanation:

A. Incorrect - Plausible since the PRT can handle in-leakage but the PRT is not designed for continuous input without any actions to cool and drain.

B. Incorrect - Plausible since cyclic temperatures could lead to excessive stresses but, with a constant leak the temperatures will not be cycling, PORV seat cutting/erosion may be a concern but not inner wall erosion.

C. Correct - The tank design is based on the requirement to condense and cool a discharge of pressurizer steam equal to 110 percent of the volume above the 100%-power pressurizer water level set-point. If the temperature in the tank rises above 126°F during plant operation, the tank is cooled by spraying in cool water and draining out the warm mixture to the Waste Disposal System. The tank is not designed to accept a continuous discharge from the pressurizer.

D. Incorrect - Plausible since damage may occur (PORV seating may erode) but it would be available for overpressure protection.

Question ID (Status) NRCAUDIT07-0805(Active)

External Topic ID: 0202-1: 0202-1, Pressurizer/Pressure Relief Level/Difficulty: F/3 Comments:

DC COOK OPS Page: 67 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

RO-C-0202-1 Pressurizer/Pressure Relief Source: Bank Previous NRC exam: 2002 KA - 007000 K3.01 Pressurizer Relief Tank/Quench Tank System (PRTS)

Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment RO - 3.3 SRO - 3.6 CFR - 41.7 / 45.6 KA Justification - Question tests candidates knowledge of conditions that may lead to loss of PRT (RUPTURE Disc) and subsequent impact on Containment.

Associated objective(s):

(RO-C-0020920412-E1) Given a set of plant conditions which includes a Leaking Pressurizer Power Operated Relief Valve, predict the response of the plant without operator intervention including final plant configuration, without the use of references and without error.

DC COOK OPS Page: 68 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 35 ID: CM-0217 Points: 1.00 Unit 2 is operating at 100% power with the East Component Cooling Water (CCW) pump supplying system loads. A load shed occurs on T21A.

Which ONE of the following describes a correct response of the 2 East and 2 West CCW pumps?

A. East CCW pump continues to run, and the West CCW pump will not start.

B. East CCW pump will be shed from its bus, then the Blackout Sequencer will start the West CCW pump.

C. East CCW pump will be shed from its bus, then the West CCW pump will auto start on low pressure.

D. East CCW pump continues to run, and the Blackout Sequencer will start the West CCW pump.

Answer: D Answer Explanation:

A. Incorrect - The East Pump supplied from T21D continues to run. The West CCW pump auto starts following time delays.

B. Incorrect - The East Pump supplied from T21D continues to run.

C. Incorrect - This would be true if the East Pump was supplied from T21A.

D. Correct - The East Pump supplied from T21D continues to run and the West Pump will auto start after a time delay once the EDG loads the bus.

Question ID (Status) CM-0217(Active)

External Topic ID: 1600: 1600, Component Cooling Water Level/Difficulty: H/3 Comments:

Reference:

OHP-4030-132-217A DG1CD Load Sequencing & ESF Testing data Sheet 3, RO-C-01600 Component Cooling Water Source: Bank Previous NRC exam: 2018 DC COOK OPS Page: 69 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO KA - 008000 K2.02 Component Cooling Water System (CCWS)

Knowledge of bus power supplies to the following: CCW pump, including emergency backup RO - 3.0 SRO - 3.2 CFR - 41.7 KA Justification - This meets the K/A in that the candidate must know the correct power supply to the CCW pumps and their response to a loss of power.

Associated objective(s):

(RO-C-01600-E12) predict the impact and response of the Component Cooling Water System, DC COOK OPS Page: 70 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 36 ID: 2008NRC-0611 Points: 1.00 Which ONE of the following contains ONLY conditions that will cause the UNIT 1 pressurizer heaters to automatically de-energize?

1. Pressurizer Level Control Channel - 5% below program level.

2. Pressurizer Level Control Channel - less than 17% level.

3. Pressurizer Level Bi-stable Channel - less than 17% level.

4. Pressurizer Level Cold Calibration Channel - less than 17% level.

A. 1 and 3 B. 2 and 3 C. 2 and 4 D. 1 and 4 Answer: B Answer Explanation:

A. Incorrect - Protection is provided by the control channel but at 17%. Plausible since at 5% above program the heaters energize.

B. Correct - Protection is provided when either the CONTROL or BISTABLE channel is

<17%

C. Incorrect - The Interlock does NOT exist on the cold cal channel - plausible since this instrument is used for main control as the RCS is cooled down and it reads lower than hot cals (hot cal channels don't provide true protection at lower temperatures.

D. Incorrect - Protection is provided by the control channel but at 17%. Plausible since at 5% above program the heaters energize. The Interlock does NOT exist on the cold cal channel - plausible since this instrument is used for main control as the RCS is cooled down and it reads lower than hot cals (hot cal channels don't provide true protection at lower temperatures.

Question ID (Status) 2008NRC-0611(Active)

External Topic ID: 0202-1: 0202-1, Pressurizer/Pressure Relief Level/Difficulty: F/2 Comments:

DC COOK OPS Page: 71 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

OHP-4024-108 drop 3, Pressurizer Level High Dev Backup Htrs On, Drop 5, Pressurizer Level Low All Htrs Off, SOD-00202-003, RO-C-00202 Source: Bank Previous NRC exam: 2008 KA - 010000 K4.02 Pressurizer Pressure Control System (PZR PCS)

Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following: Prevention of uncovering PZR heaters RO - 3.0 SRO - 3.4 CFR - 41.7 KA Justification - Requires the knowledge of the heater protection interlocks provided by the PZR level control system.

Associated objective(s):

(RO-C-00202-E16) Describe the conditions that will cause the component to trip, automatically/manually start and/or automatically/manually reposition.

a. Pressurizer Spray Valves

b. Pressurizer PORVs

c. Pressurizer Heaters

d. Pressurizer Safeties

e. Charging Flow Control Valve QRV-251 DC COOK OPS Page: 72 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 37 ID: 2007-0434 Points: 1.00 Given the following:

Unit 2 is at 50% power with all controls in Automatic.

A failure of turbine first stage pressure instrumentation causes rods to slowly withdraw.

Rods continue to withdraw slowly when placed in Manual.

Assuming NO (further) operator actions, which ONE of the following trips will be the first to ensure DNB parameters are NOT exceeded for this transient?

A. Overpower-Delta Temperature B. Power Range High Flux (high setpoint)

C. Overtemperature-Delta Temperature D. Pressurizer High Level Answer: C Answer Explanation:

A. Incorrect - Fuel integrity and total core power protection.

C. Correct - FSAR Chapter 14 Transient and accident analysis describes that OTDT is provided to address a slow control rod withdrawal transient at lower power levels.

B. Incorrect - Fuel integrity protection D. Incorrect - RCS pressure protection Question ID (Status) 2007-0434(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

FSAR chapter 14.1.0.6 Reactor Protection System (RPS) Setpoints and Time Delays Source: Bank Previous NRC exam: 2014 DC COOK OPS Page: 73 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO KA - 012000 K5.01 Reactor Protection System Knowledge of the operational implications of the following concepts as they apply to the RPS: DNB RO - 3.3 SRO - 3.8 CFR - 41.5 / 45.7 KA Justification - Question tests candidate knowledge of the trips designed for protection from DNB.

Associated objective(s):

(TRANS2C) Given the following primary side reactivity transients:

a. Uncontrolled RCCA Withdrawal from a Subcritical Reactor

b. Uncontrolled RCCA Withdrawal at Power

c. Chemical and Volume Control System Malfunction (Uncontrolled Boron Dilution)

Discuss the transient including initial conditions and assumptions, plant response, final plant conditions, and overall assessment of the operation of the plant, in accordance with Chapter 14 of the UFSAR.

DC COOK OPS Page: 74 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 38 ID: CM-39479 Points: 1.00 The Unit 2 reactor failed to automatically trip when the Reactor Coolant Pumps tripped.

The following conditions exist after a MANUAL Turbine Trip is attempted per OHP-4023-FR-S-1, Response to Nuclear Power Generation/ATWS step 3:

Turbine Stop Valve Closed Status Lights - 1, 2, and 3 Lit

Turbine Stop Valve Closed Status Lights - 4 NOT Lit

MAIN TURBINE STOP VALVE CLOSED alarm - Lit

AMSAC INITIATED alarm - Lit Which ONE of the following is the NEXT action the operator is required to take?

A. Shut the Main Steam Stop Valves.

B. Actuate ATWS Turbine Runback.

C. Verify AFW Pumps running.

D. Manually actuate AMSAC.

Answer: B Answer Explanation:

A. Incorrect - Closing the Main Steam Stop valves is only performed if a manual load reduction does not work.

B. Correct - The turbine is verified tripped by checking all 4 status lights closed. The alarms are lit based on 1 stop valve closed and the AMSAC initiated.

Since the turbine is not tripped, 2-OHP-4023-FR-S-1 requires that load be manually reduced.

C. Incorrect - Checking the AFW Pumps running is step 4 but not the next action since the turbine trip has not been verified.

D. Incorrect - Step 2 of 2-OHP-4023-FR-S-1 actuates AMSAC. This is NOT performed in the turbine trip verification step as it is in the 2-OHP-4023-E-0, Reactor Trip or Safety Injection, procedure.

Question ID (Status) CM-39479(Active)

External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: H/3 Comments:

DC COOK OPS Page: 75 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

2-OHP-4023-FR-S-1, Response to Nuclear Power Generation/ATWS Step 3 Source: Bank Previous NRC exam: 2012 KA - 012000 2.4.49 Reactor Protection System Emergency Procedures/Plan Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

RO - 4.6 SRO - 4.4 CFR - 41.10 / 43.2 / 45.6 KA Justification - Question requires operator evaluate a turbine trip per FR-S.1 and determine Correct actions to take to complete the trip.

Associated objective(s):

(RO-C-EOP04-E13) List the Immediate Operator Actions for FR-S.1 DC COOK OPS Page: 76 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 39 ID: RO32NRC-039 Points: 1.00 Given the following conditions:

The East Charging Pump has been in service

Letdown flow is 75 gpm

QMO-225, East CCP Mini-Flow to RCP Seal Water Hx Shutoff Valve is OPEN

Subsequently, a SBLOCA occurred

Containment Pressure <1 psig Following the trip, the RO identifies that NPP-151 has just failed high (2500 psig).

RCS pressure is 1700 psig and lowering on all the other RCS\Pressurizer Channels.

Which of the following describes the status of the white lights above charging pump leak-off valves QMO-225, East CCP Mini-Flow, and QMO-226, West CCP Mini-Flow, and the position of each valve?

A. QMO-225: White light ON, valve OPEN QMO-226: White light ON, valve CLOSED B. QMO-225: White light OFF, valve OPEN QMO-226: White light ON, valve CLOSED C. QMO-225: White light ON, valve CLOSED QMO-226: White light ON, valve CLOSED D. QMO-225: White light OFF, valve OPEN QMO-226: White light OFF, valve OPEN Answer: A Answer Explanation:

A. Correct - The White Lights are associated with the Train SI actuated Status, so they should both be lit. QMO-225 will not close since it is controlled by NPP-151 pressure.

B. Incorrect - This is plausible if Train A SI is not actuated, the white light above QMO-225 will remain de-energized and the valve will remain open. Train B SSPS functions properly, therefore with RCS pressure <1825 psig, QMO-226 will be closed.

DC COOK OPS Page: 77 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO C. Incorrect - QMO-225 white light would be ON, but the valve would remain open with NLP-151 at 2235. Plausible if the RCS pressure <1825 psig was not channel specific.

D. Incorrect - SI on both trains would have been actuated by the 2 remaining Pressure Channels going < 1775 psig, so the QMO-226 white light would be lit, and the valve would be closed with RCS pressure <1825 psig.

Question ID (Status) RO32NRC-039(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

OHP-4023-E-0 Reactor Trip or Safety Injection, OHP-4022-013-009 Pressurizer Pressure Instrument Malfunction Source: New KA - 013000 K6.01 Engineered Safety Features Actuation System (ESFAS)

Knowledge of the effect of a loss or malfunction of the following will have on the ESFAS: Sensors and detectors RO - 2.7 SRO - 3.1 CFR - 41.7 / 45.7 KA Justification: Question requires candidate to determine the impact of the loss of a pressurizer pressure channel on the SI and CVCS ELO ESFAS actuation signals.

Associated objective(s):

(RO-C-AOP0330412-E3) Given a set of plant conditions including a Pressurizer Pressure Instrument Malfunction or Controller Failure explain the procedural mitigation strategy for the malfunction in accordance with OHP-4022-013-009, Pressurizer Pressure Instrument Malfunction.

DC COOK OPS Page: 78 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 40 ID: NRCAUDIT07-1073A Points: 1.00 A malfunction of the Containment Chilled Water (CHW) system has resulted in excessive CHW flow to the Containment Ventilation Units.

Under this condition, which ONE of the following sets of containment readings would result in Unit 2 containment still being within Technical Specification limits?

A. Containment pressure - 0.18 PSIG, Upper containment temperature 57°F, Lower containment temperature 75°F.

B. Containment pressure 0.35 PSIG, Upper containment temperature 95°F, Lower containment temperature 97°F.

C. Containment pressure 0.15 PSIG, Upper containment temperature 105°F, Lower containment temperature 75°F.

D. Containment pressure - 1.05 PSIG, Upper containment temperature 75°F, Lower containment temperature 105°F.

Answer: D Answer Explanation:

LCO 3.6.4 Containment pressure shall be > (-) 1.5 psig and < (+) 0.3 psig. LCO 3.6.5 Containment average air temperature shall be:

a. >60°F and < 100°F for the containment upper compartment and

b. >60°F and < 120°F for the containment lower compartment.

A. Incorrect - Pressure is low but within limits. Upper temp is too low. Plausible since temperatures are cool as would be expected from too much flow.

B. Incorrect - Pressure too high, plausible since upper temp is below lower temp &

Pressure is still reasonable.

C. Incorrect - Upper Temp too high. Plausible if candidate doesn't know that Upper limit is lower.

D. Correct - Pressure is negative but within limits and temperatures are cool as would be expected from too much flow.

DC COOK OPS Page: 79 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) NRCAUDIT07-1073A (Active)

External Topic ID: 3400: 3400, Containment Level/Difficulty: F/3 Comments:

Reference:

Unit 2 T.S. 3.6.4 and 3.6.5 Source: Modified Previous NRC exam: 2018 KA - 022000 A1.02 Containment Cooling System (CCS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment pressure RO - 3.6 SRO - 3.8 CFR - 41.5 / 45.5 KA Justification - This meets the K/A because the candidate must know the Containment limits that are being monitored.

Original Question # - Modified from NRC EXAM 2012, NRCAUDIT07-1073, RO23-090-4-Q72, RO25 Audit-37 Original Question KA - SYS022 A1.04 Modified Question stem from loss of CHW to excessive flow. Changed all distractors to make temps cooler and lower pressures.

Previously Distractor A was correct answer with positive pressure and higher temps.

Now distractor D is Correct with lower pressure and reduced temps.

Associated objective(s):

(RO-C-02800-E13) Given a set of plant conditions and/or results of a Surveillance Test, determine all applicable TS, the System Operability, and the most limiting LCO.

DC COOK OPS Page: 80 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 41 ID: 2008NRC-0321 Points: 1.00 Given the following plant conditions on Unit 1:

The unit was operating at 100% power when an INADVERTENT Phase A Containment Isolation occurred on Train A.

The Crew has reset Phase A Containment Isolation and attempted to restore Control Air to Containment.

The Control Air Containment Isolation Valves could not be opened.

Which ONE of the following describes short-term impact of the loss of air on the restoration efforts of the crew and the required compensatory actions?

A. RCP NESW Motor Air cooling water can NOT be restored. Trip the reactor and stop 3 RCPs. Perform a containment pressure relief.

B. Glycol Cooling to the ice condenser can NOT be restored. Stop all Unit 1 Ice Condenser Air Handling Units (AHUs). Monitor ice bed temperatures to ensure they remain at an acceptable level.

C. RCS overpressure protection has been lost (PORVs will NOT open). Begin a reactor shutdown and be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D. RCP Seal Injection is available, but Seal Return can NOT be restored. Drain the PRT as required to maintain an acceptable level.

Answer: B Answer Explanation:

A. Incorrect - NESW to RCP Motor Cooling valves are located outside containment and close on a Phase B Isolation. Actions are Correct for Loss of NESW.

B. Correct - Glycol Cooling inside Containment Isolation valves VCR-11 and VCR-21 will NOT open. The Ice Condenser AHU's are stopped if the glycol system is shut down for more than 30 minutes. Technical Specifications requires that temperatures are maintained < 27°F.

C. Incorrect - PORVs NRV-152 and NRV-153 have local reservoirs. Technical Specifications require a shutdown if all PORVS are lost.

D. Incorrect - RCP Seal Injection is not isolated and Seal Return QCM-250 and QCM-350 are motor operated valves. RCP Seal Leakoff valves QRV-10, 20, 30, and 40 are fail open. Seal Return would go to the PRT if the Containment DC COOK OPS Page: 81 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Isolation valves were closed.

Question ID (Status) 2008NRC-0321(Active)

External Topic ID: 1000: 1000, Ice Condenser Level/Difficulty: H/3 Comments:

Reference:

RO-C-01000 Ice Condenser, TS 3.6.11, 12-OHP-4024-135 Drop 51 Glycol Expansion Tank Level High-High Source: Bank Previous NRC exam: 2014 KA - 025000 A2.04 Ice Condenser System Ability to (a) predict the impacts of the following malfunctions or operations on the Ice Condenser System and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations:

Containment isolation RO - 3.0 SRO - 3.2 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Question tests knowledge of the impact an inadvertent Phase A Containment Isolation has on the Ice Condenser cooling system & subsequent actions.

Question Source -NRC Exam 2014 Q#14 modified from Cook NRC Exam 2004-060-2 Associated objective(s):

(RO-C-01000-E8) Describe the conditions that will cause the Ice Condenser system component to trip, automatically / manually start and/or automatically / manually reposition.

a. Glycol Pumps

b. Glycol Chillers

c. Air Handling Units

d. Ice Condenser Floor Cooling Loops (pumps, defrost heaters)

e. Valves DC COOK OPS Page: 82 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 42 ID: RO-C-00900-E12-7 Points: 1.00 Following a LOCA inside containment in which pressure reached 2.8 psig, the operator notes the following indications:

Containment pressure is 3.1 psig and raising slowly

Refueling Water Storage Tank (RWST) level is 80% and lowering slowly

WHITE CTS Timer ON status lamps are lit Assuming NO operator actions, TWO MINUTES after pressure reached 2.8 psig the operator would observe:

__(1)__ IMO 210/220 & IMO 211/221 CTS Pump Discharge Valves

__(2)__ IMO-212/222 Eductor Supply Valves

__(3)__ CTS pumps (1) (2) (3)

A. Closed Closed NOT Running B. Open Open Running C. Open Open NOT Running D. Open Closed Running Answer: C Answer Explanation:

The CTS valves will receive the OPEN signal when pressure rises above 2.8 PSIG. The discharge valve stroke times are ~ 1.5 minutes. The Pump Start is delayed for about 4 minutes.

A. Incorrect - Only the pump start is delayed, the valves stroke open on the signal B. Incorrect - This would have been to correct response prior to the timer.

C. Correct - Valves will align and pump will wait for the timer D. Incorrect - Spray Additive tank and eductor would still be open. This would be plausible if the Spray Add tank flow was delayed and not the pump start DC COOK OPS Page: 83 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) RO-C-00900-E12-7(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

1-OHP-4023-E-0 Reactor Trip or Safety Injection Step 6, SOD -00901-001, Containment Spray and Hydrogen Recombiners Source: Modified Previous NRC exam: 2018 KA - 026000 A3.01 Containment Spray System (CSS)

Ability to monitor automatic operation of the CSS, including: Pump starts and Correct MOV positioning RO/SRO Value - (4.3 / 4.5) CFR - 41.7 / 45.5 KA Justification - Question tests candidate ability to monitor the pumps and valves for correct operation following CTS actuation.

Modified from RO-C-00900-E12-6 from NRC exam 2018.

Changed to CTS has just actuated (But before the timer has timed out) and changed from suction valves to discharge valves. Removed low RWST and Spray ADD tank Alarm from the Stem.

Changed Distractor C to make open, open, not running and make it Correct answer instead of A.

Associated objective(s):

(RO-C-00900-E12) Describe the conditions that will cause the following Containment Spray system and/or hydrogen removal or monitoring system components to trip, automatically/manually start and/or automatically/manually reposition:

a. CTS pump start and trip

b. CTS pump discharge valve opening

c. SAT outlet valve opening

d. SAT outlet valve closure

e. Eductor supply valve closure

f. CTS heat exchanger ESW outlet valve opening DC COOK OPS Page: 84 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 43 ID: CM-0415 Points: 1.00 During a cooldown on Unit 2 the following conditions exist:

RCS loop Tavg:

Loop 1: 538°F lowering

Loop 2: 542°F lowering

Loop 3: 537°F lowering

Loop 4: 540°F lowering

Steam header pressure 900 psig and lowering

Steam Dump Mode Selector switch is in STM PRESS MODE

Steam Dump Controller is in MAN set at 30% demand

ALL Steam Dumps closed The operator momentarily places the Steam Dump Control Selector Train A and Control Selector Train B switches to BYPASS INTRLK and then releases them.

What is the expected status of the Steam Dump valves following the operator's actions?

A. All valves remain closed.

B. The valves in group 1 and 2 are open and the valves in group 3 are closed.

C. The valves in group 1 and 2 are open and the valves in group 3 are partially open.

D. The valves in group 1 are open and the valves in groups 2 and 3 are closed.

Answer: D Answer Explanation:

A. Incorrect - Incorrect because Group 1 will open. This is a plausible answer if the student does not believe the interlock will be bypassed in the conditions given in the question.

B. Incorrect - Incorrect but plausible if the student confuses the normal turbine trip circuit operation with the less than P-12 circuit operation C. Incorrect - Incorrect but plausible if the student does not understand that the P-12 bypass limits the groups that can be opened.

D. Correct - When the steam dumps have the P-12 interlock bypassed, only group 1 valves are bypassed. A temp mod can be installed to open more valves, but this temp mod can only be installed when RCS temp is less than 350F DC COOK OPS Page: 85 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO by procedure.

Question ID (Status) CM-0415(Active)

External Topic ID: 5200: 5200, Steam Dumps Level/Difficulty: H/3 Comments:

Reference:

OHP-4021-001-004 Plant Cooldown from Hot Standby to Cold Shutdown, OHP-4021-052-001 Steam Dump Control System Source: Bank KA - 039000 A4.07 Main and Reheat Steam System (MRSS)

Ability to manually operate and/or monitor in the control room: Steam dump valves RO - 2.8 SRO - 2.9 CFR - 41.7 / 45.5 to 45.8 KA Justification - Question requires candidate to monitor the Steam dumps for correct operation following the P- 12 Bypass.

Associated objective(s):

(RO-C-05200-E6) State the Steam Dump System response to various plant transients, including the following:

a. Normal RCS Cooldown

b. Normal Reactor Startup and Power Escalation

c. Main turbine trip

d. Load rejection

e. Relationship of Tavg setpoint in SDS to primary cooldown

f. Tavg/Tref program

g. Relationship of SG pressure to steam flow

h. Operation of loss of load bi-stables on turbine trip DC COOK OPS Page: 86 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 44 ID: RO32NRC-044 Points: 1.00 Unit ONE is at 100% power in a normal line up. Both MFPs are in Auto DP Control.

Turbine Bypass Header Pressure Transmitter, UPC-102B is disabled.

MFP Speed Control System Turbine Bypass Header Pressure Transmitter, UPC-102A, slowly begins to fail LOW.

As this signal is slowly lowering the FW Pump delta P signal is _(1)__than required, causing the feed pumps to_(2)___.

A. (1) larger; (2) slow down.

B. (1) smaller; (2) slow down.

C. (1) smaller; (2) speed up.

D. (1) larger; (2) speed up.

Answer: A Answer Explanation:

A. Correct - The Main FW Pump Speed control compares the UPC-102A/B (highest 102A since B is disabled) steam header pressure to the FW pump Discharge pressure FPC-250A/B (lowest). The speed control attempts to maintain the Main FW Pump speed such that the FW header to Steam Header DP is on Program. When the Steam Header Pressure drifts Low, it will appear that a Larger DP exists which will lower FW pump speed to try to lower FW pump Discharge header pressure.

B. Incorrect - Steam to FW discharge pressure DP will be larger. Plausible if the student does not know how the DP is being derived.

C. Incorrect - Steam to FW discharge pressure DP will be larger. The controller will lower FW pump Speed. Plausible if the student does not know how the DP is being derived.

D. Incorrect - The controller will lower FW pump Speed. Plausible if the student does DC COOK OPS Page: 87 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO not know how the impact of DP effects SGFP controls.

Question ID (Status) CM-0450A(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

TS3000, DCS BODD, Page 4-7 Source: Modified KA - 059000 A1.07 Main Feedwater (MFW) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW System controls including:

Feed Pump speed, including normal control speed for ICS RO - 2.5 SRO - 2.6 CFR - 41.5 / 45.5 K/A Justification - Question tests ability to predict changes associated with the FW Pump Speed.

Modified Question CM-0450A to make UPC-102A fail the other direction. Changed Correct Answer from C to A.

Associated objective(s):

(RO-C-05501-E5) Describe the principles of Main Feed Pump Turbine control including system response to changes in input parameters.

DC COOK OPS Page: 88 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 45 ID: RO32NRC-045 Points: 1.00 1-OHP-4021-055-003, Attachment 3 East Feed Pump Startup as Second Feed Pump, tracks a 4-hour time interval from the time the East FW pump is Reset until the speed of 3850 to 4200 rpm is reached.

Which ONE of the following describes the reason for this limitation?

A. The FW Pump speed must be raised slowly over a 4-hour period to minimize vibrations.

B. This tracks the 4-hour interval during FW startup that the FW pump trip input to the AFW start logic may be inoperable.

C. This tracks the 4-hour heatup time required before the FW pump is permitted to deliver flow.

D. The longer the time that a FW pump is operated at low/no load the shorter the life between overhauls.

Answer: B Answer Explanation:

A. Incorrect - Vibrations are a concern during FW pump startup but the critical speed ranges and usually hurried through to prevent operating within that range.

B. Correct - TS 3.3.2 H contains a note "Two channels on one Main Feedwater pump may be inoperable for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during the process of removing the pump from service or placing the pump in service" that is reflected in the startup procedure.

C. Incorrect - There are many references to the correct heatup of the FW pump system and oil throughout the startup procedure, but they do not contain the 4-hour limit.

D Incorrect - There is a Precaution against operating the pump at low loads for long periods time will lead to shorter life between overhauls.

Question ID (Status) RO32NRC-045(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

DC COOK OPS Page: 89 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

1-OHP-4021-055-003 Placing A Main Feed Pump in Service Attachment 3 East Feed Pump Startup as Second Feed Pump Source: New KA - 059000 2.1.32 Main Feedwater (MFW) System Conduct of Operations Ability to explain and apply system limits and precautions.

RO - 3.4 SRO - 4.0 CFR - 41.10 / 43.2 / 45.12 K/A Justification - Question requires candidate knowledge of the reason for 4-hour limit.

Associated objective(s):

(RO-C-NOP7-E36) Describe how to start a second Main Feedwater Pump during a power escalation in accordance with OHP-4021-055-003, Placing A Main Feed Pump in Service.

DC COOK OPS Page: 90 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 46 ID: CM-0477 Points: 1.00 To establish Essential Service Water (ESW) flow to the suction of the Auxiliary Feedwater Pumps (AFP), the ESW supply valves WMO-753, 754, and 744 are:

A. automatically opened on an AFP low suction pressure signal, which allows ESW flow to the AFPs.

B. automatically opened, but manual ESW isolation valves must be opened locally to allow ESW flow to the AFPs.

C. opened remotely by the Control Room operator, which allows ESW flow to the AFPs.

D. opened remotely by the Control Room operator, but manual ESW isolation valves must be opened locally to allow ESW flow to the AFPs.

Answer: D Answer Explanation:

A. Incorrect - Plausible if thought auto swap over occurred on low suction pressure.

However, there are no auto actions for ESW supply valves.

B. Incorrect - Plausible if thought auto swap over occurred (yet manual actions are still required). However, there are no auto actions for ESW supply valves.

C. Incorrect - Plausible if though that ESW could be supplied from control room action solely. However, the manual ESW isolation valves must also be opened locally to allow ESW flow to the AFPs.

D. Correct - To establish ESW flow to the suction of the AFPs, the ESW supply valves WMO-753, 754, and 744 must be opened remotely by the Control Room operator however, manual ESW isolation valves must also be opened locally to allow ESW flow to the AFPs.

Question ID (Status) CM-0477(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

OHP-4022-055-003 Loss of Condensate to Auxiliary Feedwater Pumps, DC COOK OPS Page: 91 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO SD-DCC-HP102 Source: Bank KA - 061000 K1.07 Auxiliary / Emergency Feedwater (AFW) System Knowledge of the physical connections and/or cause-effect relationships between the AFW System and the following systems: Emergency water source RO - 3.6 SRO - 3.8 CFR - 41.2 to 41.9 / 45.7 to 45.8 K/A Justification - Question requires candidate knowledge of how ESW is aligned to supply an AFW Suction source.

Associated objective(s):

(RO-C-05600-E11) List the local controls and indications available for the AFW system at the Hot Shutdown panel.

DC COOK OPS Page: 92 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 47 ID: CM-7740A Points: 1.00 Given the following conditions on Unit 2:

Unit was operating at 100%

A reactor trip occurred A transient occurred resulting in the following conditions:

2CD Emergency Diesel Generator running.

RCP #23, Circ Water Pump #21, North Hotwell, North Condensate Booster, and North Heater Drain Pumps are NOT running.

West CCP, CCW, ESW, and MDAFW Pumps are all running.

South NESW Pump is running.

East CCW, ESW, and MDAFW Pumps are all running.

North NESW Pump is running.

Which ONE of the following failures caused the transient?

A. RCP Bus 2D supply breaker tripped B. RCP Bus 2C supply breaker tripped C. Loss of ALL power to 250V DC Bus 2CD D. Bus T21D Degraded Bus Voltage Answer: A Answer Explanation:

A. Correct - When RCP Bus 2D supply breaker trips it de-energizes the #23 RCP, #21 CW, North Hotwell, North Condensate Booster, and North Heater Drain Pumps as well as the T21D bus. This will start the 2CD EDG and energize the Blackout Sequencer, loading - East CCW, ESW, NESW and MDAFW Pumps.

B. Incorrect - Plausible since 2CD EDG is running. The T21D Bus loads would not be operating and the #23 RCP, #21 CW, North Hotwell, North Condensate, and North Heater Drain Pumps would not be tripped.

C. Incorrect - Plausible since loss of DC does impact the RCP & T21 bus loads except Loss of DC would prevent T21D loads from energizing.

D. Incorrect - Plausible since Degraded Bus Voltage has auto actions to de-energize the T21D bus under certain conditions. This will start the 2CD EDG and energize the Blackout Sequencer, loading - East CCW, ESW, NESW and MDAFW Pumps. Loads on RCP bus 2D would not be affected.

DC COOK OPS Page: 93 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) CM-7740A (Active)

External Topic ID: 8200: 8200, BOP Electrical Distribution Level/Difficulty: H/3 Comments:

Reference:

2-OHP-2110-BKM-001, Figure 12, pg. 22 Source: Bank Previous NRC exam: 2007 KA - 062000 K2.01 A.C. Electrical Distribution System Knowledge of bus power supplies to the following:

Major system loads RO - 3.3 SRO - 3.4 CFR - 41.7 KA Justification - Requires knowledge of the abnormal condition caused by a loss of 4kv AC bus 2D.

Original Question # - RO28 AUDIT, CM-7740A, RO27 Audit -47, NRC EXAM 2007-56, Cook RO24 Audit -048-11, NRC Exam 2002-093 Original Question KA - 062000 K3.01, 062000 2.4.11 Associated objective(s):

(RO-C-08200-E2.) Describe the power distribution arrangement for the offsite power supply system and the balance of plant electrical system including the following line-ups. Include all transformers, voltage levels, buses, breakers, remote operated disconnect switches and supplemental diesel generators.

a. Unit 1 main generator output to the 345kV switchyard with both 345kV buses and the breaker and a half 345 kV scheme and the connecting grid lines

b. Unit 2 main generator output to the 765kV ring bus with the 765 kV breakers and the connecting grid lines

c. Preferred offsite power supply from both the 345kV and 765kV grid connections to the RCP buses with Correct 34.5 kV reserve feed breaker positions shown

d. Alternate offsite power supply from the grid connections, including the Supplemental Diesel Generators, to the 4kV safety bus (T bus) connection

e. Backfeed

f. In house balance of plant electrical distribution from the 4kV level to the 600V level DC COOK OPS Page: 94 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 48 ID: NRCAUDIT07-0699 Points: 1.00 The following conditions exist:

Unit 2 is operating at 100% power.

2CD Emergency Diesel Generator (EDG) is started for surveillance testing.

Which ONE of the following describes how a reverse power condition is prevented when closing the EDG output breaker?

A. Ensure incoming (Start) voltage is slightly higher than running (Run) voltage prior to closing the output breaker.

B. Ensure synchro scope is at 12 o'clock position prior to closing the output breaker.

C. Ensure synchro scope is rotating slowly in the 'FAST' (Clockwise) direction prior to closing the breaker.

D. Ensure running (Run) and incoming (Start) frequencies are matched prior to closing the output breaker.

Answer: C Answer Explanation:

A. Incorrect - Ensuring voltage is matched ensures no large reactive current flow on breaker closure.

B. Incorrect - The synchro scope is required to be at 12 o'clock but this does not prevent reverse power.

C. Correct - The procedure requires the Operator to ensure that the synchro scope is rotating slowly in the 'FAST' (clockwise) direction prior to closing the breaker. This causes the DG to be slightly faster than the Grid causing it to pick up some load thus preventing a reverse power trip.

D. Incorrect - Matching frequency may actually increase the chance of reverse power because the incoming machine will take less load on parallel.

Question ID (Status) NRCAUDIT07-0699(Active)

External Topic ID: 3200: 3200, Diesel Generators Level/Difficulty: F/3 Comments:

Reference:

2-OHP-4021-032-001CD DG2CD Operation Attachment 2, Section 4.2, DC COOK OPS Page: 95 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO RO-C-IF05, Motors and Generators (GP Chapter 5)

Source: Bank Previous NRC exam: 2006 KA - 062000 A4.07 A.C. Electrical Distribution System Ability to manually operate and/or monitor in the control room: Synchronizing and paralleling of different ac supplies RO - 3.1 SRO - 3.1 CFR - 41.7 / 45.5 to 45.8 KA Justification - Question requires candidate knowledge of how to synchronize the EDG with normal power.

Associated objective(s):

(RO-C-03200-E15) List the parameters used to verify proper Emergency Diesel Generator operation.

a. Frequency

b. Output Voltage

c. Phase Amps

d. Generator kW

e. Field Amps

f. Load Test Bank Amps DC COOK OPS Page: 96 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 49 ID: 2008NRC-0265 Points: 1.00 Unit 2 was operating at 100% power when a Complete Loss of Onsite and Offsite AC power occurred.

3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months later, Unit 2 dispatched operators to shed the large Non-Essential DC loads.

Given the following plant conditions:

The crew transitioned to 2-OHP-4023-ECA-0.0, Loss Of All AC Power

Power has just been restored from Emergency Power (EP).

SG pressures have been stabilized

While performing Step 36, power could NOT be restored to the battery chargers for the N train, 2AB, and 2CD 250VDC buses.

The actions of step 36 to restore 600V AC Busses and Control Room Cooling were successfully performed.

When the Batteries completely discharge:

A. all AFW flow will be lost when the AFW pump discharge valves fail closed B. the ability to start and stop ECCS pumps from the control room will be lost.

C. the Emergency Power (EP) feed breaker will trip open resulting in another Loss of AC.

D. all vital instrumentation will be lost.

Answer: B Answer Explanation:

A. Incorrect - Plausible since many AOVs will fail closed on loss of DC and the TDAFW Pump Valves use DC but will fail as is while the MDAFW valves may be operated.

B. Correct - A loss of DC control power will prevent breaker operations with the control switch (and trip functions).

C. Incorrect - Plausible since the breaker is affected by loss of DC but this breaker will not open from a loss of DC - it must receive a trip signal (overload).

D. Incorrect - Plausible since the (CRIDs) vital instruments are normally supplied from inverters that have DC backup but will remain powered from AC backup DC COOK OPS Page: 97 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO sources.

Question ID (Status) 2008NRC-0265(Active)

External Topic ID: 8201: 8201, ESS Electrical Distribution Level/Difficulty: H/3 Comments:

Reference:

RO-C-08201, Engineered Safety System Electrical Distribution System Source: Bank Previous NRC exam: 2008 KA - 063000 K3.02 D.C. Electrical Distribution System Knowledge of the effect that a loss or malfunction of the D.C. Electrical System will have on the following: Components using DC control power RO - 3.5 SRO - 3.7 CFR - 41.7 / 45.6 KA Justification - Requires the knowledge of the effect of a loss of DC will have on major loads (breaker control power) and ability of those breakers to reposition.

Associated objective(s):

(RO-C-EOP14-E9) For each of the ECA-0 series procedures, discuss the basis or reason for all Steps.

DC COOK OPS Page: 98 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 50 ID: RO26-0128 Points: 1.00 Given the following plant conditions on Unit 1:

Bus T11B normal supply breaker has opened.

DG 1AB Diesel Generator has started and is tied to the bus.

The ESF loads are sequencing on.

Which ONE of the following conditions will automatically trip the diesel generator?

A. Engine Speed of 590 rpm B. CO2 actuating in the EDG Room C. Main Bearing Temperature 198° F D. Low Lube Oil Pressure 23 psig Answer: A Answer Explanation:

A. Correct - EDG in Emergency Mode so Overspeed Trip is the only one available, 590 rpm is 114.7% of Normal 514 rpm - Trip at 110%.

B. Incorrect - CO2 is trip but not in emergency mode.

C. Incorrect - Main bearing temp of >195 is normal trip D. Incorrect - Lube oil pressure of <25 psig is normal trip Note: LOOP or SI places EDG in Emergency Mode and blocks 7 non-emergency trips.

Question ID (Status) RO26-0128(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

OHP-4024-119 ANNUNCIATOR #119 RESPONSE: STATION AUXILIARY AB - DROP 53, DG1AB TRIPS DISABLED, RO-C-03200 Emergency Diesel Generators Source: Bank Previous NRC exam: 2010 DC COOK OPS Page: 99 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO KA - 064000 K4.01 Emergency Diesel Generator (ED/G) System Knowledge of ED/G System design feature(s) and/or interlock(s) which provide for the following:

Trips while loading the ED/G (frequency, voltage, speed)

RO - 3.8 SRO - 4.1 CFR - 41.7 KA Justification - Question tests knowledge of which trips are available when EDG is started in Emergency Mode and is being loaded. (Note the only load related trip is overspeed or overcurrent.)

Question Source - RO23 Audit -059-5 (Q#54)

Associated objective(s):

(RO-C-03200-E10) Describe the conditions for the following:

a. Conditions that will cause the Diesel Engine to automatically/manually trip.

b. Conditions that will prevent the Diesel Engine from an automatic/manual trip.

c. Conditions that will cause the Diesel Engine to automatically/manually start.

d. Conditions that will prevent the Diesel Engine from an automatic/manual start.

e. Conditions that will cause an Incomplete Start signal.

f. Diesel Engine Governor response to changing speed/load.

DC COOK OPS Page: 100 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 51 ID: RO32NRC-051 Points: 1.00 While preparing the release paperwork for the Monitor tank, it is identified that the tank contains a high concentration of Tritium.

Which one of the following describes the impact on the monitored release due to the Tritium Levels?

The tank will be released A. with the same release flowrate as normally and a higher minimum dilution flowrate B. with a lower maximum release flowrate and a higher minimum dilution flowrate.

C. with a higher maximum release flowrate and with a normal setpoint on the RRS-1001A, Liquid Waste Effluent Monitor.

D. with the same release flowrate as normally and with a lowered setpoint on the RRS-1001A, Liquid Waste Effluent Monitor.

Answer: B Answer Explanation:

A. Incorrect - Whenever the tritium level is high a lower release flowrate is used.

Plausible since a higher/restricted dilution flowrate is also required during the release.

B. Correct - Whenever the tritium level is high a lower release flowrate is used. The minimum dilution flowrate is also specified and controlled.

C. Incorrect - Whenever the tritium level is high a lower release flowrate is used.

Plausible if they think the tritium is a normal occurring isotope so it may be released quicker.

D. Incorrect - Whenever the tritium level is high a lower release flowrate is used.

Plausible that the setpoint may be lowered to trigger the closure sooner.

Question ID (Status) RO32NRC-051 (Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

12-OHP-4021-006-004 TRANSFERRING DISTILLATE FROM MONITOR TANKS, Attachment 3 Step 4.4 DC COOK OPS Page: 101 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Source: New KA - 073000 K5.03 Process Radiation Monitoring (PRM) System Knowledge of the operational implications of the following concepts as they apply to the PRM System:

Relationship between radiation intensity and exposure limits RO - 2.9 SRO - 3.4 CFR - 41.5 / 45.7 KA Justification - Question requires candidate to determine the release restrictions on the PRM due to the radiation intensity of the release fluid.

Associated objective(s):

(RO-C-02200-E1) State the purposes(s) and/or functions) of the Liquid Waste system DC COOK OPS Page: 102 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 52 ID: RO32NRC-052 Points: 1.00 Unit 1 and Unit 2 were operating at 100% power with the following

Unit 1 East Essential Service Water (ESW) pump running

Unit 2 East Essential Service Water (ESW) pump running

Unit Crossties open (1-WMO-705, 1-WMO-707, 2-WMO-706, 2-WMO-708)

ESW is not aligned to any Emergency Diesel Generator (EDG) to maintain minimum ESW system flow Unit 1 Trips due to a Main Generator fault.

The following post trip conditions exist:

Reserve Transformer TR101CD is unavailable.

1CD EDG has started.

Bus T11D is energized from the EDG.

Bus T11C failed to energize from the EDG.

Which ONE of the following describes the status and or actions required for 1CD EDG cooling?

A. Normal cooling valve WMO-725 has automatically OPENED.

Alternate cooling valve WMO-727 has power and is available if required.

B. Normal cooling valve WMO-725 has automatically OPENED.

Alternate cooling valve WMO-727 is De-energized.

C. Normal cooling valve WMO-725 is De-energized.

Alternate cooling valve WMO-727 has automatically OPENED.

D. Normal cooling valve WMO-725 is De-energized.

Alternate cooling valve WMO-727 must be manually OPENED.

Answer: A Answer Explanation:

A. Correct - T11D powers the Normal EDG cooling supply and it will open on an EDG start. T11A is the power for the Alternate cooling supply and is energized.

B. Incorrect. T11D powers the Normal EDG cooling supply and it will open on an EDG start. T11A is the power for the Alternate cooling supply and is energized.

C. Incorrect - T11D powers the Normal EDG cooling supply and it will open on an EDG DC COOK OPS Page: 103 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO start. T11A is the power for the Alternate cooling supply and is energized.

Plausible is candidate thinks Normal is powered from T11C and alternate will automatically open.

D. Incorrect - T11D powers the Normal EDG cooling supply and it will open on an EDG start. T11A is the power for the Alternate cooling supply and is de-energized. This would be the response if T11D was lost.

Question ID (Status) RO32NRC-052(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

12-OHP-4021-019-001 attachment #6 Filling and Venting ESW to Unit 1 Emergency Diesel Generators Source: New KA - 076000 K2.08 Service Water System (SWS)

Knowledge of bus power supplies to the following: ESF-actuated MOVs RO - 3.1 SRO - 3.3 CFR - 41.7 KA Justification - Question requires candidate knowledge of the EDG ESW supply valve power source and operation.

Associated objective(s):

(RO-C-3201-E4) Explain how a loss of each of the support systems will affect Emergency Diesel Generator Auxiliary System operation:

a. Electrical Power System (600V AC, 250V DC, 120V AC)

b. ESW System

c. Demineralized Water System DC COOK OPS Page: 104 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 53 ID: NRC2010-62 Points: 1.00 Given the following plant conditions:

Unit 2 is in MODE 4 at 2800 F.

The Unit 2 East CCW HX is in service with 5000 GPM of ESW Flow.

The Unit 2 EAST RHR train in service with 5000 GPM of CCW Flow.

All four ESW pumps are in service

ESW crosstie valves (WMO-705/ 706/ 707/ 708) are open.

The 2CD EDG is running at a constant 3200 KW for a Surveillance test.

Which ONE of the following describes the long-term impact on the listed parameters if the Unit 2 East ESW pump trips with no operator action taken?

U2 East CCW Heat Exchanger 2CD EDG Jacket Water (JW) Heat Exchanger CCW Outlet Temperature JW Outlet Temperature A. Rises Relatively Constant B. Rises Rises C. Relatively Constant Relatively Constant D. Relatively Constant Rises Answer: A Answer Explanation:

A. Correct - ESW Flow through the CCW HX is maintained by manually throttling ESW through the CCW HX. When the East ESW pump Trips this will lower flow through the HX (since the flow is manually controlled). The Flow through the DG is maintained constant by the automatic temperature control valve.

B. Incorrect - CCW HX temperature will rise due to lower flow but DG flow and temperatures are relatively constant due to the DG temperature control Valve, plausible since the system pressure/flow will lower and student is not aware of DG temperature control valve.

C. Incorrect - CCW HX temperature will rise due to lower flow while DG flow and temperatures are relatively constant due to the DG temperature control Valve, plausible since the system pressure/flow will lower and student may DC COOK OPS Page: 105 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO think that ESW through CCW is temperature controlled and/or student may assume that the West ESW would supply flow (ESW crossties are between units not trains).

D. Incorrect - CCW HX temperature will rise due to lower flow and DG flow and temperatures are relatively constant due to the DG temperature control Valve, plausible since the system pressure/flow will lower and student may think that ESW through CCW is temperature controlled and/or student may assume that the West ESW would supply flow (ESW crossties are between units not trains).

Question ID (Status) NRC2010-62(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

OHP-4024-204 Drop 85 East CCW HX Discharge Temp Abnormal, OHP-4024-219 Drop 22 DG2AB Lube Oil Temp High Source: Bank Previous NRC exam: 2016 KA - 076000 A1.02 Service Water System (SWS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures RO - 2.6 SRO - 2.6 CFR - 41.5 / 45.5 KA Justification - Question matches KA as it tests ability to predict/monitor CCW &

DG JW parameters / temperatures associated with reduction in ESW (SWS) cooling.

Associated objective(s):

(RO-C-01900-E9) Describe the response of each component to a loss of control power or control air.

a. ESW pumps

b. ESW pump discharge valves

c. ESW pump discharge strainers

d. ESW Unit crosstie valves

e. Inlet and/or Outlet valves for the following components:

Component Cooling Water Heat Exchangers

Containment Spray Heat Exchangers

Diesel Generator Cooling (Normal and Alternate supply)

Control Room Air Conditioning Units

Auxiliary Feedwater emergency supply DC COOK OPS Page: 106 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 54 ID: RO32NRC-054 Points: 1.00 Which of the following result in DIRECTLY tripping the Plant Air Compressor as a result of the loss of NESW?

1. High first inner stage temp (>125°F)

2. High second inner-stage temp (>125°F)

3. High inlet oil temp (>150°F)

4. High Inlet Cooling Water Temperature > 120°F

5. Low Inlet Cooling Water Pressure <20 psig A. 3 and 5 ONLY B. 1,2, and 3 ONLY C. 4 and 5 ONLY D. 1, 2 and 4 ONLY Answer: B Answer Explanation:

PAC Trips Drop 3

1. High first inner stage temp (>125°F)

2. High second inner-stage temp (>125°F)

3. High inlet oil temp (>150°F)

PAC Trouble Drop 6

4. High Inlet Clg Water Temperature > 120°F

5. Low Inlet Clg Water Pressure <20 psig A. Incorrect - Low pressure is an alarm only on the PAC (a Trip on the CAC)

B. Correct - High temperatures on oil or first/second stage causes a trip C. Incorrect - These are alarms only D. Incorrect - High NESW temperature is not a trip.

Question ID (Status) RO32NRC-054(Active)

External Topic ID:

DC COOK OPS Page: 107 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Level/Difficulty: F/3 Comments:

Reference:

OHP-4024-122(222), Plant Service Drop 3: Plant Air Compressor Protection Trip Source: New KA - 078000 K4.03 Instrument Air System (IAS)

Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: Securing of SAS upon loss of cooling water RO - 3.1 SRO - 3.3 CFR - 41.7 KA Justification - Question requires candidate knowledge of the PAC (SAS) trips that come from a loss of cooling water.

Associated objective(s):

(RO-C-06401-E7) State the automatic plant and control air compressor trips in accordance with 1/2-OHP-4024-122/222 Annunciator Response: Plant Service DC COOK OPS Page: 108 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 55 ID: NRCAUDIT07-1034 Points: 1.00 Which of the following sets of components need to be manually realigned following a Phase A Containment Isolation Signal reset in accordance with 02-OHP-4022-034-003, Recovery from an Inadvertent Containment Isolation Phase A?

1) Control Air to Containment

2) Cooling Water to Containment Upper Vent Fans

3) RCP Seal Return

4) Cooling Water to Reactor Supports

5) Instrument Room Ventilation Fan

6) Containment Equalization (CEQ) Fans A. 1, 2, 3, 5 B. 1, 3, 4, 5 C. 2, 3, 4, 6 D. 1, 4, 5, 6 Answer: B Answer Explanation:

Control Air to Containment, RCP Seal Return, and Cooling Water to Reactor Supports isolate on a Phase A. The Instrument Room Ventilation Fans stop on a Phase A.

Containment Upper Vent Fans automatically restart on Phase A reset. There is no effect on Containment Equalization (CEQ) Fans from a Phase A.

A - Incorrect - Cooling Water to Containment Upper Vent Fans does not isolate on a Phase A Isolation B - Correct - See above C - Incorrect - Cooling Water to Containment Upper Vent Fans does not isolate on a Phase A Isolation. CEQ Fans start at same setpoint as a Phase A (1.1 psig) but not from the Phase A actuation relays / logic.

D - Incorrect - CEQ Fans start at same setpoint as a Phase A (1.1 psig) but not from the Phase A actuation relays / logic.

Question ID (Status) NRCAUDIT07-1034(Active)

External Topic ID:

DC COOK OPS Page: 109 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Level/Difficulty: H/3 Comments:

Reference:

02-OHP-4022-034-003, Recovery from an Inadvertent Containment Isolation Phase A Source: Bank KA - 103000 A2.03 Containment System Ability to (a) predict the impacts of the following malfunctions or operations on the Containment System and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation RO - 3.5 SRO - 3.8 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Question requires candidate knowledge of which components were impacted by the Phase A and which ones need to be positioned.

Associated objective(s):

(RO-C-AOP0520412-E2) Given a set of plant conditions including the occurrence of an Inadvertent Containment Phase "A" isolation describe the required operator actions to Correct, control, or mitigate the plant response in accordance with OHP-4022-034-003, Recovery from Inadvertent Containment Isolation Phase A.

DC COOK OPS Page: 110 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 56 ID: RO32NRC-056 Points: 1.00 Unit 2 has experienced a large LOCA with a loss of offsite power.

How would the loss of the Wide Range RVLIS indicators NLI-130 AND NLI-131 effect the performance of the Core Cooling Status Tree?

A. The Narrow Range RVLIS indicators would be used once an RCP is started.

B. No Impact since the Wide range indicators are only used with the RCPs running.

C. No Impact since the Wide range indicators are only used for Mid -Loop operations D. The UPPER Plenum RVLIS indicators would be used since they overlap the Wide range indicators.

Answer: B Answer Explanation:

A. INCORRECT- The Wide Range RVLIS is tied to the RCP running.

B. CORRECT - The Status trees check to see if the RCPs are running and then use the WR to determine actions based on level.

C. INCORRECT - The Narrow range are used during Mid-Loop operations. The wide range are used when the RCPS are running (not during Mid-Loop)

D. INCORRECT - The Upper Plenum is used when the RCPs are stopped for the Inventory Status tree. The range is limited to the vessel head.

Question ID (Status) RO32NRC-056(Active)

External Topic ID:

Level/Difficulty: F/4 Comments:

Reference:

TS 3.3.3 Source: New KA - 002000 K6.03 Reactor Coolant System (RCS)

Knowledge of the effect of a loss or malfunction on the following RCS components:

Reactor vessel level indication DC COOK OPS Page: 111 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO RO - 3.1 SRO - 3.6 CFR - 41.7 / 45.7 KA Justification - Question requires candidate to determine how a loss of a RVLIS channel effects the Critical safety Function status Trees.

Associated objective(s):

(RO-C-00200-E5) Explain the following about the Reactor Vessel Level Indication System (RVLIS):

a. how operation of Reactor Coolant Pumps (RCPs) affects each RVLIS Range b.why RVLIS wide range pressure is used preferentially when computing sub-cooling and making decisions in the Emergency Operating Procedures (EOPs) the meaning of significant RVLIS indications used in EOPs indications symptomatic of inoperability of RVLIS DC COOK OPS Page: 112 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 57 ID: RO32NRC-057 Points: 1.00 Unit 1 was at 73% with a power escalation in progress The following plant conditions exist:

Rod H-8 in Control Bank D Group 2 is at 167 steps.

The remaining Control Bank D rods are at 193 steps.

A faulty card has been replaced.

Reactor Engineering has concurred that Rod H-8 should be re-aligned with Control Bank D

The crew is implementing 1-OHP-4022-012-005, Dropped or Misaligned Rod Attachment B, Recovery by Positioning a Single Rod

After moving Rod H-8 three steps, a Control Rod Urgent Failure alarm occurs.

Continued rod withdrawal will cause:

A. ALL Bank D Group 2 IRPI AND Group 2 Demand Counter indications to indicate withdraw as the rod is recovered.

B. ONLY Bank D rod H-8 IRPI indications to indicate withdraw as the rod is recovered. Demand Counters do NOT change in Bank Select.

C. ONLY Bank D rod H-8 IRPI and Group 2 Demand Counter indications to indicate withdraw as the rod is recovered.

D. ONLY rod H-8 IRPI will indicate withdraw and NO Group 2 Demand Counter indications will indicate withdraw as the rod is recovered.

Answer: C Answer Explanation:

During rod recovery the disconnect coils are opened for all unaffected rods. This will cause the Urgent failure in the opposite power cabinet and those rods will not move.

The step counter will not change for the unaffected group only for the group that has rod movement. The IRPI indication are coils for the individual rods.

A. Incorrect - The IRPI Indicators are rod specific and will not change for the other rods in the group.

B. Incorrect - The Step Counters will indicate the steps changed. The PA counter will not.

C. Correct - Only the rod with the lift coil still closed will move along with just group 2 DC COOK OPS Page: 113 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO step counter.

D. Incorrect - The Step Counters will indicate the steps changed.

Question ID (Status) RO32NRC-057(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

1-OHP-4022-012-005 Dropped or Misaligned Rod Attachment B, 1-OHP-4024-110 Drop 26 Rod Control Urgent Failure Source: New KA - 014000 A1.02 Rod Position Indication System (RPIS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls including:

Control rod position indication on control room panels RO - 3.2 SRO - 3.6 CFR - 41.5 / 45.5 KA Justification - Question requires candidate to determine the Correct IRPI and Step Counter indications based on the rod recovery method.

Associated objective(s):

(RO-C-AOP0240412-E3) Given a set of plant conditions and the occurrence of an abnormal event, without use of references, explain the procedural mitigation strategy for Dropped Rod, Misaligned Rod, or RPI Failure in accordance with plant procedures, and standards and expectations for performance.

DC COOK OPS Page: 114 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 58 ID: NRCAUDIT07-0082 Points: 1.00 Unit 2 is operating at 100% power. The 43-TSAT-2 Thermocouple Selector Switch is selected to use a single thermocouple (Auctioneering function is NOT Working).

The following conditions exist:

Subcooling Meter indicates 425°F

RCS Thot indicates 608°F

RCS pressure indicates 2200 psig Which ONE of the following malfunctions describes:

(1) these indications (2) the means used to determine subcooling?

A. (1) A SHORT is causing the thermocouple to read HIGH.

(2) The operator should use OHP-4023-SUP-001, SubCooling Margin Determination or use PPC values.

B. (1) An OPEN is causing the thermocouple to read LOW.

(2) The operator should use OHP-4023-SUP-001, SubCooling Margin Determination or use PPC values.

C. (1) A SHORT is causing the thermocouple to read HIGH.

(2) The operator should defeat the failed thermocouple at the Incore TC Recorder.

D. (1) An OPEN is causing the thermocouple to read LOW.

(2) The operator should defeat the failed thermocouple at the Incore TC Recorder.

Answer: B Answer Explanation:

A failed OPEN TC will indicate LOW (200°F) causing the meter to read 425°F or excessive subcooling. OHP-4023-SUP-001 SubCooling Margin Determination is used whenever it is desired to determine subcooling. Selecting the T-SAT-RTD position allows the use of the RTDs for calculation. The PPC also provides indications of individual TCs and calculations that use the Average TC margin to saturation.

A. Incorrect - A short will cause the TC to indicate low (200°F). A high reading would indicate an "OL" or a negative number indicating saturation.

B. Correct - See above DC COOK OPS Page: 115 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO C. Incorrect - A short will cause the TC to indicate low (200°F). A high reading would indicate an "OL" or a negative number indicating saturation. TCs are not defeated from the recorder.

D. Incorrect - TCs are not defeated from the recorder.

Question ID (Status) NRCAUDIT07-0082(Active)

External Topic ID:

Level/Difficulty: H/4 Comments:

Reference:

RO-C-01301, Incore Nuclear Instrumentation System, RO-C-IF27A Sensors and Detectors Source: Bank Previous NRC exam: 2004 KA - 017000 A2.01 In-Core Temperature Monitor (ITM) System Ability to (a) predict the impacts of the following malfunctions or operations on the ITM System and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations: Thermocouple open and short circuits.

RO - 3.1 SRO - 3.5 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Question requires candidate to determine the failure and action for a TC Open wire.

Associated objective(s):

(RO-C-01301-E12) Given a description of plant conditions and/or the results of a Surveillance Test, determine the following:

a. All applicable Technical Specifications

b. System Operability

c. Most limiting LCO DC COOK OPS Page: 116 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 59 ID: NRC2010-59 Points: 1.00 Which ONE of the following lists the Unit 1 Control Room Ventilation system damper alignment for operation during a high alarm on ERS-7401, U1 Control Room Area Radiation Monitor?

1-HV-ACR-DA-1/1A 1-HV-ACR-DA-2 1-HV-ACR-DA-2A 1-HV-ACR-DA-3 Outside air to CR Outside air to CR PRZN Outside air to CR PRZN CR air to PRZN A. OPEN CLOSED PARTIAL OPEN OPEN B. CLOSED PARTIAL OPEN CLOSED OPEN C. OPEN PARTIAL OPEN CLOSED CLOSED D. CLOSED CLOSED PARTIAL OPEN CLOSED Answer: B Answer Explanation:

A. Incorrect - Dampers 1/1A will be closed on an ERS-7401 high alarm.

B. Correct - On an ERS-7401 high alarm: Damper 1/1A will be closed, Damper 2 will be partially open, Damper 3 opens.

C. Incorrect - Damper 1/1A will be closed and Damper 3 will remain open on an ERS-7401 high alarm.

D. Incorrect - Damper 3 will remain open on an ERS-7401 high alarm.

Question ID (Status) NRC2010-59(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

12-OHP-4024-139, 1-ERS-7400 Control Room, Inst. Room, Aux Building Area Monitors Source: Bank Previous NRC exam: 2010 KA - 072000 A3.01 Area Radiation Monitoring (ARM) System DC COOK OPS Page: 117 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Ability to monitor automatic operation of the ARM system, including:

Changes in ventilation alignment RO - 2.9 SRO - 3.1 CFR - 41.7 / 45.5 KA Justification - Question tests ability to monitor changes in the Control Room Ventilation dampers caused by a high radiation alarm.

Original Question # - NRC 2010 Exam MASTER 01028C01A02-6, NRC EXAM 2007-19 Original Question KA - 000067 AA2.02 Associated objective(s):

(RO-C-2801A-E8) Describe the conditions that cause the following components to start, trip or reposition:

a. Pressurization fans

b. Chiller packages

c. Remotely operated dampers (ACR-DA-1, 1A, 2, 2A, 3)

DC COOK OPS Page: 118 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 60 ID: RO26-0077-SRO Points: 1.00 The crew has just completed actions of 1-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink that established bleed and feed.

Given the following plant conditions:

SG #13 is faulted with ZERO indicated level on SG Wide Range Level.

SG #11, 12, and 14 are intact and SG Wide Range levels are 14%.

Core Exit Thermocouples are 570°F (hottest) and slowly lowering.

RCS Hot Leg temperatures are 565°F and slowly lowering.

RCS Wide Range Pressure is 1200 psig and stable

Containment pressure is 0.8 psig and rising.

The Turbine-driven AFW Pump was just made available.

What action is taken to establish a secondary heat sink?

A. Initiate feed flow to SG #11, 12, OR 14 at less than 50 x 103 lbm/hr.

B. Initiate feed flow to SG #11, 12, OR 14 at the maximum rate until RCS Subcooling is > 40°F.

C. Initiate feed flow to SG #11, 12, AND 14 at the maximum rate until RCS Subcooling is > 40°F.

D. Initiate feed flow to SG #11, 12, AND 14 at less than 50 x 103 lbm/hr.

Answer: A Answer Explanation:

A. Correct - Since core exit TCs are lowering and there is concern with thermal stress to hot, dry SG, procedural direction is to initiate flow in controlled flow rate (50 X 103 lbm/hr) to provide heat removal while minimizing stresses.

B. Incorrect - Since core exit TCs are lowering and there is concern with thermal stress to hot, dry SG, procedural direction is to initiate flow in controlled flow rate (50 X103 lbm/hr) to provide heat removal while minimizing stresses. This answer would be Correct if core exit TCs were rising C. Incorrect - Feed Flow should only be established to ONE SG. Since core exit TCs are lowering and there is concern with thermal stress to hot, dry SG, procedural direction is to initiate flow in controlled flow rate (50 X103 lbm/hr) to provide heat removal while minimizing stresses.

D. Incorrect - Feed Flow should only be established to ONE SG.

DC COOK OPS Page: 119 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) RO26-0077-SRO(Active)

External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: H/3 Comments:

Reference:

OHP-4023-FR-H.1 Response to Loss of Secondary Heat Sink (Foldout Page item #4, Step 1 Caution, & Step 33)

Source: Bank Previous NRC exam: 2012 KA - 035000 A4.02 Steam Generator System (S/GS)

Ability to manually operate and/or monitor in the control room: Fill of dry S/G RO - 2.7 SRO - 2.8 CFR - 41.7 / 45.5 to 45.8 KA Justification - Question addresses Loss of Heat Sink and the operators ability to monitor the plant conditions and manually control AFW flow to fill the hot dry SGs.

Original Question # - NRC EXAM 2012, RO26-0077, RO26 AUDIT-62 Original Question KA - 00WE05 EA1.3 Associated objective(s):

(RO-C-EOP11-E10) For each of the FR-H series procedures discuss the basis or reason for all Steps.

NOTE: The following is a list of steps that are considered essential to understanding the purpose and overall strategy contained in the FR-H series procedures.

FR-H.1

Determination if secondary heat sink is required

Determination if bleed and feed is required (Loss of Heat Sink)

Establishing AFW flow, MFW flow and Condensate flow

Tripping RCPs

Verifying adequate RCS feed path

Verifying adequate RCS bleed path

Starting all available charging pumps

Feeding a dry S/G

Determining if SGs being fed are intact

Determining if one HHSI pump should be stopped

Determining if a transition to E-1 is required

Determining if a transition to ES-1.1 is required DC COOK OPS Page: 120 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 61 ID: CM-0427A Points: 1.00 Given the following initial conditions:

Unit 2 is at 70% power and stable

All controls are in Normal or Automatic position Annunciator Panel 216 Drop 92, Heater Drain Pump Motor Overload AND Drop 93, Heater Drain Pump Discharge Pressure Low, alarms are in. The South Heater Drain Pump indicates stopped.

Which ONE of the following describes (1) the actions that will occur and (2) where it will be verified?

A. (1) 2-HRV-461, Heater 4A Alternate Drain to Condenser Hotwell OPENS.

(2) Verified on the Feedwater Heaters / MSR Panel Bailey controller.

B. (1) 2-HRV-461, Heater 4A Alternate Drain to Condenser Hotwell OPENS.

(2) Verified on the Feedwater Heater HMI.

C. (1) 2-HRV-462, Heater 4B Alternate Drain to Condenser Hotwell OPENS.

(2) Verified on the Feedwater Heaters / MSR Panel Bailey controller.

D. (1) 2-HRV-462, Heater 4B Alternate Drain to Condenser Hotwell OPENS.

(2) Verified on the Feedwater Heater HMI.

Answer: D Answer Explanation:

The Trip of the South Heater Drain pump on Unit 2 will cause the Level to rise in the 4B heater which will open the alternate drain. This valve is controlled by the HMI panel in Unit 2. These valves were previously controlled by the bailey controllers and the other low-pressure heaters still are. The Unit 1 Heater drain is aligned to the 5 heater.

A. Incorrect - The controls are on HMI and the 4B heater is affected.

B. Incorrect - The 4B heater is affected.

C. Incorrect - The controls are on HMI.

D. Correct - HRV-462 is controlled from the HMI and it should start to throttle open.

Question ID (Status) CM-0427A(Active)

DC COOK OPS Page: 121 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

OHP-4024-216 DROP 92 Heater Drain Pump Motor Overload Trip Source: Modified KA - 056000 2.1.31 Condensate System Conduct of Operations Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

RO - 4.6 SRO - 4.3 CFR - 41.10 / 45.12 KA Justification - Question requires candidate to determine location and valve that need to be monitored due to the HD pump trip.

Modified Changed Stem to Just list trip of Heater Drain Pump and eliminate CD/Hotwell Alignments. Also changed to ask where the valve position would be checked.

Changed distractors A and C from CRV224 and Middle HD pump to the 461/462 valves and Bailey controller options. Add HMI to distractors B & D.

Associated objective(s):

(RO-C-0600270412-E2) Given a set of plant conditions including a Feedwater Heater Level Control Failure describe the required operator actions to Correct, control, or mitigate the plant response in accordance with OHP-4022-IFR-001, Instrument Failure Response.

DC COOK OPS Page: 122 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 62 ID: RO26-0169 Points: 1.00 Given the following plant conditions:

You are the Unit 1 RO.

A release of the #7 Gas Decay Tank is in progress.

The Auxiliary Building Exhaust Fan status is as follows:

1-HV-AX-1 Running 1-HV-AX-2 Off 2-HV-AX-1 Running 2-HV-AX-2 Running

Auxiliary Building Exhaust Fan 1-HV-AX-1 Trips.

Which ONE of the following describes your response concerning the release due to 1-HV-AX-1 tripping?

A. Notify the WDS operator to VERIFY that RRV-306, Waste Gas Decay Tank Release Valve has AUTOMATICALLY tripped closed.

B. Notify Unit 2 to monitor the release since it is all going out the Unit 2 Vent Stack through the 1-HV-AX-VD-3, Aux Building Ventilation Exhaust Plenums Crosstie Damper.

C. Instruct the Unit 1 Aux Tour operator to close 1-HV-AX-VD-3, Aux Building Ventilation Exhaust Plenums Crosstie Damper to direct the release through the Unit 2 Vent stack.

D. Notify the WDS operator that he must MANUALLY close RRV-306, Waste Gas Decay Tank Release Valve since dilution flow has been reduced.

Answer: A Answer Explanation:

A. Correct - Loss of all Unit 1 exhaust fans will cause a closure of RRV-306.

B. Incorrect HV-AX-VD-3, Aux Building Ventilation Exhaust Plenums Crosstie Damper is required to be closed while a GDT release is in progress. The GDTs discharge to the Unit 1 Vent Stack.

C. Incorrect HV-AX-VD-3, Aux Building Ventilation Exhaust Plenums Crosstie Damper is required to be closed while a GDT release is in progress. The DC COOK OPS Page: 123 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO GDTs discharge to the Unit 1 Vent Stack.

D. Incorrect - The RRV-306 is interlocked to automatically close on a loss of Unit 1 exhaust fans.

Question ID (Status) RO26-0169(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

12-OHP-4021-023-002, Release of Radioactive Waste from Gas Decay Tanks Att. 2 Source: Bank Previous NRC exam: 2014 KA - 071000 K1.04 Waste Gas Disposal System (WGDS)

Knowledge of the physical connections and/or cause-effect relationships between the Waste Gas Disposal System and the following systems: Station ventilation RO - 2.7 SRO - 2.8 CFR - 41.2 to 41.9 / 45.7 to 45.8 KA Justification - Question Requires candidate knowledge of how the ventilation system impact the waste gas release.

Question Source -NRC Exam 2014-23, Cook NRC Exam 2002-121-2, Master Bank AS07-17 Associated objective(s):

(RO-C-02300-E6) For the following list of systems/components, that interface with and/or provide a support function for the Gaseous Waste Disposal System, describe the purpose of each interface and/or its support function.

a. Component Cooling Water

b. Auxiliary Building Ventilation System

c. Compressed Gas (Nitrogen)

DC COOK OPS Page: 124 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 63 ID: NRCAUDIT07-0440A Points: 1.00 The following conditions exist on Unit 2:

Reactor status: Tripped from 100% power

Reactor trip breakers: OPEN

Turbine Stop Valves: Three CLOSED, One OPEN

Auto Stop oil pressure: 0 psig

Circ water pumps: 2 running

Condenser vacuum: 16.0 inches of Hg vacuum

RCS Tavg: 550° F The Steam Dump System:

A. IS dumping steam to the condensers and maintaining temperature on the Turbine Trip Controller.

B. IS NOT dumping steam to the condensers because of a blocking signal.

C. IS dumping steam to the condensers and maintaining temperature on the Load Rejection Controller.

D. IS NOT dumping steam to the condenser because it is NOT required to reduce temperature at the conditions given.

Answer: B Answer Explanation:

A. Incorrect - Plausible because student may not realize that the vacuum is too low to allow steam dump operation. Two Circ water pumps are running and only 1 is required, but vacuum must be > 19 inches HG.

B. CORRECT - Group 1 would be modulated open since temperature is > 547 Tref. A blocking signal from C-9 prevents the steam dump valves from opening with vacuum < 20.6 inches HG C. Incorrect - C-9 is NOT met; steam dump valves are blocked. Plausible if student does not correctly apply the Condenser Vacuum setpoint and does not realize that auto stop oil also feeds the turbine trip controller.

D. Incorrect - Plausible because student may not realize the turbine trip controller (C-8) is active with auto stop oil pressure at zero. C-9 is NOT met; steam dump valves are blocked. If the system was on the Load Rejection controller DC COOK OPS Page: 125 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO the Steam dumps would be closed since temperature is only 3 degrees high (5 degrees Question ID (Status) NRCAUDIT07-1019(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

SOD-05700-001 Source: Bank KA - 075000 K3.07 Circulating Water System Knowledge of the effect that a loss or malfunction of the Circulating Water System will have on the following: ESFAS RO - 3.4 SRO - 3.5 CFR - 41.7 / 45.6 KA Justification - Question requires candidate knowledge of the effect that the CW system loss of 2 CW pumps and vacuum has on ESFAS system (C9)

Associated objective(s):

(RO-C-05700-E6) Describe the function of the following Circulating Water System major components.

a. Intake Cribs and Intake Piping

b. Screen House Forebay and Bar Grills

c. Circulating Water Pumps

d. Circulating Water Crosstie Header

e. Intake Tunnels

f. Main Condensers Water Side

g. Main Feedpump Turbine Condensers Water Side

h. Turbine Auxiliary Cooling Water

i. Discharge Tunnel

j. WMO 15, 16, & 17 (Unit 1), WMO 25A, 25B, 26 & 27 (Unit 2)

k. WMO-30

l. De-icing Tunnel DC COOK OPS Page: 126 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 64 ID: 2008NRC-0641 Points: 1.00 Given the following plant conditions:

Unit 2 Plant Air Compressor (PAC) is running.

Unit 1 Plant Air Compressor (PAC) is out of service for maintenance.

Both Units' Control Air Compressors (CACs) are stopped and in AUTO.

A large leak occurs on the piping connecting the Unit 1 Plant Air Receiver to the Unit 1 side of the plant air header.

Plant Air Header pressure is slowly lowering.

Which ONE of the following statements describes:

(1) The method for isolating the leak?

(2) The Control Air Supply System status following isolation?

A. (1) Close PRV-10 OR PRV-11, Plant Air Header Crosstie Valves to Unit 2 (2) Control Air for both units will be supplied from the Unit 2 PAC B. (1) Close PRV-10 AND PRV-11, Plant Air Header Crosstie Valves to Unit 2 (2) Control Air for both units will be supplied from the Unit 2 PAC C. (1) Close PRV-10 AND PRV-11, Plant Air Header Crosstie Valves to Unit 2 (2) Unit 1 Control Air will be supplied by the Unit 1 Control Air Compressor AND Unit 2 Control Air remains supplied from the Unit 2 PAC D. (1) Close PRV-10 OR PRV-11, Plant Air Header Crosstie Valves to Unit 2 (2) Unit 1 Control Air will be supplied by the Unit 1 Control Air Compressor AND Unit 2 Control Air will be supplied from the Unit 2 CAC Answer: B Answer Explanation:

A. Incorrect - Both PRV-10 AND PRV-11 must be closed to isolate the Unit 1 header.

B. Correct - Both PRV-10 AND PRV-11 must be closed to isolate the Unit 1 header.

Both control air supplies come downstream of the header crosstie valves, so U2 PAC will continue to supply both units' control air headers.

C. Incorrect - Both control air supplies come downstream of the header crosstie valves, so U2 PAC will continue to supply both units' control air headers.

D. Incorrect - Both PRV-10 AND PRV-11 must be closed to isolate the Unit 1 header.

DC COOK OPS Page: 127 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Both control air supplies come downstream of the header crosstie valves, so U2 PAC will continue to supply both units' control air headers.

Question ID (Status) 2008NRC-0641(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

OHP-4022-064-001 Control Air Malfunction, SOD-06401-002, Plant Air System Source: Bank Previous NRC exam: 2014 KA - 079000 K4.01 Station Air System (SAS)

Knowledge of SAS design feature(s) and/or interlock(s) which provide for the following: Cross-connect with IAS RO - 2.9 SRO - 3.2 CFR - 41.7 KA Justification - Question tests knowledge of interrelations for the piping supplies between the plant air (SAS) and the control air systems (IAS).

Associated objective(s):

(RO-C-06401-E11) List the parameters used to verify proper operation of the Compressed Air System including:

a. Turbine Building Plant Air Header Pressure Indicator (PPI-10,20)

b. Plant Air Header Total Flow (PFR-1)

DC COOK OPS Page: 128 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 65 ID: RO26-0101 Points: 1.00 A fire has been reported in the Unit 1 Control Room Cable Vault (CRCV).

Which ONE of the following describes the actions the Control Room Operator must take?

A. 1. Verify automatic water spray into the CRCV.

2. Verify personnel out of the CRCV.

3. Manually actuate CO2 discharge into the CRCV.

4. After seven (7) minutes, manually re-actuate CO2 discharge.

B. 1. Verify automatic Halon actuation into the CRCV.

2. Verify personnel out of the CRCV.

3. Manually actuate water spray into the CRCV.

4. After seven (7) minutes, manually re-actuate water spray.

C. 1. Verify automatic Halon actuation into the CRCV.

2. Verify personnel out of the CRCV.

3. Manually actuate CO2 discharge into the CRCV.

4. After seven (7) minutes, manually re-actuate CO2 discharge.

D. 1. Verify automatic CO2 actuation into the CRCV.

2. Verify personnel out of the CRCV.

3. Manually actuate Halon discharge into the CRCV.

4. After seven (7) minutes, manually re-actuate Halon discharge.

Answer: C Answer Explanation:

A. Incorrect - Plausible as water spray does exist in Unit 2. There is NO water spray in Unit 1 CRCV.

B. Incorrect - Plausible as water spray does exist in Unit 2. There is NO water spray in Unit 1 CRCV.

C. Correct - Halon will automatically actuate, followed by a manual CO2 actuation per 1-OHP-4024-101, Drop 56.

D. Incorrect - Plausible as these are the correct types of fire protection. However, in the wrong order and using the wrong type of actuation method.

DC COOK OPS Page: 129 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) RO26-0101(Active)

External Topic ID: AS19: AS19, Miscellaneous Fire Protection Level/Difficulty: F/3 Comments:

Reference:

1-OHP-4024-101, Drop 56 & 57 Source: Bank KA - 086000 K5.04 Fire Protection System (FPS)

Knowledge of the operational implications of the following concepts as they apply to the Fire Protection System: Hazards to personnel as a result of fire type and methods of protection RO - 2.9 SRO - 3.5 CFR - 41.5 / 45.7 KA Justification - Question tests knowledge of the suppression methods for the cable spreading room, the personnel concerns and the order of execution to put the fire out.

Original Question # - RO28 AUDIT, RO26-0101, RO26 AUDIT-74, RO23 Audit 075-5 (RO#069 /SRO#069)

Original Question KA - 194001 2.4.25, SYS 086 K4.06 Associated objective(s):

(RO-C-AS18-E3) Describe three methods used to initiate a CO2 discharge into a protected area.

(RO-C-AS19-E3) State the fire protection systems used to protect the control room cable vault in both Unit 1 and 2 and describe how each is activated.

DC COOK OPS Page: 130 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 66 ID: CM-3169 Points: 1.00 Which ONE of the following non-licensed individuals is permitted to manipulate the controls of the reactor?

A. A person enrolled in the license training program under the direct supervision of the licensed reactor operator.

B. Any STA who is directed by the reactor operator or shift manager to perform specific panel operations.

C. Any operations department personnel under the direct supervision of the licensed reactor operator.

D. An Instrument Technician performing an alignment in accordance with an approved procedure.

Answer: A Answer Explanation:

PMP-4043-EQC-001 Section 3.6.2 The following guidelines shall be used by on duty Control Room Operators when allowing personnel access to the Control Panels:

a. No personnel except a licensed Senior Reactor Operator, Reactor Operator, or individual enrolled as part of the licensee training program to qualify for an Operators license under the direct supervision of a licensed Reactor Operator, are allowed to manipulate controls that affect reactivity or the power level of a nuclear reactor.

A. Correct - Only an SRO, RO or an ILT candidate under supervision can manipulate controls.

B. Incorrect - STA is not allowed to manipulate reactor controls.

C. Incorrect - personnel must be in an approved training program.

D. Incorrect - They may be able to perform work and manipulate controls that do NOT impact reactivity.

Question ID (Status) CM-3169(Active)

External Topic ID: ADMN-14: ADMN-14, Standards Level/Difficulty: F/2 Comments:

DC COOK OPS Page: 131 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

PMP-4043-EQC-001 Equipment Control Sec. 3.6 Source: Bank KA - 194001 2.1.2 Generic Conduct of Operations Knowledge of operator responsibilities during all modes of plant operation.

RO - 4.1 SRO - 4.4 CFR - 41.10 / 45.13 KA Justification - Question tests knowledge of operator responsibility for control panel operations.

Associated objective(s):

(RO-C-ADM14-E8) Given PMP-4010-CRC-001, Control Room Conduct, PMP-4043-EQC-001, Equipment Control, and OHI-4000 Attachments 3, Command and Control, Attachment 6, Control Room Conduct, and Attachment 7, Control Board Monitoring, Describe the administrative requirements placed on the following aspects of Control Room Conduct:

A. General requirements B. Command and Control function C. Meals in the Control Room D. Control board awareness DC COOK OPS Page: 132 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 67 ID: NRC2010-68 Points: 1.00 OHI-4000 Attachment 5, Conservative Decision Making states that:

"WHEN faced with unexpected or uncertain conditions, THEN personnel must promptly identify a transition point at which efforts to keep the unit on-line or on schedule are no longer conservative, nor reasonable."

Once this point is reached:

A. the reactor must be tripped immediately.

B. senior management must be notified to determine course of action.

C. actions must be taken to place the unit in a safe condition without hesitation.

D. the NRC must be notified, and actions taken to address the problem within one hour.

Answer: C Answer Explanation:

A. Incorrect - tripping the reactor is only one of the options available and may not be the prudent choice based on the transition point determined.

B. Incorrect - Action must be taken without hesitation.

C. Correct - OHI-4000, Att. 5, Step 3.4 states "WHEN faced with unexpected or uncertain conditions, THEN personnel must promptly identify a transition point at which efforts to keep the unit on-line or on schedule are no longer conservative, nor reasonable. Once this point is reached, actions to place the unit in a safe condition by reducing power, tripping the reactor, or suspending core alterations must be taken without hesitation.

D. Incorrect - NRC Notification is not required and action must be taken without hesitation.

Question ID (Status) NRC2010-68(Active)

External Topic ID:

Level/Difficulty: F/2 Comments:

Reference:

OHI-4000, Conduct of Operations: Standards Attachment 5, DC COOK OPS Page: 133 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Conservative Decision Making Source: Bank Previous NRC exam: 2010 KA - 194001 2.1.39 Generic Conduct of Operations Knowledge of conservative decision-making practices.

RO - 3.6 SRO - 4.3 CFR - 41.10 / 43.5 / 45.12 KA Justification - Question tests Knowledge of conservative decision-making practices.

Associated objective(s):

(RO-C-ADM14-E3) Given PMI-4010, Plant Operations Policy, and OHI-4000 , Conservative Decision Making, Define conservative decision making and describe the types of events which would be considered as conservative decision making.

DC COOK OPS Page: 134 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 68 ID: CM-7855 Points: 1.00 Unit 1 is in Mode 6 and in the process of core offload.

Which ONE of the following limiting conditions requires immediate suspension of all CORE ALTERATIONS?

A. Source Range Channel N32 is INOPERABLE. All other Source Range Channels are OPERABLE with the Audio Count Rate selected to Source Range Channel N31.

B. Loss of direct communications between control room and personnel at the refueling station.

C. Loss of all Fuel Handling Area Supply Fans.

D. Time since entering mode 3 is 155 hours0.00179 days 0.0431 hours 2.562831e-4 weeks 5.89775e-5 months .

Answer: B Answer Explanation:

OHP-4030-127-037 Data Sheet 2 A. Incorrect- Tech Spec 3.9.2 requires 2 operable source range channels. Only two out of four source range monitors are required OPERABLE. Must have N-31 or N-32 (or N-21, if 1-TM-17-10-R0 installed) OPERABLE for audible source range indication.

B. Correct - TRS 8.9.1.1 requires verification of direct communication between Control Room and refueling station C. Incorrect- Fuel Handling Ventilation is required for movement in SFP area not containment (Unlatch requires no SFP crane movements)

D. Incorrect - TRM 8.9.2 requires the reactor be subcritical greater than or equal to 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months during movement of irradiated fuel assemblies in the reactor pressure vessel.

Question ID (Status) CM-7855(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

DC COOK OPS Page: 135 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

Reference:

Admin TRO 8.9.1, 01-OHP-4030-127-037, Refueling Surveillance Data Sheet 2 Source: Bank Previous NRC exam: 2006 KA - 194001 2.1.41 Generic Conduct of Operations Knowledge of the refueling process.

RO - 2.8 SRO - 3.7 CFR - 41.2 / 41.10 / 43.6 / 45.13 KA Justification - Question tests RO Knowledge of refueling process and requirements.

Associated objective(s):

(RO-C-ADM13-E3) Given a set of plant conditions and the operations refueling surveillance procedure(s) determine if refueling operations can be initiated or continue without error.

DC COOK OPS Page: 136 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 69 ID: NRCAUDIT07-0063 Points: 1.00 An AEO performing rounds identifies that the TDAFW pump discharge valves for Unit 1 are throttled while the Unit 2 valves are fully open. The AEO requests permission to reposition the Unit 1 valves to be fully open.

Which ONE of the following describes the required response?

A. Change the Unit 1 valve positions to the FULL OPEN position.

B. Do NOT change Unit 1 valve positions. The Unit 2 valves should be placed in the THROTTLED position.

C. Change the Unit 1 valve positions to the FULL OPEN position and change the Unit 2 valves to the THROTTLED position.

D. Do NOT change valve positions on either unit. The valves are correct as positioned.

Answer: D Answer Explanation:

A. Incorrect - The Unit 1 valves should be THROTTLED due to SG overfill concerns.

B. Incorrect - The Unit 2 valves should be FULL OPEN.

C. Incorrect - The Unit 1 valves should be THROTTLED due to SG overfill concerns.

The Unit 2 valves should be FULL OPEN.

D. Correct - The Unit 1 valves should be THROTTLED due to SG overfill concerns. The Unit 2 valves should be FULL OPEN.

Question ID (Status) NRCAUDIT07-0063(Active)

External Topic ID: 5600: 5600, Auxiliary Feedwater Level/Difficulty: F/3 Comments:

Reference:

01(02)-OHP-4021-056-002, Auxiliary Feed Pump Operation Attachment 1 Source: Bank Previous NRC exam: 2004 KA - 194001 2.2.3 DC COOK OPS Page: 137 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Generic Equipment Control (multi-unit license) Knowledge of the design, procedural, and operational differences between units.

RO - 3.8 SRO - 3.9 CFR - 41.5 / 41.6 / 41.7 / 41.10 / 45.12 KA Justification - Question requires candidate to know the unit differences between TD AFP valve alignments.

Associated objective(s):

(RO-C-05600-E2) State the design function of the AFW system and the condensate storage tank.

DC COOK OPS Page: 138 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 70 ID: RO-C-ADM07-E2-1 Points: 1.00 If blocking an Annunciator alarm that is the direct result of the performance of a procedure, THEN a blocked alarm review and subsequent logging of the blocked alarm is NOT required in accordance with PMP-4043-APC-001, Abnormal Position Control provided which combination of statements is true?

1. The procedure identifies alternate alarms that can be used to monitor the system(s) for which the alarm to be blocked is associated with.

2. The procedure specifies what annunciators are to be blocked.

3. The procedure has a sign-off stating all needed compensatory actions have been placed in effect.

4. The procedure has steps specifying verification of the initiation AND restoration of the annunciator blocks.

5. The procedure specifies what compensatory measures must be placed in effect prior to blocking the alarm (if none are required, the procedure must so state).

6. The procedure directs the annunciator is to be blocked only after the first time it is received.

A. 1,2,3,5 B. 1,3,4,5 C. 2,3,4,6 D. 2,3,4,5 Answer: D Answer Explanation:

From PMP-4043-APC-001 IF the alarming annunciator is the direct result of the performance of a procedure, THEN a blocked alarm review and subsequent logging of the blocked alarm is NOT required provided:

The procedure specifies what annunciators are to be blocked (2).

The procedure specifies what compensatory measures must be placed in effect prior to blocking the alarm (if none are required, the procedure must so state) (5).

DC COOK OPS Page: 139 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO

The procedure has a sign-off stating all needed compensatory actions have been placed in effect (3).

The procedure has steps specifying verification of the initiation AND restoration of the annunciator blocks (4).

A. Incorrect - Item 1 is not required.

B. Incorrect - Item 1 is not required.

C. Incorrect - Item 6 is not required D. Correct - See above Question ID (Status) RO-C-ADM07-E2-1(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

PMP-4043-APC-001, Abnormal Position Control Source: Bank KA - 194001 2.2.43 Generic Equipment Control Knowledge of the process used to track inoperable alarms.

RO - 3.0 SRO - 3.3 CFR - 41.10 / 43.5 / 45.13 KA Justification - Question matches KA as candidate is required to demonstrate knowledge of process for blocking and tracking inoperable alarms IAW station procedures.

Associated objective(s):

(RO-C-ADM07-E2) Given PMP-4043-APC-001, describe the requirements to block an alarm.

DC COOK OPS Page: 140 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 71 ID: CM-39599 Points: 1.00 Given the following events and conditions:

Unit 2 is responding to a Large Break LOCA with 10% failed fuel.

2-ICM-305, Recirc Sump to East RHR / CTS Pumps isolation valve would NOT open during transfer to cold leg recirculation.

2-ICM-305 breaker has tripped and will NOT reset

RP projects that expected dose rates in the area of 2-ICM-305 will be very high, possibly exceeding 50 Rem/hr.

The Site Emergency Director (SED) has determined that manual alignment of 2-ICM-305 is justified for lifesaving protection of the general public inside the 10 mile EPZ (public health and safety).

A team of voluntary workers, who are fully aware of the risks involved and have been approved for the allowable dose extensions, are prepared to establish conditions to manually open 2-ICM-305.

Which ONE of the following exposure limits would apply for the voluntary workers?

A. Do NOT exceed 5 Rem TEDE.

B. Do NOT exceed 10 Rem TEDE.

C. Do NOT exceed 25 Rem TEDE.

D. The workers may exceed 25 Rem TEDE.

Answer: D Answer Explanation:

A. Incorrect - This is the maximum allowable TEDE for non-emergency conditions.

B. Incorrect - This is the maximum allowable annual TEDE for equipment protection during accident conditions.

C. Incorrect - This is the maximum allowable annual TEDE for life saving or public safety on a non-volunteer basis.

D. Correct - Greater than 25 Rem is allowed on a volunteer basis for protection of the public health and safety.

DC COOK OPS Page: 141 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) CM-39599(Active)

External Topic ID: GFRP: GFRP, Generic Fundamentals - Radiation Protection Level/Difficulty: F/2 Comments:

Reference:

RMT 2080-FRM-011 Instructions for Dose Extension and Authorization Source: Bank KA - 194001 2.3.4 Generic Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions.

RO - 3.2 SRO - 3.7 CFR - 41.12 / 43.4 / 45.10 KA Justification - Candidate is required to demonstrate knowledge of emergency radiation limits.

Associated objective(s):

(RO-C-RP02-E6) State the administrative dose levels of RMT-2080-TSC-001, Activation and Operation of the TSC:

a) Protecting Valuable Equipment b) Protection Large Population Groups c) Life Saving Actions DC COOK OPS Page: 142 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 72 ID: RO32 NRC-072 Points: 1.00 An Electronic Dosimeter (Self Reading Dosimeter), for a watch stander in the Auxiliary Building, will measure what kind(s) of radiation?

A. Gamma only B. Neutron and Beta only C. Gamma and Beta only D. Neutron, Beta, and Gamma Answer: A Answer Explanation:

A. Correct - The ED-SRD measures Gamma only.

B. Incorrect - The ED-SRD measures Gamma only. Plausible since Neutron is a major concern around the reactor and both Beta and Gamma are measured by personal DLR's.

C. Incorrect - The ED-SRD measures Gamma only. Plausible as both Beta and Gamma are measured by personal DLR's.

B. Incorrect - The ED-SRD only measures Gamma only. Plausible since Neutron is a major concern around the reactor and both Beta and Gamma are measured by personal DLR's.

Question ID (Status) RO32NRC-072 (Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

DMC-3000 Personal Electronic Dosimeter Manual Source: New KA - 194001 2.3.15 Generic Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

DC COOK OPS Page: 143 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO RO - 2.9 SRO - 3.1 CFR - 41.12 / 43.4 / 45.9 KA Justification - Candidate is required to demonstrate knowledge of personnel monitoring equipment.

Associated objective(s):

(RO-C-RP03-E7) Identify the detector best suited for survey or radiation monitoring.

DC COOK OPS Page: 144 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 73 ID: RO26-0033-SRO Points: 1.00 Given the following conditions on U2:

The crew has entered OHP-4022-016-004, Loss of CCW do to a loss of all CCW.

The crew tripped the reactor.

Which ONE of the following describes the correct procedural implementation?

A. Trip the RCPs after reactor trip has been verified.

OHP-4022-016-004, Loss of CCW, may be performed concurrently after the immediate actions are complete.

B. Trip the RCPs after the immediate actions of OHP-4023-E-0 are complete.

OHP-4022-016-004, Loss of CCW, is NOT needed since the EOP network addresses a loss of CCW.

C. Trip the RCPs after the immediate actions of OHP-4023-E-0 are complete.

Steps from OHP-4022-016-004, Loss of CCW, may NOT be performed until completion of OHP-4023-ES-0.1, Reactor Trip Response.

D. Implement OHP-4022-016-004, Loss of CCW, until CCW is restored.

Trip the RCPs after reactor trip has been verified.

Perform OHP-4023-E-0, Reactor Trip or Safety Injection, steps as time allows.

Answer: A Answer Explanation:

A. Correct - OHI-4023, Abnormal/Emergency Procedure User's Guide allows Abnormal Procedures to be implemented concurrently with Emergency Procedures after the immediate actions are complete. The RCPs are tripped after the reactor trip has been verified per the Note in OHP-4022-016-004.

B. Incorrect - Performance of OHP-4023-E-0 is required upon the reactor trip, but the operators must continue to perform OHP-4022-016-004 to address the loss of CCW.

C. Incorrect - User's Guide allows Abnormal Procedures to be implemented concurrently with Emergency Procedures.

D. Incorrect - The Unit Supervisor should direct action of OHP-4023-E-0, first, NOT as time allows. OHP-4023-E-0 actions take priority over OHP-4022-016-004.

Question ID (Status) RO26-0033-SRO(Active)

DC COOK OPS Page: 145 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO External Topic ID: PROC-1: PROC-1, Procedure Usage Level/Difficulty: F/3 Comments:

Reference:

OHI-4023 Abnormal/Emergency Procedure User's section 4.6.10, 2-OHP-4022-016-004 step 2 & Note Source: Bank Previous NRC exam: 2014 KA - 194001 2.4.11 Generic Emergency Procedures/Plan Knowledge of abnormal condition procedures.

RO - 4.0 SRO - 4.2 CFR - 41.10 / 43.5 / 45.13 K/A Justification - Required the knowledge of the strategies for implementing the EOPs coincident with a Loss of CCW AOP.

Associated objective(s):

(RO-C-EOP01-E25) Describe the Rules of Usage associated with the use of AOPs and ARPs during the implementation of the EOPs.

DC COOK OPS Page: 146 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 74 ID: CM-0655 Points: 1.00 Which ONE of the following conditions would require EOP implementation using ADVERSE containment values?

A. Containment radiation level is 200,000 R/hr and stable B. Containment pressure is 1.1 psig and rising C. Containment pressure is 4.5 psig and lowering with Containment Spray in service D. Containment Sump water level is 95% and rising Answer: A Answer Explanation:

Adverse containment values are required to be used when containment pressure is >5 psig or >105 R/Hr. When pressure lowers to <5 psig normal values may be used as long as the integrated dose is <106 R.

A. Correct - 200,000 R/Hr is greater that the trigger of 1 x 105.

B. Incorrect - This is greater than SI setpoint but not above 5 psig used for adverse containment.

C. Incorrect - This would be higher than 2.8 psig but not above 5 psig used for adverse containment.

D. Incorrect - This would indicate a very large break but does not impact the Adverse Containment determination.

Question ID (Status) CM-0655(Active)

External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: F/3 Comments:

Reference:

OHI-4023 Abnormal/Emergency Procedure User's Guide Source: Bank KA - 194001 2.4.17 Generic Emergency Procedures/Plan DC COOK OPS Page: 147 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Knowledge of EOP terms and definitions.

RO - 3.9 SRO - 4.3 CFR - 41.10 / 45.13 K/A Justification - Candidate is required to demonstrate knowledge of the EOP defined Adverse Containment.

Associated objective(s):

(RO-C-EOP01-E8) State the conditions that define "Adverse Containment".

DC COOK OPS Page: 148 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 75 ID: RO32NRC-075 Points: 1.00 A fire has been confirmed within the Turbine Building.

Which set of the following contain the Operations responsibilities in accordance with PMP-2270-FRP-001, Fire Response Plan Attachment 1 Operations Fire Response Responsibilities?

1. Sound Fire Siren.

2. Notify Security to perform "Fire Page Out".

3. Call 911 to request Offsite assistance.

4. Notify WCC\SRO to assume duties of Fire Brigade Leader.

5. Notify Fire Protection Shift Supervisor to assemble Fire Brigade.

6. Announce location on PA.

A. 2,4,5 B. 1,2,6 C. 1,3,5 D. 3,4,6 Answer: B Answer Explanation:

PMP-2270-FRP-001 Attachment 1 directs the operators to Sound Fire Siren (1),

Announce location on PA (6), and Notify Security to perform "Fire Page Out"(2). If the Fire Brigade leader requests, (3) the operators will notify security to request offsite assistance. (5) The Fire brigade will be notified through the page out and the (4)

WCC\SRO reports to act as the advisor to the Fire Brigade Leader.

A. Incorrect - (4) WCC\SRO is not leader and doesn't (5) notify Fire Protection.

B. Correct - see above explanation C. Incorrect - (3) doesnt call 911 for offsite and doesn't (5) notify Fire Protection.

D. Incorrect - (3) doesnt call 911 for offsite and (4) WCC\SRO is not leader Question ID (Status) RO32NRC-075(Active)

External Topic ID:

Level/Difficulty: F/3 DC COOK OPS Page: 149 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Comments:

Reference:

PMP-2270-FRP-001, Fire Response Plan, Attachment 1 Operations Fire Response Responsibilities Source: New KA - 194001 2.4.27 Generic Emergency Procedures/Plan Knowledge of "fire in the plant" procedure.

RO - 3.4 SRO - 3.9 CFR - 41.10 / 43.5 / 45.13 K/A Justification - Question requires candidate knowledge of the Fire Protection Response Procedure.

Associated objective(s):

(RO-C-0660620801-E3) Given a set of plant conditions including Fires in the Plant, explain the procedural mitigation strategy for Fires in the Plant, in accordance with applicable Annunciator Response Procedures and 12-OHP-4025-001-002, Fire Response Guidelines.

DC COOK OPS Page: 150 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 76 ID: RO26-0016 Points: 1.00 SRO ONLY Given the following plant conditions:

A LOCA has occurred.

The crew has just entered OHP-4023-E-1, Loss of Reactor or Secondary Coolant.

The following parameters exist:

All SG pressures 730 psig and slowly trending down

All SG levels controlled at 40% NR

PRZ level off-scale high

Containment Pressure 2.5 psig

RWST level 74% and decreasing slowly

RCS pressure 875 psig and decreasing slowly

Highest CET 500°F Based on these indications, which ONE of the following procedures will the Unit Supervisor direct the crew to enter next?

A. OHP-4023-ES-1.1, SI Termination B. OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization C. OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation D. OHP-4023-E-2, Faulted Steam Generator Isolation Answer: B Answer Explanation:

A. Incorrect - SI termination criteria not met as RCS Pressure is lowering.

B. Correct - RCS Pressure > 300 psig and not stable, requires RCS cooldown via 4023-ES-1.2.

C. Incorrect - RCS pressure and RWST level are high. Entry to OHP-4023-ES-1.3 on low RWST level.

D. Incorrect - SG pressures are trending down because RCS temperature is trending down and AFW flows maintaining levels.

DC COOK OPS Page: 151 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) RO26-0016(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

OHP-4023-E-1, Loss of Reactor or Secondary Coolant Source: Bank Previous NRC exam: 2016 KA - 000009 EA2.38 Small Break LOCA Ability to determine and interpret the following as they apply to a small break LOCA:

Existence of head bubble RO - 3.9 SRO - 4.3 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Justification - Requires the ability to assess a small break LOCA (PRZ Steam Break) with a high pressurizer level and low vessel inventory due to the head bubble and determine the EOP procedure transition to the correct mitigating procedure (ES-1.2).

SRO Justification - Requires assessment of conditions and selection of correct procedure.

RO-C-EOP09-E40-17 is a modified version with C as correct answer used on NRC Exam 2016 Associated objective(s):

(RO-C-EOP09-E40) For each of the E-1 Series procedures, identify the Procedure Transitions.

DC COOK OPS Page: 152 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 77 ID: CM-2738 Points: 1.00 SRO ONLY A Loss of an RHR pump in mode 5 could impact two Technical Specifications:

RCS Loops - Mode 5, Loops Filled

RCS Loops - Mode 5, Loops Not Filled Which ONE of the following describes the reason why these two Limiting Conditions for Operation (LCO) are different?

A. With the reactor coolant loops filled, reactor coolant loops can be used to meet decay heat removal requirements.

B. With the reactor coolant loops NOT filled, one RHR loop is required for RCS cooling with the other required for Refueling Cavity makeup.

C. With the reactor coolant loops filled, the SI Accumulators can be used to meet decay heat removal requirements.

D. With the reactor coolant loops NOT filled, both RHR loops are required to provide sufficient volumetric flow rate.

Answer: A Answer Explanation:

A. Correct - The RCS loops may act as a heat removal method if the loops are filled and the SGs have adequate level.

B. Incorrect - Either Loop of RHR is sufficient for cooling and mixing.

C. Incorrect - Accumulator injection is used in some AOP/EOP situations where other cooling methods are lost, but this is not the reason for the loops filled requirements.

D. Incorrect - Makeup can be performed using the operating loop.

Question ID (Status) CM-2738(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

Tech Specs 3.4.7 & 3.4.8; 1(2)-OHP-4021-002-005 RCS Draining DC COOK OPS Page: 153 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Source: Bank KA - 000025 2.1.28 Loss of Residual Heat Removal System (RHRS)

Conduct of Operations Knowledge of the purpose and function of major system components and controls.

RO - 4.1 SRO - 4.1 CFR - 41.7 KA Justification - Question requires knowledge of the system purpose and requirements during shutdown conditions.

SRO Justification - Question requires knowledge to Technical Specification bases for the different specifications.

Associated objective(s):

(RO-C-01700-E13) Explain the basis for the following RHR System Technical Specification LCOs, Action Statements, and Surveillance Requirements:

a. 3.4.6 RCS Loops - MODE 4

b. 3.4.7 RCS Loops - Mode 5, Loops Filled

c. 3.4.8 RCS Loops - Mode 5, Loops Not Filled

d. 3.4.12 Low Temperature Overpressure Protection

e. 3.5.2 ECCS - Operating

f. 3.5.3 ECCS - Shutdown

g. 3.9.4 RHR and Coolant Circulation - High Water Level Circulation

h. 3.9.5 RHR and Coolant Circulation - Low Water Level DC COOK OPS Page: 154 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 78 ID: RO32NRC-078 Points: 1.00 SRO ONLY Given the following:

A tube rupture has occurred in the #23 Steam Generator.

The crew entered 2-OHP-4023-E-3, Steam Generator Tube Rupture

#23 Steam Generator level is 22% and slowly rising.

A Manual Reactor Trip and Safety Injection were performed.

RCS Pressure is currently 2035 psig.

RCS Tavg is 549°F.

Attempts have been made to close the Steam Generator Stop Valves, but NONE would close.

Which ONE of the following is the action that should be taken based on the failure to close the Steam Generator Stop valves?

A. Transition to 2-OHP-4023-ECA-2-1, Uncontrolled Depressurization Of All Steam Generators to address the failed Steam Generator Stop valves.

B. Transition to 2-OHP-4023-ECA-3.1, SGTR With Loss Of Reactor Coolant -

Subcooled Recovery Desired to begin the RCS cooldown and depressurization.

C. Transition to 2-OHP-4023-ECA-3.2, SGTR With Loss Of Reactor Coolant -

Saturated Recovery Desired to begin the RCS depressurization and ECCS flow reduction.

D. Remain in 2-OHP-4023-E-3, Steam Generator Tube Rupture to perform a cooldown using only the steam dumps.

B Answer:

Answer Explanation:

A. Incorrect - This is the normal procedure used if the Stop valves wouldn't close and a secondary break exists, but with a tube rupture the crew is directed back to E-3.

B. Correct - If the Ruptured SG cannot be isolated a controlled cooldown is performed within ECA-3.1.

C. Incorrect - ECA-3.2 is only entered from ECA-3.1 if RWST water is lost or Steam DC COOK OPS Page: 155 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Generator overfill is imminent (>91%).

D. Incorrect - While a cooldown will be performed using the Steam dumps, E-3 attempts to cool the RCS at a maximum rate to a target temperature which is below ruptured Steam Generator saturation pressure. This is not possible so ECA-3.1 must be used.

Question ID (Status) RO32NRC-078(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

12-OHP-4023-E-3 Background, 2-OHP-4023-E-3 SGTR Source: New KA - 000038 EA2.12 Steam Generator Tube Rupture (SGTR)

Ability to determine and interpret the following as they apply to a SGTR:

Status of MSIV activating system RO - 3.9 SRO - 4.2 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification: Question requires candidate to determine correct action based on the failure of the SG Stop valves to close.

SRO Justification: Question requires assessment of conditions and determination of correct procedural transition from RNO.

Associated objective(s):

(RO-C-AOP0170412-E3) Given a set of plant conditions and the occurrence of an abnormal event, without use of references, explain the procedural mitigation strategy for a Steam Generator Tube Leak in accordance with plant procedures, and standards and expectations for performance.

DC COOK OPS Page: 156 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 79 ID: RO-C-03200-E20-13 Points: 1.00 SRO ONLY OPEN REFERENCE Unit 1 is 100% power on June 26.

Given the following timeline:

1000 1W RHR pump is removed from service due to a mechanical seal leak.

1100 Supplemental DG #1 seizes during a surveillance test.

1200 U1 CD D/G is removed from service due to an oil leak.

In accordance with Technical Specifications, the Unit is required to be placed in Mode 3 no later than ____________________.

A. June 26 at 1900 B. June 26 at 2300 C. June 29 at 1700 D. June 29 at 1800 Answer: B Answer Explanation:

A. Incorrect - Plausible if the applicant applies actions of LCO 3.0.3 immediately upon inoperability of CD D/G.

B. Correct- With 1W RHR out of service, the CD D/G issue creates inoperabilitys on both trains due to safety related equipment supported by CD D/G. This constitutes a condition not addressed by a specific LCO, therefore LCO 3.0.3 should apply. However, LCO 3.8.1 (Condition B) allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to declare supported features inoperable when redundant features are inoperable. Therefore, CD D/G supported equipment must be declared inoperable by 1600 on 6/26/15 and LCO 3.0.3 is entered at that time, requiring shutdown to Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

C. Incorrect - Plausible because this contains the correct time for LCO 3.5.2 only (1W RHR pump).

DC COOK OPS Page: 157 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO D. Incorrect - Plausible because this contains the correct time for LCO 3.8.1 only (CD D/G).

C - ACTION B.1 not met Requires SDG or EDG restored in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Question ID (Status) RO-C-03200-E20-13(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

(Provided)

Tech. Spec. 3.5.2, ECCS - Operating Tech. Spec. 3.8.1, AC Sources - Operating Source: Bank KA - 000056 2.2.38 Loss of Offsite Power Equipment Control Knowledge of conditions and limitations in the facility license.

RO - 3.6 SRO - 4.5 CFR - 41.7 / 43.1 / 45.13 KA Justification: The KA is matched because the applicant is presented with conditions involving equipment out of service for maintenance, and two hours later, a degraded power source is removed from service for maintenance. This equipment is required to address the loss of offsite power. The SRO applicant is then required to analyze the effect of the combination of these pieces of equipment on the LCOs in effect, including LCO 3.0.3.

SRO Justification: The SRO applicant is then required to analyze the effect of the combination of these pieces of equipment on the LCOs in effect, including LCO.

Associated objective(s):

(RO-C-TS01-E11) For a specific plant system or component, apply Technical Specifications to determine:

If the system or component is Tech. Spec. required

Limiting Conditions for Operation

Applicable Modes

Actions required

Surveillance Requirements DC COOK OPS Page: 158 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 80 ID: RO32NRC-080 Points: 1.00 SRO ONLY OPEN REFERENCE 1000 - Unit ONE is performing calibration for the Degraded Voltage relays for the T11D bus.

1200 - MTI has reported the discovery of a manufacturing defect that makes all three relays unreliable. They estimate it will take 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to get replacements and install them.

1300 - Transmission calls the Control Room to report a grid disturbance that results in the Inoperability of the off-site circuit feed to Transformer #9.

Reserve Feed is in the Preferred Lineup.

What Technical Specification LCOs are required to be entered?

A. LCO 3.3.5 Condition A ONLY / LCO 3.8.1 Condition A ONLY B. LCO 3.3.5 Condition A & B ONLY / LCO 3.8.1 Condition A & B ONLY C. LCO 3.3.5 Condition C ONLY / LCO 3.8.1 Condition E ONLY D. LCO 3.3.5 Condition A, B, & C / LCO 3.8.1 Condition A, B, and E Answer: D Answer Explanation:

LCO 3.3.5 Condition A is entered based on MTI reporting all three Degraded Voltage relays having issues.

LCO 3.3.5 Condition B is entered based on MTI reporting all three Degraded Voltage relays having issues.

LCO 3.3.5 Condition C is entered based on the time required Action Completion time not being met.

LCO 3.8.1 Condition A is entered based on the Inoperability of the Off-site feed to TR-9.

LCO 3.8.1 Condition B is entered based on the Degraded Voltage relay issue requiring LCO 3.3.5 Condition C to be entered which makes the CD EDG Inoperable.

LCO 3.8.1 Condition E is entered as both one required offsite circuit and one required EDG is Inoperable.

A. Incorrect - See above explanation.

B. Incorrect - See above explanation.

DC COOK OPS Page: 159 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO C. Incorrect - See above explanation.

D. Correct - See above explanation.

Question ID (Status) RO32NRC-080(Active)

External Topic ID:

Level/Difficulty: H 3 Comments:

Reference:

TS 3.3.5, TS 3.8.1 Reference Provided TS 3.3.5, TS 3.8.1 Source: New KA 000077 Generator Voltage and Electric Grid Disturbance AA2.09 Ability to determine the following as they apply to Generator Voltage and Electric Grid Disturbance: Operational status of Emergency Diesel Generators RO 3.9 SRO 4.3 (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)

KA Justification - Question matches KA as it presents the candidate with conditions that must be analyzed dealing with grid disturbance and EDG Operability.

SRO Justification:

Facility operating limitations in the TS and their bases. Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1). Candidate must identify Tech Spec required actions based on the degraded Voltage relay Inoperability and apply that LCO in addition to the AC Sources Operating (3.8.1) LCO actions.

Associated objective(s):

(RO-C-0821040401-E3) Given a set of plant conditions including the occurrence of Abnormal Electrical Grid Voltage and Frequency explain the procedural mitigation strategy for the malfunction in accordance with 12-OHP-4022-082-004, Degraded Offsite Ac Voltage Response.

DC COOK OPS Page: 160 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 81 ID: NRCAUDIT07-0130A-SRO Points: 1.00 SRO ONLY OPEN REFERENCE The control room operators are responding to a RED path on the Heat Sink Critical Safety Function (CSF) caused by a loss of all Auxiliary Feed Water (AFW) on Unit 2.

The following plant conditions exist:

All Reactor Coolant Pumps (RCPs) have been tripped

East Centrifugal Charging Pump (CCP) is tripped

West CCP is operating

Containment Pressure is 5.3 psig

Control Air to Containment remains isolated due to a large air leak

Wide Range (WR) Steam Generator (SG) levels are lowering at 0.5%/minute SG 1 SG 2 SG 3 SG 4

SG (WR) [%] 19 19 20 25 The AEO reports that the Turbine Driven Auxiliary Feed Pump will be restored within 10 minutes.

Which ONE of the following describes the correct operator response?

A. Immediately initiate bleed and feed. Continue bleed and feed until AFW flow has restored 1 SG Narrow Range (NR) level to greater than 28%.

B. Immediately initiate bleed and feed. Terminate bleed and feed as soon as AFW flow of 240,000 PPH is restored.

C. Continue efforts to restore AFW flow. Do NOT initiate bleed and feed until 2 SG WR levels are less than 16%.

D. Continue efforts to restore AFW flow. Do NOT initiate bleed and feed until 3 SG WR levels are less than 13%.

Answer: A Answer Explanation:

A. Correct - Bleed and feed must be initiated immediately upon reaching the criteria.

The effectiveness depends on the timeliness of initiation. This requires bleed and feed to be initiated when 3 SG WR levels are <21% (adverse)

(Step 3). Bleed and feed is continued until at least 1 SG NR level is >28%

DC COOK OPS Page: 161 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO (step 33).

B. Incorrect - RCS bleed and feed may be stopped when at least 1 SG NR level is

>28% (adverse). Plausible if the procedure exit criteria for establishing AFW flow of 240,000 PPH when recovery occurs before initiation of bleed and feed are applied to termination of bleed and feed.

C. Incorrect - Bleed and feed must be initiated immediately upon reaching the criteria.

The effectiveness depends on the timeliness of initiation. Plausible if candidate assumes Normal Containment and only 2 SGs.

D. Incorrect - Bleed and feed must be initiated immediately upon reaching the criteria.

The effectiveness depends on the timeliness of initiation. Plausible if candidate recognizes value from normal adequate heat sink.

Question ID (Status) NRCAUDIT07-0130A-SRO (Active) External Topic ID: PROC-3:

PROC-3, Procedure Knowledge Level/Difficulty: H/4 Comments:

Reference:

2-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink Step 3 Source: Bank Reference Provided: 2-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink FOP KA - 00WE05 2.4.47 Loss of Secondary Heat Sink Emergency Procedures/Plan Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

RO - 4.2 SRO - 4.2 CFR - 41.10 / 43.5 / 45.12 KA Justification - This meets the K/A because the candidate must recognize the trend of lowering SG level with only 2 PORVs and take the timely action to initiate feed and bleed during the loss of heat sink.

SRO Justification: SRO must assess plant conditions and determine appropriate steps.

Original Question # - RO28 AUDIT Changed distractors B, C and D (changed to Cook FR-H.1 Criteria). Also changed initial SG WR levels in stem. TCW Associated objective(s):

(RO-C-EOP11-E7) For a specified set of conditions, determine if initiation of RCS bleed and feed is necessary.

DC COOK OPS Page: 162 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 82 ID: RO32NRC-082 Points: 1.00 SRO ONLY OPEN REFERENCE Given the following conditions:

Unit 2 reactor power is 85% and stable.

Main Generator load is 1045 MW and stable.

C Main Condenser Vacuum is 27.4 inches Hg and stable.

The following transient sequence occurs:

1100: A Condenser air leak develops in the C Main Condenser Vacuum to lower at a constant rate of 0.2 inches Hg/minute.

1105: A ramp down at 12.3 MW/minute (1%/Min) is commenced in an attempt to stabilize condenser pressure.

Field activities are ongoing to find and isolate the condenser air leak.

(Assume the Load DROP rate remains constant and the rate of Main Condenser pressure RISE remains constant throughout the event)

1. With the above conditions, which of the following is the closest approximate time that the unit will first reach the unacceptable region (curve 1) of Figure TDB FIG-11-19?

2. What action would be directed by the SRO?

A. 1. 1113

2. Direct a Reactor Trip prior to reaching the unacceptable region.

B. 1. 1113

2. Begin a rapid power reduction when you reach the unacceptable region.

C. 1. 1120

2. Direct a Reactor Trip prior to reaching the unacceptable region.

D. 1. 1120

2. Begin a rapid power reduction when you reach the unacceptable region.

Answer: C Answer Explanation:

DC COOK OPS Page: 163 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Using Figure TDB-2-FIG-11-19:

Plot initial conditions of 85% 1045 MW and 27.4" Hg Pressure rises to 1" in first 5 minutes.

Drop 12 MW (1%) and raise .0.2" Hg every subsequent minute and the plot will intercept the Ultimate Trip Curve line at approx. 70 % power, 860.6 MW and 23.4" Hg pressure.

The Lines Change 1.5 over 40% power so Trip Curve is = 24-(((power-60)/40)

1.5)

Timer Line is = 25.5-(((power-60)/40)

1.5)

Vacuum lowering is 26.4 (from 5 minutes into the event) - time

0.2 The Power Ramp begins at 1105 so A is 8 minutes of ramp, B is 12, C is 15, and D is 25 so the power decrease is the same amount.

A. Incorrect - 1113 is approx. time pressure would reach the timed trip line with Vacuum at 24.8 power @ 77% with timed trip at 24.8625. This is not the unacceptable region, but the 30-minute vacuum trip timer starts. This would be a good time to begin a rapid power reduction.

B. Incorrect - 1113 is approx. time pressure would reach the timed trip line with Vacuum at 24.8 power @ 77% with timed trip at 24.8625. This is not the unacceptable region, but the 30-minute vacuum trip timer starts. This would be a good time to begin a rapid power reduction.

C. Correct - see explanation above. A preemptive trip is warranted.

D. Incorrect - 1120 is the correct time which would result in reaching the unacceptable region. This is also the variable trip setpoint so the plant will trip at this time.

Reference Figure TDB-2-FIG-11-19 24 22.5 100 1229.4 25.5 24 60 Time Power MW Vacuum Timer Start Trip 1100 85 1044.99 27.4 24.5625 23.0625 1101 85 1044.99 27.2 24.5625 23.0625 1102 85 1044.99 27 24.5625 23.0625 1103 85 1044.99 26.8 24.5625 23.0625 1104 85 1044.99 26.6 24.5625 23.0625 1105 85 1044.99 26.4 24.5625 23.0625 1106 84 1032.696 26.2 24.6 23.1 1107 83 1020.402 26 24.6375 23.1375 1108 82 1008.108 25.8 24.675 23.175 DC COOK OPS Page: 164 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 1109 81 995.814 25.6 24.7125 23.2125 1110 80 983.52 25.4 24.75 23.25 1111 79 971.226 25.2 24.7875 23.2875 1112 78 958.932 25 24.825 23.325 1113 77 946.638 24.8 24.8625 23.3625 1114 76 934.344 24.6 24.9 23.4 1115 75 922.05 24.4 24.9375 23.4375 1116 74 909.756 24.2 24.975 23.475 1117 73 897.462 24 25.0125 23.5125 1118 72 885.168 23.8 25.05 23.55 1119 71 872.874 23.6 25.0875 23.5875 1120 70 860.58 23.4 25.125 23.625 1121 69 848.286 23.2 25.1625 23.6625 1122 68 835.992 23 25.2 23.7 1123 67 823.698 22.8 25.2375 23.7375 1124 66 811.404 22.6 25.275 23.775 1125 65 799.11 22.4 25.3125 23.8125 1126 64 786.816 22.2 25.35 23.85 1127 63 774.522 22 25.3875 23.8875 1128 62 762.228 21.8 25.425 23.925 1129 61 749.934 21.6 25.4625 23.9625 1130 60 737.64 21.4 25.5 24 1131 59 725.346 21.2 25.5 24.0375 1132 58 713.052 21 25.5 24.075 1133 57 700.758 20.8 25.5 24.1125 1134 56 688.464 20.6 25.5 24.15 1135 55 676.17 20.4 25.5 24.1875 Question ID (Status) RO32NRC-082(Active)

External Topic ID:

Level/Difficulty: H/4 Comments:

Reference:

OHP-4024 DCS-MT DROP 134-136 Cond Low Vac, TDB 2-FIG-11-19 Source: New KA - 000051 A2.02 Loss of Condenser Vacuum Ability to determine and interpret the following as they apply to Loss of Condenser vacuum: Conditions requiring a Reactor and/or Turbine trip.

RO - 3.9 SRO - 4.1 CFR - 41.7 / 41.10 / 43.5 / 45.1 DC COOK OPS Page: 165 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO KA Justification - The question meets the K/A, requires examinee ability to determine conditions requiring unit trip as applied to loss of condenser vacuum.

SRO Justification: SRO must determine action to take based on assessment of plant conditions.

Associated objective(s):

(RO-C-0530190812-E5) Given a set of plant conditions including a Steam Seal Controller Failure/Loss of Main Condenser describe the required operator actions to Correct, control, or mitigate the plant response in accordance with the applicable Annunciator Response Procedures.

DC COOK OPS Page: 166 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 83 ID: NRCAUDIT07-0388 Points: 1.00 SRO ONLY Unit 1 experienced a large break LOCA.

The following conditions exist:

East RHR Pump is tagged out for maintenance.

Containment pressure is stable at 9 psig.

RWST level has lowered to less than 25%.

East CTS Pump running on the Recirc. Sump.

The crew has performed OHP-4023-FR-Z.1, Response to High Containment Pressure.

The crew implemented OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation, after the Recirc Sump to West RHR/CTS Pumps valve (ICM-306) failed to open while attempting to open the valve per OHP-4023-ES-1.3, Transfer to Cold Leg Recirculation.

NOTE: OHP-4023-FR-Z.1 requires both CTS pumps to be in operation, but OHP-4023-ECA-1.1 limits the operators to only one CTS pump.

Which procedural guidance on CTS pumps takes priority under these conditions and what is the basis for this requirement?

A. OHP-4023-FR-Z.1 takes priority because a total loss of RHR causes the CTS system to become relatively more important to reduce containment pressure.

B. OHP-4023-FR-Z.1 takes priority because red path FRPs always have priority over ECA procedures.

C. OHP-4023-ECA-1.1 takes priority because it conserves RWST water level as long as possible for injection while providing sufficient CTS flow to mitigate containment pressure.

D. OHP-4023-ECA-1.1 takes priority because ECA procedures always have priority over FRPs.

Answer: C DC COOK OPS Page: 167 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Answer Explanation:

A. Incorrect - ECA-1 .1 CTS operation takes priority over FR-Z.1 Plausible: Although a loss of RHR and containment sump recirc causes a loss of the containment heat sink, the supply for SI comes from the RWST which will be drawn down until containment sump recirculation can be established.

B. Incorrect - ECA-1 .1 CTS operation takes priority over FR-Z.1 Plausible: FRPs normally take priority over most EOPs.

C. Correct - ECA-1 .1 operation of the CTS pumps takes priority over FR-Z.1 in order to limit RWST usage per FR-Z.1 step 2.

D. Incorrect - ECAs do not always have priority over FRPs. Plausible: Some ECAs take priority e.g. ECA-0.0 has priority over FRPs in that FRPs are suspended until power is restored.

Question ID (Status) NRCAUDIT07-0388(Active)

External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: F/3 Comments:

REFERENCE:

1-OHP-4023-ECA-1.1 Loss of Emergency Coolant Recirculation, 01-OHP-4023-FR-Z.1, Response to High Containment Pressure Source: Bank Previous NRC exam: 2012 A - 00WE14 EA2.2 High Containment Pressure Ability to determine and interpret the following as they apply to the High Containment Pressure: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments RO - 3.3 SRO - 3.8 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Question requires operator ability to determine the correct procedural priority based on limitations (loss of recirc / RWST level) and precautions (Z.1) and the reason.

SRO Justification: SRO must assess plant conditions and determine appropriate steps based on cautions and procedural hierarchy.

Associated objective(s):

(RO-C-EOP09-E42) For each of the E-1 Series procedures, discuss any EOP Key Decision Points DC COOK OPS Page: 168 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 84 ID: NRCAUDIT07-0757 Points: 1.00 SRO ONLY The following conditions exist:

Large Break LOCA is in progress

Containment pressure is 1.9 psig and stable

You notice an ORANGE condition on CONTAINMENT Critical Safety Function Status Tree due to the "FLOOD LEVEL" lights being Lit.

(1) What action will be directed by OHP-4023-FR-FR Z.2, Response to Containment Flooding?

(2) What is the concern if these actions are not successful?

A. (1) Divert RHR flow from the Containment Sump to the RWST to lower Containment Level.

(2) High water levels could result in critical components needed for plant recovery being damaged and rendered inoperable.

B. (1) Identify and isolate the source of excess water using control board indications and Containment Sump samples.

(2) Water levels could reach the bottom of the reactor vessel resulting in thermal shock and vessel failure.

C. (1) Identify and isolate the source of excess water using control board indications and Containment Sump samples.

(2) High water levels could result in critical components needed for plant recovery being damaged and rendered inoperable.

D. (1) Stop both containment spray pumps.

(2) Water levels could reach the bottom of the reactor vessel resulting in thermal shock and vessel failure.

Answer: C Answer Explanation:

Containment design basis flood level takes into account the entire water contents of the RCS, RWST, Ice condenser ice bed melt, and SI accumulators, plus the added mass of a LOCA and a steam line or feedline break inside containment. NESW and CCW may be major contributors to exceeding "flood" level and causing a loss of equipment required for long term cooling.

A. Incorrect - Water is not pumped out of containment using the RHR pumps.

DC COOK OPS Page: 169 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO B. Incorrect - Water reaching the Reactor vessel would not cause thermal shock or vessel failure.

C. Correct - See above D. Incorrect - The CTS pumps are not stopped due to high level. Water reaching the Reactor vessel would not cause thermal shock or vessel failure.

Question ID (Status) NRCAUDIT07-0757(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

12-OHP-4023-FR-Z-2, Response to Containment Flooding Background Source: Bank Previous NRC exam: 2008 KA - 00WE15 2.4.6 Containment Flooding Emergency Procedures/Plan Knowledge of EOP mitigation strategies.

RO - 3.7 SRO - 4.7 CFR - 41.10 / 43.5 / 45.13 KA Justification: Question requires knowledge of the mitigating strategy for Containment flooding when in Emergency Procedures.

SRO Justification: Question requires SRO assessment of plant conditions and then selection of a section of a procedure to mitigate or recover.

Source: Cook 2008 Exam, Cook 2006 NRC Exam Associated objective(s):

(RO-C-EOP13-E7) For each of the FR-Z series procedures discuss the basis or reason for all Steps.

DC COOK OPS Page: 170 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 85 ID: NRCAUDIT07-0367-SRO Points: 1.00 SRO ONLY Unit 1 has implemented 1-OHP-4023-ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel when a YELLOW path occurs on Reactor Coolant Inventory due to RVLIS Upper Plenum indication < 92%.

Which ONE of the following statements describes the correct decision and reasoning as to whether to use 1-OHP-4023-FR-I-3, Response to Voids in Reactor Vessel, to address this YELLOW path procedure?

A. Implement 1-OHP-4023-FR-I-3 because EOP usage requires transitioning from an ES procedure for a YELLOW path if there are no other higher priority critical safety functions.

B. Implement 1-OHP-4023-FR-I-3 in order to vent the reactor vessel void through the head vent and collapse the void to allow the cooldown to continue.

C. Do NOT implement 1-OHP-4023-FR-I-3 because this procedure requires starting one RCP, which can NOT be done with a void in the reactor vessel due to the potential for gas binding.

D. Do NOT implement 1-OHP-4023-FR-I-3 because ES-0.3 expects a void in the head to form and controls the formation and growth of the void.

Answer: D Answer Explanation:

A. Incorrect - EOP usage allows but does NOT require transitioning to yellow path FR procedures. Plausible if student believes YELLOW path FRs are required.

B. Incorrect - Steps in 1-OHP-4023-FR-I-3 are not performed because of the Caution prior to Step 1. Plausible if student believes that a natural circulation cooldown cannot continue with a RV head void.

C. Incorrect - FR-I.3 is written to start an RCP only if it is suspected the void is not due to non-condensable gasses. Plausible if student does not understand FR-I-3 restrictions for RCP start.

D. Correct - EOP usage allows but does NOT require transitioning to yellow path FR procedures. ES-0.3 expects a void in the head to form and controls the formation and growth of the void.

DC COOK OPS Page: 171 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) NRCAUDIT07-0367-SRO (Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

OHP-4023-ES-0.3 Natural Circulation Cooldown with Steam Void in Vessel Source: Bank KA - 00WE10 EA2.1 Natural Circulation with Steam Void in Vessel with/without RVLIS Ability to determine and interpret the following as they apply to the Natural Circulation with Steam Void in Vessel with/without RVLIS: Facility conditions and selection of appropriate procedures during abnormal and emergency operations RO - 3.2 SRO - 3.9 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification: Question requires knowledge of the steps performed and how those actions may impact entry conditions for other procedures.

SRO Justification: SRO must assess plant conditions and determine if a transition is required.

Associated objective(s):

(RO-C-EOP01-E22) Describe the rules of usage for implementation of the Functional Restoration Procedures.

DC COOK OPS Page: 172 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 86 ID: NRCAUDIT07-1058 Points: 1.00 SRO ONLY Following a failure of the PRZ Level Control System, Volume Control Tank (VCT) level rose to 70% and VCT pressure rose to 55 psig.

This caused Reactor Coolant Pump (RCP) seal injection flow to ____(1)_____ and will require the VCT to be vented to _____(2)_____ .

_____(1)_____ _____(2)_____

A. slightly rise restore operability for the Emergency Boration Path from the BAST B. remain the same restore operability for the Emergency Boration Path from the BAST C. remain the same prevent exceeding the RCS Hydrogen Concentration TRM limits D. slightly rise prevent exceeding the RCS Hydrogen Concentration TRM limits Answer: B Answer Explanation:

Raising VCT level without venting the VCT will cause Pressure to rise. This will provide more NPSH to the CCP. However, Seal Injection flow is controlled by the position of QRV-251 and QRV-200. The CCP discharge pressure may rise but the QRV-251 will throttle to maintain charging flow and QRV-200 will maintain Seal Injection flow constant.

A. Incorrect - # 1 Seal leakoff flow will not change & the VCT must be vented to restore BA Flowpath.

B. Correct - See above C. Incorrect - The RCS Hydrogen concentration is not limited in the TRM. (Design is to maintain enough H2 toscavenge O2.)

D. Incorrect - # 1 Seal leakoff flow will not change & the RCS Hydrogen concentration is not limited in the TRM. (Design is to maintain enough H2 toscavenge O2.)

DC COOK OPS Page: 173 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) NRCAUDIT07-1058(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

1-OHP-4021-005-007 Operation of Emergency Boration Flow Paths P&L

3.1 Source

Modified - 2006 ILE (Q26)

KA - 003000 A2.04 Reactor Coolant Pump System (RCPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations: Effects of fluctuation of VCT pressure on RCP seal injection flow RO - 2.4 SRO - 2.8 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification: Question requires candidate to predict impacts of higher VCT press on seal injection (and on the BA operability) and identify the correct reason to lower pressure.

SRO Justification: Question requires SRO to assess conditions and identify an inoperable system based on information contained with the surveillance and bases.

Source: Cook 2006 NRC Exam #26 Changed to Include Venting VCT in Stem. Changed Distractors to Include BA Flowpath Operability & H2 Original Quest. KA - 003.a2.05 Associated objective(s):

(RO-C-00300-E15) Describe the purpose/function of the following instruments used to verify proper operation of the CVCS:

a. Temperature

b. Pressure

c. Flow

d. Level (RO-C-00201-E19) Given a description of plant conditions, predict the impact on and response of the Reactor Coolant Pumps.

DC COOK OPS Page: 174 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 87 ID: RO32NRC-087 Points: 1.00 SRO ONLY OPEN REFERENCE Unit 1 is operating at 100% power.

During the Unit 1 performance of SR 3.3.1.5, ACTUATION LOGIC TEST, for the RTS Instrumentation automatic trip logic for the A train, entry into the Required Actions of TS 3.3.1 is delayed. The A train is bypassed at 1000 on July 14 to perform the test.

The B train is OPERABLE.

One (1) hour after the A train is bypassed and while still in bypass, it is determined that the train cannot perform its intended function. What TS ACTIONS are required?

Declare the train inoperable and restore the train to OPERABLE status by:

A. 1400 on July 14.

B. 1700 on July 14.

C. 2000 on July 14.

D. 1100 on July 16.

Answer: B Answer Explanation:

Declare the train inoperable and enter the Required Actions and associated Completion Times for the inoperable train. Enter Condition J of TS 3.3.1 and restore the train to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or by 1700. Once the train is known to be inoperable the 4-hour allowed bypass time is no longer applicable and the 6-hour time clock to repair must be entered immediately.

Ref: LCO 3.0.2 TS 3.3.1, Condition J Bases A. Incorrect - This is the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the time the Train may be bypassed.

B. Correct - This is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from the time of discovery.

C. Incorrect - This is the remainder of the 4-hour bypass time and then the 6-hour restoration time (10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months from initial bypass)

DC COOK OPS Page: 175 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO D. Incorrect - This is the Requirements for Condition B which applies in Mode 3, 4,or 5 (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovery)

Question ID (Status) RO32NRC-087(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference PROVIDED TS 3.3.1 pages 3.3.1-1 through 3.3.1-5 and 3.3.1-15 (Page 4 of table only)

Source: New KA - 013000 2.2.40 Engineered Safety Features Actuation System (ESFAS)

Equipment Control Ability to apply Technical Specifications for a system.

RO - 3.4 SRO - 4.7 CFR - 41.10 / 43.2 / 43.5 / 45.3 KA Justification: Question requires candidate to apply Technical Specification.

SRO Justification: Question requires candidate to determine the correct action statement required and to properly apply the time limits based on LCO 3.0.2.

Associated objective(s):

(RO-C-TS01-E11) For a specific plant system or component, apply Technical Specifications to determine:

If the system or component is Tech. Spec. required

Limiting Conditions for Operation

Applicable Modes

Actions required

Surveillance Requirements DC COOK OPS Page: 176 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 88 ID: RO32NRC-088 Points: 1.00 SRO ONLY Given the following conditions:

Unit 2 control room operators are responding to a SGTR with a loss of offsite power.

02-OHP-4023-E-3, Steam Generator Tube Rupture is in progress.

Chemistry has confirmed through sampling that SG 23 is ruptured.

All SG pressures are 1032 psig.

ALL SG PORVs were partially open prior to performing step 3 to isolate flow from SG 23.

The crew has just completed step 3.

The STA reports that Steam Line Radiation monitors for #23 and #24 Steam Generator PORVs are in High Alarm.

Which ONE of the following describes the action required per 02-OHP-4023-E-3 and the reason for this action?

A. Return to step 1 of 02-OHP-4023-E-3 since SG 24 now shows signs of rupture.

B. Continue with Step 4 of 02-OHP-4023-E-3 to cooldown and depressurize, since the radiation alarms are caused by shine C. Transition to 02-OHP-4023-ECA-3.1, SGTR with Loss of Reactor Coolant -

Subcooled Recovery Desired since multiple SGs are ruptured.

D. Transition to 02-OHP-4023-ECA-3.1, SGTR with Loss of Reactor Coolant -

Subcooled Recovery Desired since the ruptured SG NOT isolated from all of the intact SGs.

Answer: A Answer Explanation:

A. Correct - The High alarm on MRA-2602 indicates leakage on SG 24. SG24 is not located near SG 23 so the indications are not from Shine. The E-3 foldout page directs the return to step 1 to isolate subsequent leaking SGs.

B. Incorrect - The High alarm on MRA-2602 indicates leakage on SG 24. SG24 is not located near SG 23 so the indications are not from Shine.

C. Incorrect - ECA-3.1 is used only if the ruptured SG(s) cannot be isolated from at least DC COOK OPS Page: 177 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 1 intact SG. Plausible since the multiple failures may be addressed by an ECA.

D. Incorrect - ECA-3.1 is used only if the ruptured SG(s) cannot be isolated from at least 1 intact SG. Plausible since the multiple failures may be addressed by an ECA.

Question ID (Status) RO32NRC-088(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

2-OHP-4023-E-3 Steam Generator Tube Rupture Foldout page #5 Source: New KA - 039000 A2.03 Main and Reheat Steam System (MRSS)

Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations:

Indications and alarms for main steam and area radiation monitors (during SGTR)

RO - 3.4 SRO - 3.7 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification: Question addresses the impacts of a MS SG PORV steam release during a SGTR and the impacts of the alarming channels and procedural action required.

SRO Justification: Question requires SRO knowledge of the procedural transition required with the SGTR procedure.

Associated objective(s):

(RO-C-EOP08-E17) For the E-3 procedures and the ECA-3 series procedures discuss the basis or reasons for all Foldout page items, in accordance with the applicable Plant specific Background documents.

DC COOK OPS Page: 178 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 89 ID: NRCAUDIT07-0975-SRO Points: 1.00 SRO ONLY You are the Unit Supervisor. Unit 2 is at 100% power.

Panel 215 Drop 48 - BATTERY N UNDERVOLTAGE has just alarmed.

Investigation revealed that a metal plate has shorted the battery terminals.

Which ONE of the following identifies the effects on the operability and capability of the Auxiliary Feedwater System?

A. The TDAFW Pump will NOT start and the FMO-211, 221, 231, & 241 TDAFW to SG Isolation valves are failed in the open position. Declare the TDAFW Train INOPERABLE.

B. The TDAFW Pump will start and the FMO-211, 221, 231, & 241 TDAFW to SG Isolation valves are failed in the open position. Declare the N Train Battery INOPERABLE.

C. The TDAFW Pump will start but the MCM-221 SG Steam supply to TDAFW Pump Isolation valve is failed in the closed position. Declare the TDAFW Pump INOPERABLE.

D. The TDAFW Pump will NOT start and the FMO-211, 221, 231, & 241 TDAFW to SG Isolation valves are failed in the closed position. Declare the TDAFW Train INOPERABLE.

Answer: A Answer Explanation:

A. Correct - The Train N battery supplies power to the TDAFW pump start circuitry, Trip

& Throttle Valve, SG FW Valves and Test Valves. The TDAFW pump valves are normally open and so they will fail in the open position (as-is).

TS 3.7.5 Condition B allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

B. Incorrect - The TDAFW Pump will not start. If Train N battery is inoperable the TDAFW pump must be declared inoperable.

C. Incorrect - The TDAFW pump will not start. Steam supply to TDAFW Pump Isolation valve is normally open.

DC COOK OPS Page: 179 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO D. Incorrect - The valves are normally open.

Question ID (Status) NRCAUDIT07-0975-SRO (Active)

External Topic ID: 5600: 5600, Auxiliary Feedwater Level/Difficulty: H/3 SRO ONLY Comments:

Reference:

TS 3.7.5 & 3.8.4 Source: Bank Previous NRC exam: 2007 KA - 061000 2.2.37 Auxiliary / Emergency Feedwater (AFW) System Equipment Control Ability to determine operability and/or availability of safety related equipment.

RO - 3.6 SRO - 4.6 CFR - 41.7 / 43.5 / 45.12 KA Justification - Requires system knowledge to predict the impact and use of TS to determine operability actions for AFW battery failures.

SRO Justification: Questions requires a determination of the Correct Technical Specification call to make based on the support system inoperability.

Original Question # - RO28 AUDIT, NRCAUDIT07-0975, Cook 2006 NRC Exam -

092-5 Original Question KA - 061000 A2.03 Associated objective(s):

(RO-C-05600-E13) Given a description of plant conditions, determine applicable TRM, TS, System Operability, and most limiting LCO.

DC COOK OPS Page: 180 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 90 ID: RO32NRC-090 Points: 1.00 SRO ONLY During the performance of Attachment 2, Fast Speed Start the EDG, of 1-OHP-4030-132-027-AB, AB Diesel Generator Operability Test, load is reduced to 900KW prior to aligning the EDG to Bus T11B and then it is reduced to 500 KW prior to opening the T11B feed breaker.

1. Which one of the following describes the reason for this unloading sequence?

2. Is the EDG considered OPERABLE or INOPERABLE while performing the test?

A. 1. The load is reduced prior to closing the T11B breaker to prevent overloading the T11B feed breaker. The Load must be less than 1000 KW before opening T11B breaker to prevent EDG overspeed.

2. The EDG is INOPERABLE.

B. 1. The load is reduced prior to closing and opening the T11B breaker to minimize the transients placed on the EDG and busses.

2. The EDG is INOPERABLE.

C. 1. The load is reduced prior to closing the T11B breaker to prevent overloading the T11B feed breaker. The Load must be less than 1000 KW before opening T11B breaker to prevent EDG overspeed.

2. The EDG is OPERABLE.

D. 1. The load is reduced prior to closing and opening the T11B breaker to minimize the transients placed on the EDG and busses.

2. The EDG is OPERABLE.

Answer: B Answer Explanation:

A. Incorrect - The load is reduced to prevent transients. Plausible since T11B is the valve bus and does not carry the same loads as the T11A Bus. The EDG speed would rise when opened under load but the TS requires that the EDG does not overspeed from full load. The test procedure declares the EDG INOPERABLE.

B. Correct - The load is reduced to prevent transients. The test procedure declares the EDG INOPERABLE.

C. Incorrect - The load is reduced to prevent transients. Plausible since T11B is the valve bus and does not carry the same loads as the T11A Bus. The EDG DC COOK OPS Page: 181 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO speed would rise when opened under load but the TS requires that the EDG does not overspeed from full load. Many Surveillances allow the conduct without declaring the equipment INOPERABLE.

D. Incorrect - The load is reduced to prevent transients. Many Surveillances allow the conduct without declaring the equipment INOPERABLE.

Question ID (Status) RO32NRC-090(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

1-OHP-4030-132-027-AB AB Diesel Generator Operability Test Attachment #2 DG1AB Fast Speed Start step 4.1 RO-C-IF05, Motors and Generators Source: New KA - 064000 A2.11 Emergency Diesel Generator (ED/G) System Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G System and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations:

Conditions (minimum load) required for unloading an ED/G RO - 2.6 SRO - 2.9 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification: Question requires candidate to determine the reason for minimum EDG loading and operability.

SRO Justification: Question requires knowledge of procedural steps that declare the equipment Inoperable beyond the normal operability status.

Associated objective(s):

(RO-C-03200-E17) Explain the basis for the following Technical Specification LCOs, Action Statements, and Surveillance Requirements:

a. 3.8.1 A.C. Sources Operating

b. 3.8.2 A.C. Sources Shutdown

c. 3.8.3 Diesel Fuel Oil DC COOK OPS Page: 182 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 91 ID: RO32NRC-091 Points: 1.00 SRO ONLY Given the following conditions on Unit 1:

The plant is operating at 6% power preparing for Turbine roll.

NLP-151, PZR Level Channel 1 failed high 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ago.

The bi-stables have been tripped and all actions are complete as per 1-OHP-4022-013-010, Pressurizer Level Instrument Malfunction.

PZR level is currently 25% on the remaining PZR Level channels.

Which ONE of the following describes the effects on the plant if NLP-153, PZR Level Channel 3 fails low and the effect on Unit Supervisor's decision to trip bi-stables for the Channel 3 failure?

Note: Assume NO operator actions.

A. Letdown will Isolate and heaters will de-energize.

Bi-stables may be tripped without causing a reactor trip. Power must remain less than 10%.

B. Letdown will Isolate and heaters will de-energize.

Bi-stables should NOT be tripped since a reactor trip will be generated. Power must be reduced to less than 5%.

C. Letdown will remain in service and heaters will de-energize.

Bi-stables should NOT be tripped since a reactor trip will be generated. Power must remain less than 10%.

D. Letdown will remain in service and heaters will de-energize.

Bi-stables may be tripped without causing a reactor trip. Power must be reduced to less than 5%.

Answer: A Answer Explanation:

A. Correct - With Channel 1 NLP-151 in the tripped condition, the High level Rx Trip signal will be made up for 1 channel (1/2 coincidence on remaining channels). The level control selector switch for the pressurizer is in the 2/3 position with channel 3 NLP-153 as the controlling channel. When it fails DC COOK OPS Page: 183 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO low letdown will isolate and the heaters will de-energize. When the bi-stables are tripped, a reactor trip signal will be generated but it is blocked by P-7 (Reactor and Turbine power both below 10%). Plant startup can NOT continue. Power must be maintained below 10% (P-7).

B. Incorrect - Letdown will isolate. Reactor will not trip (See Answer A). Plausible as student could be focused on Technical Specification impacts and not consider operational impacts causing Letdown isolation.

C. Incorrect - Letdown will isolate. Power does not need to be reduced. Power just must remain less than P-7 (10%). Plausible as student could confuse Mode change requirements with Tech Spec compliance.

D. Incorrect - Reactor will not trip (See Answer A). Power does not need to be reduced.

Power just must remain less than P-7 (10%). Plausible as student could believe a trip signal could be generated with the additional bistable actuation.

Question ID (Status) RO32NRC-091(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

SOD-00202-003, Pressurizer Level Control TS 3.3.1 actions D & N. Table 3.3.1-1 Function 9 Source: Bank KA - 011000 A2.10 Pressurizer Level Control System (PZR LCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations: Failure of PZR level instrument - high RO - 3.4 SRO - 3.6 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Requires the ability to predict the impacts of multiple PZR Level channel failures (including a high failure), and to use the AOPs and TS to determine the appropriate actions.

SRO Justification - SRO must determine whether to trip bi-stables based on failures and power levels and address the TS restriction for power.

Original Question # - NRC02-045-3 (SRO41), RO24 Audit-096-8, NRC Exam 2010-093 Associated objective(s):

(RO-C-00202-E17) Explain the basis for the Limiting Condition of Operation (LCO), Applicability, Required Actions and Surveillance Requirement for the following Technical Specification and Technical Requirements Manual Items:

a. 3.3.1, Reactor Trip System (RTS) Instrumentation

1) Function 8a, Pressurizer Pressure low

2) Function 8b, Pressurizer Pressure high

3) Function 9, Pressurizer Water Level high DC COOK OPS Page: 184 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 92 ID: RO32NRC-092 Points: 1.00 SRO ONLY OPEN REFERENCE Given the following conditions:

Unit 2 is at 100% power.

Radiation Monitoring Panel Alert alarm was received on SRA 2900 at about midnight (0000).

A RCS to Steam Generator Tube Leak is suspected.

The crew is performing 12-OHP-4024-139, Annunciator Response: Radiation for 2-SRA-2900

The activities from SRA-2905 have been recorded at 15-minute intervals on Data Sheet #1 starting at the 15-minute mark and plotted on 2-Figure 19.19a.

The SJAE flowrate is 17 scfm.

SEE GRAPH PROVIDED ON NEXT PAGE Which ONE of the following describes the appropriate actions in accordance with procedure?

A. Trip the reactor and initiate safety injection.

B. Reduce power to less than 50% in one hour and be in MODE 3 in the next two hours.

C. Reduce power to be in MODE 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D. Reduce power to be in MODE 3 in the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

DC COOK OPS Page: 185 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Answer: B Answer Explanation:

The slope of the line shows a rate of change of > 30gpd/hr over the 90-105 minute time frame. Leakage also exceeds > 75 gpd about the same time so step 3.1.6 applies.

A. Incorrect - A RX trip and SI are not warranted given this leakage. Plausible since the procedure cautions that a rapid change between 75 gpd and 150 gpd are a SGTR precursor.

DC COOK OPS Page: 186 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO B. Correct - the plant must be below 50% in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and modes 3 in the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. Incorrect - This is the TS required action but is less restrictive than the procedural requirement. (Also 3.1.8)

D. Incorrect - This is required if the slope was less than 30 gpd and the total was < 150 gpd.

Question ID (Status) RO32NRC-092(Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

Reference:

12-OHP-4024-139 SRA-2900 pages 121-128 Source: New KA - 055000 2.1.23 Condenser Air Removal System (CARS)

Conduct of Operations Ability to perform specific system and integrated plant procedures during all modes of plant operation.

RO - 4.3 SRO - 4.4 CFR - 41.10 / 43.5 / 45.2 / 45.6 KA Justification: Question requires candidate to determine plant shutdown requirements based on the SJAE radiation monitor readings.

SRO Justification: SRO must interpret the readings and trend and determine the procedural shutdown requirements.

Associated objective(s):

(RO-C-AOP0170412-E2) Given a set of plant conditions and the occurrence of an abnormal event, without use of references, explain the required operator actions to stabilize plant conditions after a Steam Generator Tube Leak prior to formal procedure implementation in accordance with plant procedures, and standards and expectations for performance.

DC COOK OPS Page: 187 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 93 ID: 2008NRC-0605-SRO Points: 1.00 SRO ONLY OPEN REFERENCE Given the following conditions on Unit 1:

The unit is operating at 100% power.

A Containment Pressure Relief is in progress.

VRS-1201 Upper Containment Normal Range Monitor failed due to a power supply failure.

The Containment Pressure Relief is stopped.

Which ONE of the following describes the restrictions placed on Containment Purge/Pressure Relief Operations?

A. Containment Purge/Pressure Relief operations may NOT be performed until VRS-1201 is restored to operable status.

B. Containment Purge/Pressure Relief operations may continue under administrative controls provided that VRS-1201 is restored to operable status prior to entering Mode 4 following the next refueling outage.

C. Containment Purge/Pressure Relief operations may continue under administrative controls for up to 7 days, provided that area surveys of upper containment are performed at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D. Containment Purge/Pressure Relief operations may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months before VRS-1201 is required to be restored to operable status.

Answer: B Answer Explanation:

A. Incorrect - Condition A of TS 3.3.6 allows Purge Operations if 2/3 channels per train are operable.

B. Correct - Condition A of TS 3.3.6 allows Purge Operations if 2/3 channels per train are operable. The failed channel must be fixed during the next refueling outage.

C. Incorrect. Sampling every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is required per TRM 8.3.8 if BOTH VRS-1101 &

1201 are inoperable. Other Channels within TRM 8.3.8 have a 7-day limit.

D. Incorrect - Purge operations are not limited. These requirements are from the Technical Specification 3.3.6 Condition C.

DC COOK OPS Page: 188 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Question ID (Status) 2008NRC-0605-SRO (Active)

External Topic ID: TS01-3: TS01-3, Technical Specifications/TRM - Application1350:

1350, Radiation Monitoring System Level/Difficulty: H/3 SRO ONLY Comments:

Reference:

Tech Specs 3.3.6, TRM 8.3.8 Reference Provided - TS 3.3.6, TRM 8.3.8 Source: Bank Previous NRC exam: 2018 KA - 072000 A2.02 Area Radiation Monitoring (ARM) System Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system and (b) based on those predictions, use procedures to Correct, control, or mitigate the consequences of those malfunctions or operations:

Detector failure RO - 2.8 SRO - 2.9 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification: Candidate is required to demonstrate ability to identify the impacts of an ARM failure and apply Tech Spec restrictions to control operation of the Containment purge System.

SRO Justification: Question has candidate assess the plant conditions and determine the correct actions based detailed procedural steps (RNO). Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Original Question # - RO28 AUDIT, 2008NRC-0605, RO27 Audit -91, NRC EXAM 2008-92, NRC Exam 2018 Original Question KA - 072000 A2.01 Associated objective(s):

(RO-C-01350-E9) Given a description of plant conditions, determine all applicable LCOs, System Operability, and most limiting LCO.

DC COOK OPS Page: 189 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 94 ID: RO32NRC-094 Points: 1.00 SRO ONLY Given the following plant conditions:

Unit One is operating at 100% power A loss of Offsite Power and failure of both Diesel Generators has occurred The operating crew has entered OHP-4023-ECA-0.0 Operators are performing a 100°F/hr cooldown in OHP-4023-ECA-0.0 The dispatcher has informed the Control Room that Emergency Power (EP) is available for station use NOTE:

OHP-4023-E-0, Reactor Trip Or Safety Injection OHP-4023-ECA-0.0, Loss Of All AC Power OHP-4023-SUP-009, Restoration of 4KV Power from EP What actions should the Unit Supervisor perform?

A. Direct the RO to stop the cooldown and transition to OHP-4023-SUP-009 to restore Offsite Power.

B. Direct the RO to complete the cooldown and continue in OHP-4023-ECA-0.0.

C. Direct the RO to continue the cooldown and in parallel have the BOP perform OHP-4023-SUP-009.

D. Direct the RO to complete the cooldown and then restore Offsite Power per OHP-4023-SUP-009.

Answer: C Answer Explanation:

A. Incorrect - Plausible since the candidate may not recognize that the step to restore AC Power is a continuous action step and continue through ECA-0.0 until the next transition out of that procedure. Also, the operator would have to complete the cooldown step.

B. Incorrect - Plausible since the step to restore AC Power is a continuous action step and the candidate may believe that the step is a transition to OHP-4023-SUP-009 vice while continuing.

C. Correct - The operating crew would continue performing actions in ECA-0.0 and DC COOK OPS Page: 190 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO perform OHP-4023-SUP-009 in parallel to restore AC Power.

D. Incorrect - Plausible since the candidate may not recognize that the step to restore AC Power is a continuous action step and determine that the current step must be completed and then go back and re-perform the Offsite Power restoration step.

Question ID (Status) RO32NRC-094(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

OHP-4023-ECA-0.0, Loss of All AC Power Source: New KA - 194001 2.1.6 Generic Conduct of Operations Ability to manage the control room crew during plant transients.

RO - 3.8 SRO - 4.8 CFR - 41.10 / 43.5 / 45.12 / 45.13 K/A Justification: Question requires the candidate to direct and coordinate the implementation and use of Emergency and Equipment Restoration procedures by the operating crew.

SRO Justification: Question deals with execution of procedure steps and hierarchy.

Associated objective(s):

(RO-C-EOP14-E7) For each of the ECA-0 series procedures, identify the Major Action Categories and discuss the bases for each.

DC COOK OPS Page: 191 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 95 ID: RO26-0153 Points: 1.00 SRO ONLY OPEN REFERENCE Given the following plant conditions:

You are the Unit 2 Unit Supervisor. Maintenance has completed an inspection of the Unit 2 East CCP and has signed off the clearance (which was on the Breaker and Control Switches only).

A credible insider threat has resulted in activation of the Vital Area Two-Person Line-Of-Sight Rule.

Unit 2 has experienced a reactor trip from 100% power.

A loss of Bus T21A due to a fault on the bus occurred on the reactor trip.

An AEO requests permission to suspend the Two-Person Line-Of-Sight Rule so that he may proceed to rack in the Unit 2 East CCP.

There are no other operators currently available to assist him.

The Two-Person Line-Of-Sight Rule:

A. may NOT be suspended but you may assign a Security officer to accompany the Operator to rack the breaker in.

B. may NOT be suspended, you must wait until another operator becomes available.

C. may be suspended if approved by a Security Officer and the Shift Manager.

D. may be suspended if approved by you and the Shift Manager.

Answer: D Answer Explanation:

A. Incorrect - The rule requires two individuals that are knowledgeable with the task being performed. A security officer would not be familiar with the task.

B. Incorrect - The rule could be suspended for plant safety C. Incorrect - Suspension of the rule requires an SRO and SM.

D. Correct - Suspension of the rule is allowed with Shift manager and SRO concurrence only if personnel or plant safety would be adversely impacted. The Loss of All CCP could lead to a loss of RCP seals, so the rule could be suspended DC COOK OPS Page: 192 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO for plant safety.

Question ID (Status) RO26-0153(Active)

External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

PMP-2060-SEC-006 Security Requirements for Plant Personnel Attachment 4 Attachment Provided: PMP-2060-SEC-006 Security Requirements for Plant Personnel Attachment 4 Source: Bank Previous NRC exam: 2006 KA - 194001 2.1.13 Generic Conduct of Operations Knowledge of facility requirements for controlling vital / controlled access.

RO - 2.5 SRO - 3.2 CFR - 41.10 / 43.5 / 45.9 / 45.10 SCLR - 1P K/A Justification - This question tests ability of SRO to make correct determination based on plant conditions of whether to suspend requirements for access to vital areas.

SRO Justification: The decision to waive the requirements are a procedurally driven SRO task.

Question Source - NRC Exam 2006-95-10, COOK 2004 Q#93 Associated objective(s):

DC COOK OPS Page: 193 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 96 ID: NRCAUDIT07-0962-SROA Points: 1.00 SRO ONLY - OPEN REFERENCE Given the following:

An On The Spot Change (OTSC) has been written to a surveillance procedure to run the North Safety Injection pump with the discharge valve throttled 75% open and collect motor data.

The plant conditions required for the above evolution are NOT described in current procedures or the Updated Safety Analysis Report.

The OTSC author is the System Engineer, who has brought it to you for review and approval.

The appropriate action for the SRO is to:

A. NOT approve until the Qualified Technical Review has been performed.

B. review and approve the OTSC without restriction.

C. NOT approve. A revision of the surveillance procedure is required.

D. review and approve the OTSC ONLY if a 50.59 screening/evaluation has been approved.

Answer: C Answer Explanation:

The Throttling of the Pump Discharge valve changes the intent of the procedure and therefore an OTSC is Not appropriate. This requires following the normal alteration process.

A - Incorrect - an OTSC is not appropriate and a 50.59 evaluation is required.

B - Incorrect - an OTSC is not appropriate and a 50.59 evaluation is required.

C. Correct - See above D - Incorrect - an OTSC is not appropriate.

Question ID (Status) NRCAUDIT07-0962-SROA (Active)

External Topic ID: ADMN-12: ADMN-12, Procedures Level/Difficulty: F/2 DC COOK OPS Page: 194 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Comments:

Reference:

PMP-2010-PRC-002 Procedure Alteration, Review, and Approval Figures 2, 4, and Attachment 3 Attachment Provided: PMP-2010-PRC-002 Procedure Alteration, Review, and Approval Figures 2 & 4 and Attachment 3 Source: Bank KA - 194001 2.2.11 Generic Equipment Control Knowledge of the process for controlling temporary design changes.

RO - 2.3 SRO - 3.3 CFR - 41.10 / 43.3 / 45.13 KA Justification: Question requires knowledge of whether a change to a surveillance procedure follows the process to ensure the safety margin as defined in technical specification is not reduced.

SRO Justification: Question requires SRO knowledge of admin procedures dealing with procedure changes and the 50.59 screening process.

Associated objective(s):

(RO-C-ADM12-E11) Given PMP-2010-PRC-002 Procedure Alteration, Review, and Approval, determine if a procedure deficiency meets the requirement for performing an OTSC.

DC COOK OPS Page: 195 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 97 ID: CM-7159-SRO Points: 1.00 SRO ONLY - OPEN REFERENCE The plant is at 100% power.

It has just been discovered that a required quarterly surveillance test for a component required to be operable in modes 1, 2 and 3 has not been completed

The surveillance interval expired two months ago.

A risk evaluation has NOT been performed.

There are no indications of any problems with the component.

At this time, is the LCO concerning this surveillance being met, including why?

A. Yes. Declaring the LCO not met may be delayed for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform the surveillance regardless of the time since the last performance.

B. Yes. Since it is believed that the component will perform its intended function, a new surveillance interval may be declared to avoid entry into the ACTIONS.

C. No. SR 3.0.1 requires that the component be declared inoperable, the LCO be declared not met, and the associated Required Actions be performed.

D. No. SR 3.0.3 only applies if the time from when the SR 3.0.2 interval is exceeded is not greater than the specified Frequency.

Answer: A Answer Explanation:

A. Correct - Since the Surveillance was not performed within the required frequency SR 3.0.3 allows a 24-hour period to perform the surveillance.

B. Incorrect - The interval is not reset and must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. Incorrect - Since the only failure is the omission of the surveillance the delay time is allowed without declaring the equipment inoperable.

D. Incorrect - SR 3.0.3 is applicable Question ID (Status) CM-7159-SRO(Active)

External Topic ID: TS01-1: TS01-1, Technical Specifications/TRM - LCO/TRO Recognition Level/Difficulty: H/2 SRO ONLY DC COOK OPS Page: 196 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO Comments:

Reference:

Technical Specifications SR 3.0.3 Reference Provided TS SR 3.0.1 - SR 3.0.4 Source: Bank KA - 194001 2.2.22 Generic Equipment Control Knowledge of limiting conditions for operations and safety limits.

RO - 4.0 SRO - 4.7 CFR - 41.5 / 43.2 / 45.2 KA Justification: Question requires knowledge of Technical Specification LCO implementation.

SRO Justification: Question requires SRO knowledge of TS usage and application rules.

Associated objective(s):

(RO-C-TS01-E11) For a specific plant system or component, apply Technical Specifications to determine:

If the system or component is Tech. Spec. required

Limiting Conditions for Operation

Applicable Modes

Actions required

Surveillance Requirements

Bases DC COOK OPS Page: 197 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 98 ID: CM-1166-SROA Points: 1.00 SRO ONLY Which ONE of the following responses correctly reflects the bases for Reactor Coolant Specific Activity in Technical Specification?

A. The short-lived radioactive isotope fission products will have decayed prior to any fuel movement.

B. Limitations on specific activity in the RCS reduces corrosion product activation and subsequent challenges to RCS integrity.

C. Limitations on the allowable concentrations of radionuclides are established to prevent exceeding 10CFR20 limits during a LOCA.

D. Limitations on the allowable concentrations of radionuclides are established to minimize the dose consequences during a steam line break or a steam generator tube rupture.

Answer: D Answer Explanation:

A. INCORRECT. See Answer D. Plausible as this is a concern during refueling evolutions for dose to employees moving fuel.

B. INCORRECT. See Answer D. Plausible as this identifies a challenge to corrosion affecting the metal in the RCS for long term RCS operation.

C. INCORRECT. See Answer D. Plausible for accident dose concerns but the CFR reference is incorrect, and the TS is specifically written for SGTR and SLB.

D. CORRECT. The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67 (Ref. 1). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.

Question ID (Status) CM-1166-SROA (Active)

DC COOK OPS Page: 198 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO External Topic ID:

Level/Difficulty: F/3 Comments:

Reference:

TS and Bases 3.4.16 RCS Specific Activity Source: Bank KA - 194001 2.3.14 Generic Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

RO - 3.4 SRO - 3.8 CFR - 41.12 / 43.4 / 45.10 KA Justification - Question requires operator knowledge of the radiation hazards and mitigating actions required for high RCS activity.

SRO Justification: Question requires knowledge of TS bases.

Associated objective(s):

(RO-C-00200-E13) Given a description of plant conditions and a copy of the Technical Specifications, determine the following for the Reactor Coolant System:

a. All applicable Condition(s) and Required Action(s)

b. The most limiting Condition(s) and Required Action(s)without error.

DC COOK OPS Page: 199 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 99 ID: CM-7713 Points: 1.00 SRO ONLY The following sequence of events has occurred:

The unit has tripped and experienced a Safety Injection.

While performing 02-OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization, an ORANGE path condition was noted for the Core Cooling Critical Safety Function

2-OHP-4023-FR-C.2, Response to Degraded Core Cooling was entered

While performing steps of this procedure, the Shift Technical Advisor reports a RED path condition exists for both Core Cooling and Containment Critical Safety Functions

NO other abnormal conditions were noted NOTE:

2-OHP-4023-FR-C.1, Response to Inadequate Core Cooling 2-OHP-4023-FR-Z.1, Response to Containment High Pressure Based on these plant conditions, which one of the following is the appropriate action for the Unit Supervisor to take?

A. Stop performing 02-OHP-4023-FR-C.2, and immediately transition to 02-OHP-4023-FR-C.1.

B. Complete actions of 02-OHP-4023-FR-C.2, and then transition to 02-OHP-4023-FR-Z.1.

C. Complete actions of 02-OHP-4023-FR-C.2, and then transition to 02-OHP-4023-FR-C.1.

D. Stop performing 02-OHP-4023-FR-C.2, and immediately transition to 02-OHP-4023-FR-Z.1.

Answer: A Answer Explanation:

Rules of usage for functional restoration procedures dictate leaving a lower priority procedure immediately to perform the highest priority procedure which is required.

DC COOK OPS Page: 200 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO A. CORRECT - Immediately go to highest priority red path (C-1).

B. INCORRECT - Plausible if rules of usage misapplied to complete a functional restoration procedure which has been entered and then transition; this would also require incorrect prioritization of Z over C.

C. INCORRECT - Plausible if rules of usage misapplied to complete a functional restoration procedure which has been entered and then transition to the highest priority procedure which is required.

D. INCORRECT - Plausible if rules of usage correctly applied to immediately leave a lower priority functional restoration procedure when a higher priority procedure exists; this would also require incorrect prioritization of Z over C.

Question ID (Status) CM-7713 (Active)

External Topic ID:

Level/Difficulty: H/3 Comments:

REFERENCE:

OHI-4023 Abnormal/Emergency Procedure User's Guide, Attachment 5 Section 5 Source: Bank Previous NRC exam: 2010 KA - 194001 2.4.21 Generic Emergency Procedures/Plan Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

RO - 4.0 SRO - 4.6 CFR - 41.7 / 43.5 / 45.12 KA Justification: Question requires knowledge of the safety function assessment for core cooling.

SRO Justification: Question requires SRO knowledge of procedural hierarchy.

Associated objective(s):

(RO-C-EOP10-E2) Define Saturated Core Cooling (SCC).

DC COOK OPS Page: 201 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO 100 ID: CM-1046 Points: 1.00 SRO ONLY The Shift Manager is relieved of Site Emergency Director (SED) duties when:

A. an Alert or higher condition is declared and the Operations Support Center (OSC) has completed accountability.

B. the EOF has been activated and all positions are staffed, and responsibility for offsite communications has been transferred.

C. the initial offsite notifications are complete and the Technical Support Center (TSC) is declared operational.

D. the TSC has been activated, the on-coming SED has been briefed, and the Emergency Turnover Checklist completed.

Answer: D Answer Explanation:

From RMT-2080-CHK-001, Site Emergency Director Checklist WHEN the TSC has been activated, the on-coming SED has been briefed, and the Emergency Turnover Checklist completed, THEN conduct turnover and transfer Command and Control Function using Data Sheet 2, Emergency Turnover Checklist.

A. Incorrect - See above.

B. Incorrect - See above.

C. Incorrect - See above.

D. Correct - See above Question ID (Status) CM-1046(Active)

External Topic ID: EP01: EP01, Emergency Preparedness Overview Level/Difficulty: F/2 Comments:

Reference:

PMP-2080-EPP-100, Emergency Response step 3.2 SM-SED Checklist and data Sheet #2 Emergency Turnover Checklist, RMT-2080-CHK-001, Site Emergency Director Checklist Source: Bank DC COOK OPS Page: 202 of 203 26 February 2020

EXAMINATION ANSWER KEY RO32NRC-SRO KA - 194001 2.4.40 Generic Emergency Procedures/Plan Knowledge of SRO responsibilities in emergency plan implementation.

RO - 2.7 SRO - 4.5 CFR - 41.10 / 43.5 / 45.11 KA Justification: Question requires knowledge of when the turnover for site emergency director is complete.

SRO Justification: Question addresses SRO task of SED.

Associated objective(s):

(ST-C-EP01-E4) Describe the turnover process for the SM to be relieved of SED duties.

In accordance with PMP-2080-EPP-100, Emergency Response.

DC COOK OPS Page: 203 of 203 26 February 2020

Q# Reference Provided 14 Steam Tables 79 T.S. 3.5.2, T.S. 3.8.1 80 T.S 3.3.5, T.S. 3.8.1 (see above question) 81 1-OHP-4023-FR-H-1 Fold Out Page 82 TDB 2-FIG-11 Modified 87 T.S. 3.3.1 (pg 3.3.1-1 to 3.3.1-5, 3.3.1-15) 92 2-FIG-19-19.e (completed - Attached to question) 93 T.S. 3.3.6, TRM 8.3.8 95 PMP-2060-SEC-006 Attachment 4 96 PMP-2010-PRC-002 Figures 2&4, Attachment 3 97 T.S. SR-3.0.1 to 3.0.4 References to be provided with no identification as to the question they apply to.

ML20356A257

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